ML20057F133
| ML20057F133 | |
| Person / Time | |
|---|---|
| Issue date: | 10/12/1993 |
| From: | Meador S NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| NUDOCS 9310140168 | |
| Download: ML20057F133 (136) | |
Text
Pbl INDEX OF BRIEFING PAPERS BPI License Renewal
[NRR]
BP2 Design Certification Process
[NRR]
BP3 Below Regulatory Concern
[RES]
BP4 Resident Inspector Program
[NRR]
BPS Assistance to Regulatory Bodies of Russia & Ukraine
[0E00]
BP6 Emergency Planning and Preparedness
[NRR]
BP7 Emergency Response Data System (ERDS)
[AE0D]
BP8 Systematic Assessment of Licensee Performance (SALP)
[NRR]
BP9 Chernobyl Status
[RES]
BP10 Three Mile Island 2 Status
[NRR]
BPil Licensee Performance Indicators
[AE0D]
BP12 Nuclear Reactor Risk to the Public
[NRR]
BP13 Advanced Reactors Program
[NRR]
BP14 High-level Radioactive Waste
[NMSS]
BP15 Remediation of Residual Radioactivity at Non-Routine Contaminated Sites i
[NMSS]
BP16 Enforcement Conference
[0E]
BPl7 Inspection of Nuclear Power Plants
[NRR]
I 9310140168 931012 PDR ORG NE E
BP18 Environmental Effects of Nuclear Power Plants
[NRR]
BP19 Uranium Mill Tailings
[NMSS]
BP20 State Compliance with the 1993 and 1996 Milestones of the Low Level Radioactive Waste Policy Amendment Act of 1985
[NMSS]
BP21 Incident Investigation Teams
[AE0D]
BP22 Rosemont Transmitters
[NRR]
BP23 Harassment and Intimidation in the Nuclear Industry
[0E]
BP24 Decommissioning of Nuclear Power Plants
[NRR]
I BP25 BWR Water Level Instrument Errors Due to Noncondensible Gas
[NRR]
BP26 Fire Barrier Issues
[NRR]
BP27 Steam Generator Tube Plugging Criteria 1
[NRR]
j BP28 Medical Issues
[NMSS]
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BP1 (09/93)
Background:
Nuclear power provides approximately 20 percent of the electric power produced in the United States.
The term of the initial operating license for a nuclear power plant is limited to 40 years.
The first plant's 40-year operating license will expire in the year 2000 and approximately 20 percent of the rest will expire by the end of the year 2010.
The Department of Energy estimates that by the year 2000, significant new electric generating capacity will be needed.
The timely renewal of these operating licenses for an additional 20 years, where appropriate to do so, represents an important contribution to ensuring an adequate energy supply for the nation during the first half of the 21st century.
License Renewal:
The Atomic Energy Act limits commercial power reactor licenses to 40 years but also permits the renewal of such licenses.
Two rules (10 CFR Part 54 and Part 51) and supporting regulatory guidance establish the technical steps, the procedural steps, and the criteria to determine if regulatory requirements are met for license renewal.
License renewal is based on two key principles.
The first principle is that the regulatory process is adequate to ensure that the licensing basis of all currently operating plants provides an acceptable level of safety, with the exception of age-related degradation unique to license renewal and possibly some few other issues during extended operation.
The second l
principle is that each plant's current licensing basis is l
required to be maintained during the renewal term including i
management of aging of systems, structures, and components needed to ensure safe operation and accident mitigation.
In other i
words, the foundation of license renewal rests on the determination that currently operating plants were initially shown to have adequate levels of safety and this level has been enhanced through evolution of the licensing bases.
Additionally, i
NRC regulatory activities have provided ongoing assurance that the licensing bases continue to provide an acceptable level of safety.
Specifically, a nuclear power plant may apply to the Commission to renew its license for a period of 20 years or less.
This application would be subject to public hearings - a formal, adjudicatory process.
The license renewal review will be based I
on issues primarily related to aging management and on how the renewal application addresses degradation that would occur in the period of extended operation.
