POLICY ISSUE

(Information)

July 24, 2003 SECY-03-0124

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 

Contact:

Ronald C. Emrit, RES

(301) 415-6447
The Commissioners 2

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.
The Commissioners 3

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.
The Commissioners 4

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 LightWater 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.
The Commissioners 5

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, conditionmonitoring 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
The Commissioners 6

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.
The Commissioners 7

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
The Commissioners 7

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

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WMDean, AO

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OFFICE REAHFB C REAHFB C Tech Editor C C:REAHFB C D:DSARE

NAME REmrit:dfw HVanderMolen ABaranek JFlack FEltawila

DATE 06/18/03* 06/18/03* 06/23/03* 06/20/03* 06/20/03*

OFFICE D:NMSS C D:NRR C D:RES C DEDMRS C EDO C

NAME MVirgilio SCollins AThadani CPaperiello WTravers

DATE 07/01/03* 06/25/03* 07/09/03* 07/23/03 072403
Completion Schedule for Open GSIs as of July 8, 2003

GSI

Number

Title Lead

Office

Identification

Date

Completion D

Initial

Screening

Technical

Assessment

80 Pipe Break Effects on Control Rod Drive Hydraulic

Lines in the Drywells of BWR MARK I and IIContainments

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 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 HydrogenCombustion During a Severe Accident

NRR 05/2001 05/2002 12/17/2002 

191 Assessment of Debris Accumulation on PWR Sump

Performance

NRR 09/1996 09/1996 09/2001 

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 
Completion Schedule for Open GSIs as of July 8, 2003

GSI

Number

Title Lead

Office

Identification

Date

Completion D

Initial

Screening

Technical

Assessment

2 

NMSS-14 Surety Estimates for Groundwater Restoration at InSitu 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