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SECY-03-0124
July 24, 2003
| FOR: |
The Commissioners |
| FROM: |
William D. Travers
Executive Director for Operations /RA/ |
| SUBJECT: |
SUMMARY OF ACTIVITIES RELATED TO GENERIC SAFETY ISSUES |
PURPOSE:
To provide the annual summary of activities related to generic safety
issues (GSIs).
BACKGROUND:
It has been the practice of the staff since 1983 to provide the Commission
with an annual update of the progress made in resolving GSIs. This practice
was reinforced by the Commission in a staff requirements memorandum (SRM)
of May 8, 1998, in response to SECY-98-030, "Implementation of DSI-22,
Research." In the SRM, the Commission directed the staff to provide an
annual summary of activities related to open reactor and non-reactor GSIs.
DISCUSSION:
The NRC program for addressing reactor and non-reactor generic issues
is delineated in Management Directive (MD) 6.4, "Generic Issues Program."
MD 6.4 was issued in December 2001. The program described in MD 6.4
consists of seven stages: (1) identification, (2) initial screening, (3)
technical assessment, (4) regulation and guidance development, (5) regulation
and guidance issuance, (6) implementation, and (7) verification. Candidate
generic issues may be identified by organizations or individuals internal
or external to the NRC. Generally, safety concerns associated with operating
events, research results, or risk assessments form the basis for the identification
of GSIs by the staff, the ACRS, industry, or the public. After an issue
is identified, initial screening is performed to determine whether it
should be processed as a GSI, excluded from further analysis, or sent
to another NRC program for review. In the technical assessment stage,
a determination is made as to whether the issue involves adequate protection,
safety enhancement, or burden reduction. Technical findings are used as
the basis for developing or revising rules, guidance, and programs. In
the final three stages, regulation or guidance is issued by the NRC, implemented
by licensees or certificate holders, and verified by the NRC. GSIs identified
after March 1999 have been processed in accordance with MD 6.4. The Office
of Nuclear Regulatory Research (RES) tracks the status of all generic
issues in the agency-wide Generic Issue Management Control System (GIMCS)
and documents the technical assessments and dispositions of all issues
in NUREG-0933, "A Prioritization of Generic Safety Issues."
REACTOR GSIs
For generic issues associated with nuclear reactor power plants, RES
is responsible for screening all new generic issues and performing
the technical assessments of GSIs.
The Office of Nuclear Reactor Regulation (NRR) is responsible for the
development and issuance of regulation or guidance that may be recommended
in the technical assessments, and staff verification of the resultant
regulation or guidance. An adequate protection evaluation is conducted
for each new GSI identified to determine whether plants should continue
operating while the issue is being resolved. The following is a summary
of the activities related to reactor GSIs since the last report to the
Commission in SECY-02-0148 on August 2, 2002.
Identification
Three new GSIs were identified for initial screening:
| 193 |
BWR ECCS Suction Concerns: This issue addresses
the concern for the possible failure of ECCS caused by unanticipated,
large quantities of entrained gas in the suction piping from BWR suppression
pools. The issue applies to MARK I or II containments during LOCAs,
and could potentially cause pump failure or degraded performance caused
by gas binding, vapor locking, or cavitation. |
| 194 |
Implications of Updated Probabilistic Seismic Hazard Estimates:
This issue addresses the concern for the seismic design bases of all
nuclear power plants in and around the East Tennessee Seismic Zone,
based on the new composite seismicity model for the region. |
| 195 |
Hydrogen Combustion in Foreign BWR Piping: This
issue addresses the accumulation of combustible gas mixtures in piping.