The Commission expects the initial
j BP1 (09/93) license renewal review process to take approximately 5 years based on a detailed technical review and hearing process.
The Commission estimates that an applicant would need approximately 3-to 5 years to prepare its application.
An applicant may apply as early as 20 years before the expiration of its current license.
Regulations:
The decision whether to seek license renewal rests with the licensees.
They must make business decisions as to whether they are likely to satisfy NRC requirements and evaluate the costs that will be incurred to do so.
NRC's task is to establish a reasonable process and safety standards so that they can make timely decisions whether to seek license renewal.
The Commission's regulations governing license renewal are contained in 10 CFR Part 54 and require the applicant to describe and justify how they will identify and screen all the systems, structures, and components (SSCs) important to license renewal.
These SSCs include all safety related SSCs, all SSCs whose failure could directly affect safety related functions, and all SSCs subject to operability requirements contained in the facility's technical specifications limiting conditions for operation.
Further, the regulations include SSCs relied on to demonstrate compliance with the Commission's regulations for fire protection, environmental qualification, pressurized thermal shock, anticipated transients without scram, and station blackout.
The regulations recognize that in the screening process, applicants may identify areas in which new or modified programs will be developed to assess and manage age-related degradation.
i A parallel rulemaking effort concerning environmental issues related to license renewal is presently in progress.
The public comment period for the draft Generic Environmental Impact Statement (GEIS) and the proposed rule revision to 10 CFR Part 51 ended in March 1992.
This rulemaking is based on the belief that certain environmental issues should be treated generically, rather than in each plant specific licensing review.
The public comments on the draft GEIS and proposed rule have raised concerns i
related to both procedural aspects of the rule and NRC policy for j
treatment of environmental issues.
The Commission has reached agreement with the Council on Environmental Quality and the j
Environmental Protection Agency concerning the procedural l
concerns and expects to resolve the remaining policy issues in the near future.
As directed by the Commission, the staff will conduct discussions (in the form of workshops) with the commenters to resolve the policy issues.
These workshops will be held in locations convenient to the commenters but are not firmly scheduled.
The Commission's resolution of the procedural aspects of the rule and a discussion of the policy issues were made
BP1 (09/93) public in SECY-93-032, "10 CFR Part 51 Rulemaking on Environmental Review for Renewal of Nuclear Power Plant Operating Licenses," dated February 9, 1993.
The Commission's direction to conduct discussions with commenters is contained in a staff 1
requirements memorandum, dated April 22, 1993.
Industry Activities:
The NRC staff has reviewed a series of industry technical reports on evaluation of age-related degradation effects on a variety of structures and components important to license renewal.
The l
staff plans to incorporate appropriate technical information from i
the industry technical reports into the standard review plan for license renewal.
In 1989, the Department of Energy and the Electric Power Research Institute selected Yankee Rowe and Monticello to be lead plants to demonstrate the license renewal process for the industry.
The Northern States Power Company (operator of Monticello) had been preparing a renewal application until late 1992 when the effort was placed on indefinite hold.
Economic and regulatory reasons were cited as determining factors.
Also in 1992, concerns regarding the integrity of Yankee Rowe's reactor pressure vessel led to a decision to suspend operations rather than seek license renewal.
Current Status:
Since publication of the final license renewal rule, a number of significant policy issues have been identified including the question of whether the maintenance and license renewal rules can be integrated further, the appropriate scope of the license renewal rule, and the appropriate interpretation of age-related degradation unique to license renewal.
The staff, in SECY 049, Implementation of 10 CFR Part 54, ' Requirements for Renewal of Operating Licenses For Nuclear Power Plants'," of March 1, 1993, and SECY-93-113, " Additional Implementation Information for 10 CFR Part 54, ' Requirements for Renewal of Operating Licenses for Nuclear Power Plants'," of April 30, 1993, proposed approaches for implementing the license renewal rule to resolve
-j industry concerns about the integrated plant assessment and
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discussed the resolution of a variety of other issues the industry identified since the final license renewal rule (10 CFR Part 54) was published in December, 1991.
After considering the staff proposals, the Commission directed the staff to convene a public workshop to examine the. extent to which greater reliance can be placed on the maintenance rule and other existing licensee activities and programs for purposes of license renewal.
The workshop will take place on September 30, 1993.
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BP1 (09/93)
Currently, several individual industry efforts are underway that have progressed to the stage of interaction with the NRC.
The Babcock & Wilcox (B&W) Owners Group, representing seven operating B&W units, has formulated a generic license renewal program and the Baltimore Gas and' Electric Company has developed a specific approach for its Calvert_ Cliffs plant.
The staff has been reviewing submitted system, structure, and component screening methodologies and meeting with utility representatives to understand the proposed approaches.
The Westinghouse Owners Group recently met with the staff to describe a 5 year program for life cycle management / license renewal.
The Nuclear Management and Resources Council (NUMARC) is the_ leading industry group interacting with the NRC on license renewal policy issues.
NUMARC is attempting to establish an industry consensus approach that will resolve their concerns.
Contact:
Scott Newberry, Director, License Renewal and Environmental l
Review Project Directorate, Associate Directorate for Advanced Reactors and License Renewal, Office of Nuclear Reactor Regulation, U.S. NRC, Washington, D.C.
20555.
(301) 504-1183.
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BP1 (09/93)
HIGHLIGHTS OF LICENSE RENEWAL Nuclear power provides approximately 20 percent of the e
electric power produced in the United States.
License renewal could represent an important contribution to e
the nation's future energy supply.
Atomic Energy Act limits initial licenses to 40 years but allows for renewal of the license.
Code of Federal Regulations 10 CFR Part 54 allows for license renewal up to 20 years and provides procedures and requirements for license renewal applications.
Application review focused on age-related degradation of structures, systems, and components important to license renewal.
Generic Environmental Impact Statement and amendment to 10 CFR 1
e Part 51 address environmental issues associated with license renewal of nuclear power plants.
)
l Applicants are encouraged to submit their applications at e
least 5 years prior to expiration of current license.
Yankee Rowe's decision to decommission announced February e
1992.
Monticello announced an indefinite delay in making a decision on whether or not to seek a renewed license.
(November 1992)
Babcock and Wilcox Owners Group'is pursuing generic resolution e
of license renewal issues affecting the B&W plant design.
Baltimore Gas and Electric Company is currently. working with staff to resolve license renewal issues for their Calvert Cliffs Nuclear Power Plant.
Commission papers, SECY-93-049, " Implementation of 10 CFR e
Part 54, ' Requirements for Renewal of Operating Licensees For Nuclear Power Plants'," of March 1, 1993, and SECY-93-113,
" Additional Implementation Information for 10 CFR Part 54,
' Requirements for Renewal of Operating Licenses for Nuclear Power Plants'," of April 30, 1993, describe the staff's proposed approach for implementing the integrated plant assessment and discusses significant license renewal issues.
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BP1 (09/93)
In SECY-93-049 the staff proposed to incorporate the areas of e
technical agreement in the industry technical reports into the standard review plan for license renewal instead of in SERs.
The Commission directed this be done in the SRM dated June 28, 1993.
e The SRM dated June 28, 1993 directed that a public workshop be held to receive industry and public input on how best to resolve license renewal issues.
The workshop will take place on September 30, 1993.
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BP2 (09/93)
DESIGN CERTIFICATION PROCESS
Background:
The Commission has long sought nuclear power plant standardization and the enhanced safety and licensing reform which standardization could make possible.
The Commission has also sought to improve the licensing environment for advanced nuclear power reactors by minimizing the uncertainty in the regulatory process.
To do this, the Commission has promulgated 10 CFR Part 52 (54 FR 15372; April 18, 1989) which sets out a sensible and stable procedural framework for standardization.
This framework involves early design consideration and certification, through rulemaking, of futur designs.
The design certification process is the key procedure device in the Commission's regulations for bringing about the sought-after goal of enhanced safety and early resolution of licensing issues.
The goals and objectives of Part 52 have been reaffirmed in the Energy Policy Act of 1992.