In several foreign events, hydrogen and oxygen gases apparently accumulated
to a combustible level, which then damaged the piping systems. |
Initial Screening
Initial screening of the following two GSIs was completed:
| 80 |
Pipe Break Effects on Control Rod Drive Hydraulic Lines
in the Drywells of BWR MARK I and II Containments: This issue
addresses the concern for the likelihood and effects of a LOCA that
could cause interactions with the control rod drive hydraulic lines
in such a way as to prevent rod insertion, creating the potential
for recriticality when the reactor core is reflooded. Initial screening
of the issue was completed in February 2003 and the staff is currently
developing an action plan for a technical assessment. |
| 192 |
Secondary Containment Drawdown Time: This issue
addressed the concern for the adequacy of the calculations, testing,
and acceptance criteria associated with the creation of a vacuum in
the reactor building of a BWR, following an engineered safeguards
actuation signal. Initial screening of the issue was completed in
June 2002 when a staff panel found that existing regulations are adequate
to address the safety concern, and the issue was dropped from further
pursuit. |
The following four GSIs are currently undergoing screening:
| 186 |
Potential Risk and Consequences of Heavy Load Drops:
This issue resulted from a staff review of licensees' programs for
handling heavy loads, which revealed a substantially greater potential
for severe consequences to result from the drop of a heavy load than
was previously envisioned. In pursuing the issue, the staff is preparing
a report on crane operating experience through 2002. This report will
serve the dual purpose of completing both the initial screening and
technical assessment of this GSI. |
| 193 |
BWR ECCS Suction Concerns: The staff has completed
a probabilistic screening analysis of the issue and is in the process
of forming a review panel. The ongoing reevaluation of 10 CFR 50.46,
"Acceptance Criteria for Emergency Core Cooling Systems for Light-Water
Nuclear Power Plants," will be considered in the screening of this
GSI. |
| 194 |
Implications of Updated Probabilistic Seismic Hazard Estimates:
A staff panel has completed its review of the issue and found that
existing NRC programs have adequately addressed the safety concern.
The panel has recommended that the issue be dropped from further pursuit. |
| 195 |
Hydrogen Combustion in Foreign BWR Piping: The
staff is currently performing a probabilistic screening analysis of
the issue for review by a panel. |
Technical Assessment
The following is the status of the ongoing technical assessment of five
GSIs:
| 80 |
Pipe Break Effects on Control Rod Drive Hydraulic Lines
in the Drywells of BWR MARK I and II Containments: This issue
was screened as described above and an action plan is being developed
by the staff for pursuit of a technical assessment. |
| 156.6.1 |
Pipe Break Effects on Systems and Components: This
issue addresses the safety concern of whether the effects of pipe
breaks inside the containment have been adequately addressed in the
designs of some plants. A risk analysis performed by the staff showed
the issue to have some safety significance but with large uncertainty.
A more comprehensive study was undertaken to review pipe failure rate
data and pipe break methodologies. The staff is presently conducting
a comparison of the findings of two studies used in the initial screening
of the issue, and the Task Action Plan for pursuing the issue is scheduled
to be revised in July 2003. The ongoing reevaluation of 10 CFR 50.46,
"Acceptance Criteria for Emergency Core Cooling Systems for Light-Water
Nuclear Power Plants," will also be considered in the technical assessment
of this GSI. |
| 163 |
Multiple Steam Generator Tube Leakage: This issue
addresses the safety concern associated with multiple steam generator
tube leaks during a main steam line break that cannot be isolated.
The issue is an integral part of the NRC Steam Generator Action Plan,
the status of which was presented to the Commission in SECY-03-0080
on May 16, 2003, and discussed at a Commission meeting on May 29,
2003. |
| 185 |
Control of Recriticality Following Small-Break LOCA in PWRs:
This issue addresses small-break-LOCA scenarios in PWRs that involve
steam generation in the core and condensation in the steam generators,
causing deborated water to accumulate in part of the RCS. Restart
of the RCS circulation may cause a recriticality event (reactivity
excursion) by moving the deborated water into the core. Specific recommendations
on the proposed course of action are scheduled to be completed in
October 2003. |
| 188 |
Steam Generator Tube Leaks/Ruptures Concurrent with Containment
Bypass: This issue addresses the effects on the validity
of steam generator tube leak and rupture analyses of resonance vibrations
in steam generator tubes during steam line break depressurization.