Design Certification:
The Associate Directorate for Advanced Reactors and License Renewal was established, in part, to provide a focal point for early staff interactions with the sponsors of advanced reactor plant designs.
The review process for advanced reactor plant designs leading to design certification is specified in Subpart B of 10 CFR Part 52.
These reviews are expected to result in the certification of individual designs following the completion of a rulemaking process.
Design certifications are issued for a duration of fifteen years.
Regulations:
The Commission's regulations governing the design certification process are contained in 10 CFR Part 52 and require an applicant to provide the technical information necessary to demonstrate compliance with the standards set out for construction permits and operating licenses in 10 CFR Parts 20, 50, 73, and 100, as those standards are technically relevant to the design of the proposed facility.
Subpart B of 10 CFR Part 52 sets out the requirements and procedures applicable to Commission issuance of rules granting standard design certification for_ nuclear power facilities separate from the filing of an application for a construction permit or combined license for such a facility.
Appendix 0 of 10 CFR Part 52 sets out procedures for the filing, staff review, and referral to the Advisory Committee on Reactor Safeguards of standard designs for an advanced nuclear power reactor.
BP2 (09/93) 1 A key provision of 10 CFR Part 52 relates to the involvement of the public in the design certification process.
Public participation is more effective under Part 52 because the bulk of the issues are considered in the design certification part of the process.
Under 10 CFR Part 52, applicants must provide information related to the Three Mile Island Requirements set forth in 10 CFR 50.34, the postulated site parameters, the resolution of unresolved and generic safety issues, and design-specific probabilistic risk l
assessment.
Also, applicants must provide inspection, tests, analyses, and acceptance criteria (ITAAC) and interface l
requirements.
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Current Status:
The status of advanced light water reactor reviews are based on the schedules presented in SECY-93-097, " Integrated Review Schedules for the Evolutionary and Advanced Light-Water Reactor j
Projects."
Specifically:
GE Nuclear Energy (GE) Advanced Boiling Water Reactor (ABWR):
The staff's safety evaluation report is expected to be issued to the Commission and ACRS by the end of December 1993.
ABB-Combustion Engineering System 80+:
The staff's safety evaluation report is expected to be issued to the Commission and ACRS by the end of February 1994.
Westinghouse AP600:
The staff completed an acceptance review of the application submitted by Westinghouse and determined that it was acceptable in December 1992.
The draft safety evaluation report is expected to be issued to the Commission and ACRS in May 1994.
GE Nuclear Energy Simplified Boiling Water Reactor (SBWR):
The staff completed an acceptance review of the application i
submitted by GE and determined that it was acceptable in March 1993.
The draft safety evaluation report is expected to be issued to the Commission and ACRS in September 1994.
Contact:
l R.
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Borchardt, Director, Standardization Project Directorate, Associate Directorate for Advanced Reactors and License Renewal, Office of Nuclear Reactor Regulation, U.S. NRC, Washington, D.C.
20555.
(301) 504-1118.
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BP2 (09/93)
HIGHLIGHTS OF DESIGN CERTIFICATION PROCESS The Commission has long sought nuclear power plant e
standardization.
Standardization will enhance safety.
e Licensing environment improved by minimizing uncertainty.
Design certification is the key procedural device for bringing about early resolution of safety issues.
Design certification review process specified in Subpart B of 10 CFR Part 52.
Design certification revieva assure compliance with standards set out for construction and operating licenses in 10 CFR Parts 20, 50, 73, and 100.
10 CFR Part 52 makes public participation more effective by e
moving the bulk of the issues up front to the design certification stage.
i Applicants must provide information related to the Three e
Mile Island requirements, the postulated site parameters, the resolution of unresolved generic safety issues, design-specific probabilistic risk assessment, ITAAC, and the interface requirements.
i The goals and objectives of Part 52 were reaffirmed in the e
Energy Policy Act of 1992.
i Commission briefing on the status of advanced reactors i
e reviews - June 2, 1993 (SECY-93-097).
GE Nuclear Energy advanced boiling water reactor final safety evaluation report is scheduled to be completed by the end of December 1993.