The issue is an integral part of the NRC Steam Generator Action Plan,
the status of which was presented to the Commission in SECY-03-0080
on May 16, 2003, and discussed at a Commission meeting on May 29,
2003. |
Regulation and Guidance Development
Regulation and guidance development continued on the following three
GSIs:
| 168 |
Environmental Qualification of Electrical Equipment:
Accelerated-aging tests on electrical equipment showed that some of
the environmentally qualified cables either failed or exhibited marginal
insulation resistance. Failure of the cables during or following a
design basis event could affect the performance of safety functions.
(The results of Okonite's testing of its single conductor, bonded-jacket
cables were specifically considered by the staff as it explored voluntary
industry initiatives to resolve the issue.) Regulatory Issue Summary
2003-09 was issued on May 2, 2003. After review and analysis of six
LOCA tests, condition-monitoring tests on instrument and control (I&C)
cables, and information provided by the nuclear industry, the staff
concluded that the existing equipment qualification process is adequate
for assuring that I&C cables will perform their intended function.
The staff is in the process of closing out this issue by August 2003. |
| 189 |
Susceptibility of Ice Condenser and MARK III Containments
to Early Failure from Hydrogen Combustion During a Severe Accident:
This issue was identified with the issuance of NUREG/CR-6427, "Assessment
of the Direct Containment Heat (DCH) Issue for Plants with Ice Condenser
Containments," when it was discovered that the early containment failure
probability in ice condensers is dominated by non-DCH hydrogen combustion
events. The issue was extended to include BWR MARK III containments
since their relatively low free volume and strength are comparable
to PWR ice condensers. The staff concluded that further action to
provide back-up power to one train of igniters is warranted for plants
with ice condenser or MARK III containments. The staff is currently
engaging the affected stakeholders in developing additional information
related to implementing various alternatives, including an option
of using the severe accident management guidelines. A stakeholders
meeting was held on June 18, 2003, and the staff is evaluating comments
received from licensees to determine whether rulemaking should be
pursued. |
| 191 |
Assessment of Debris Accumulation on PWR Sump Performance:
This issue addresses the possibility of debris accumulating on the
emergency core cooling system (ECCS) sump screen, resulting in the
loss of net positive suction head (NPSH) margin. This loss of NPSH
could impede or prevent the flow of water from the sump necessary
to meet the criteria of 10 CFR 50.46. In its technical assessment,
the staff concluded that additional actions may be warranted to ensure
an adequate NPSH margin for PWR ECCS pumps taking suction from containment
sumps. The PWR industry, with NRC oversight, is developing technical
guidance for plant-specific analyses to determine whether debris accumulation
will impede or prevent ECCS operation. Following meetings with stakeholders
on March 5 and April 29, 2003, NRC Bulletin 2003-01 was issued to
PWR licensees on June 9, 2003, to (1) confirm their compliance with
10 CFR 50.46(b)(5) and other existing applicable regulatory requirements,
or (2) describe any compensatory measures that have been implemented
to reduce the potential risk due to post-accident debris blockage,
as evaluations to determine compliance proceed. |
Since the inception of the generic issues program in 1976, the staff
has closed 833 of the 845 reactor generic issues identified. The scheduled
completion dates of the various stages of initial screening, technical
assessment, or regulation and guidance development for the 12 open reactor
GSIs are summarized in the Attachment.
NON-REACTOR GSIs
NMSS has the primarily responsibility for processing non-reactor GSIs
through all stages of MD 6.4. The status of the unresolved non-reactor
GSIs is tracked by RES in the quarterly updates of GIMCS. The following
is a summary of the activities related to non-reactor GSIs since the last
report to the Commission in SECY-02-0148:
Identification
No new GSIs were identified for screening.
Technical Assessment
The following is the status of the ongoing technical assessment of two
GSIs:
| NMSS-7 |
Criticality Benchmarks Greater than 5% Enrichment:
The staff is developing and confirming the adequacy of tools for validating
criticality calculations, including requests to process higher enrichments,
to be used in licensing nuclear facilities. This GSI has temporarily
been placed on hold due to the need to fund higher priority tasks
while plant-specific aspects of the issue are being pursued. The staff
expects to meet the current completion date for its technical assessment
by building on the plant-specific efforts being performed. |
| NMSS-14 |
Surety Estimates for Groundwater Restoration at In-Situ
Leach Facilities: This issue addresses the development of
methodologies to (1) calculate surety for ground-water restoration
activities at in situ leach uranium extraction facilities, and (2)
monitor post-restoration ground-water quality stability. |
Regulation and Guidance Development
Regulation and guidance development continued on the following GSI:
| NMSS-16 |
Adequacy of 0.05 Weight Percent Limit in Part 40:
Options on how to proceed with jurisdictional and technical issues
on the regulation of source material were forwarded to the Commission
in SECY-00-0201 in September 2000. In accordance with the SRM issued
for SECY-00-0201, the transfer provision rule was published for public
comment in the Federal Register on August 28, 2002. The staff is currently
evaluating the comments received. |
The scheduled completion dates of the technical assessment or regulation
and guidance development stages for the three open, non-reactor GSIs are
summarized in the Attachment.
CONCLUSION:
Since the last report to the Commission on August 2, 2002, one reactor
GSI was closed and 15 GSIs remain to be resolved as the staff continued
to implement the process of MD 6.4 to identify and resolve reactor and
non-reactor GSIs. The staff will continue to provide an annual update
to the Commission on activities related to GSIs and will inform the Commission
of any significant developments.
| |
/RA by Patricial Norry Acting For/
William D. Travers
Executive Director for Operations |
Attachment: Completion Schedule for Open GSIs as
of July 8, 2003
| Contact: |
Ronald C. Emrit, RES
(301) 415-6447 |
Attachment
| GSI Number |
Title |
Lead Office |
Identification Date |
Completion Date |
| Initial Screening |
Technical Assessment |
Regulation Guidance |
| 80 |
Pipe Break Effects on Control Rod Drive Hydraulic Lines in the Drywells
of BWR MARK I and II Containments |
RES |
03/1998 |
02/2003 |
TBD |
- |
| 156.6.1 |
Pipe Break Effects on Systems and Components |
RES |
02/1991 |
07/1999 |
TBD |
- |
| 163 |
Multiple Steam Generator Tube Leakage |
NRR |
06/1992 |
01/1997 |
09/2005 |
- |
| 168 |
Environmental Qualification of Electrical Equipment |
NRR |
04/1993 |
04/1993 |
06/2002 |
TBD |
| 185 |
Control of Recriticality Following Small-Break LOCA in PWRs |
RES |
01/1999 |
07/2000 |
09/2005 |
- |
| 186 |
Potential Risk and Consequences of Heavy Load Drops in Nuclear Power
Plants |
RES |
04/1999 |
08/2003 |
08/2003 |
- |
| 188 |
Steam Generator Tube Leaks/Ruptures Concurrent with Containment
Bypass |
RES |
06/2000 |
05/2001 |
09/2004 |
- |
| 189 |
Susceptibility of Ice Condenser and MARK III Containments to Early
Failure from Hydrogen Combustion During a Severe Accident |
NRR |
05/2001 |
05/2002 |
12/17/2002 |
TBD |
| 191 |
Assessment of Debris Accumulation on PWR Sump Performance |
NRR |
09/1996 |
09/1996 |
09/2001 |
03/2007 |
| 193 |
BWR ECCS Suction Concerns |
RES |
05/2002 |
07/2003 |
- |
- |
| 194 |
Implications of Updated Probabilistic Seismic Hazard Estimates |
RES |
06/2002 |
07/2003 |
- |
- |
| 195 |
Hydrogen Combustion in Foreign BWR Piping |
RES |
02/2003 |
08/2003 |
- |
- |
| NMSS-07 |
Criticality Benchmarks Greater than 5% Enrichment |
NMSS |
05/1998 |
06/1998 |
06/2005 |
- |
| NMSS-14 |
Surety Estimates for Groundwater Restoration at In-Situ Leach Facilities |
NMSS |
06/1998 |
07/1998 |
08/2003 |
- |
| NMSS-16 |
Adequacy of 0.05 Weight Percent Limit in Part 40 |
NMSS |
06/1998 |
07/1998 |
03/2002 |
TBD |
|