ABB-Combustion Engineering System 80+ final safety e
evaluation report is scheduled to be completed by the end of February 1994.
Westinghouse AP600 application was accepted December 1992, e
and the design certification process is on-going.
GE Nuclear Energy simplified boiling water reactor e
application was accepted March 1993, and the design certification process is on-going.
BP3 (09/93) i BELOW REGULATORY CONCERN
Background:
On July 3, 1990, the Commission published a Below Regulatory Concern (BRC) Policy Statement in the Federal Reaister.
The BRC Policy was intended to guide a broad range of Commission actions, including exemptions from Commission regulations, as well as the development of generic health and safety standards such as those involved in the rulemaking on radiological criteria for i
decommissioning.
Subsequent to the publication of the BRC Policy, the Commission placed an indefinite moratorium on the implementation of the BRC Policy because of the broad public concern expressed over the new Policy.
Section 2901 of the recently enacted National Energy Policy Act of 1992 (H.R. 1776) revoked the Commission's July 1990 BRC Policy Statement.
Section i
2901 also revoked the Commission's policy statement of August 29, 1986 that established criteria to guide Commission exemption decisions on specific low-level radioactive waste streams.
This latter policy had been developed in order to comply with Section 10 of the Low-level Radioactive Waste Policy Amendments Act of 1985.
The Commission issued a formal withdrawal of these two policy statements in the Federal Reaister on August 24, 1993 (58 FR 44610).
BRC Policy Statement:
The BRC Policy was intended to provide a framework for making a
decisions on whether to grant specific exemptions in categories such as:
(1) the cleanup or release of sites containing residual radioactivity; (2) the distribution of consumer products containing small amounts of radioactivity; (3) the disposal of i
certain wastes containing very low levels of radioactivity; or (4) the recycling or reuse of radioactive materials that have very low levels of radioactivity.
Materials in the above categories with sufficiently low levels of radiation would have been exempt from regulatory controls.
However, the issuance of the BRC Policy resulted in extensive comment and public concern.
The public reaction resulted in the introduction of legislation on the national level, as well as by a number of State and local governments, that would prevent the BRC Policy from taking effect.
Current Status:
After the Commission placed the indefinite moratorium on the implementation of the BRC Policy, it decided to initiate rulemaking to address the critical need for generic site cleanup and decommissioning standards for NRC-licensed facilities.
The P
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BP3 (09/93)
Commission determined that it should proceed with a fresh l
approach to the development of these standards that is independent of the-now defunct BRC Policy.
The Commission intends to enhance the participation of affected interests in the rulemaking by soliciting commentary from these interests on the rulemaking issues before the staff develops the draft proposed i
rule.
The Commission has just completed a series of workshops to solicit commentary from affected interests on the fundamental approaches and issues that must be addressed in establishing the radiological criteria for decommissioning.
The workshops were held in various locations throughout the United Jtates beginning in January 1993 and were open to the public.
The NRC staff will l
use the workshop comments in the development of a draft proposed rule-on the site cleanup criteria for decommissioning.
It is-anticipated that the Commission will issue the proposed rule for public comment in June 1994.
l Contacts:
Francis X. Cameron, OGC (301) 504-1642 l
Donald A.
Cool, RES (301) 492-3785 Cynthia G. Jones, NMSS (301)504-2629 Michael F. Weber, NMSS (301)504-1298 i
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HIGHLIGHTS o
The BRC Policy Statement was an attempt to establish a framework for making decisions on whether to grant specific exemptions from Commission regulations.
o The categories considered for possible specific exemptions were:
(1) cleanup or release of sites containing residual radioactivity; (2) distribution of consumer products containing small amounts of radioactivity; (3) disposal of slightly contaminated wastes; or (4) recycling or reuse of radioactive materials that have very low levels of radioactivity.
o The BRC Policy Statement was issued on July 3, 1990, in the Federal Reaister.
o In response to public concern, the Commission declared a moratorium on the implementation of the BRC policy on June 28, 1991, o
With this moratorium, NRC initiated a phased consensus-building process on BRC issues.
o On November 12, 1991, the last of the potential representatives of environmental interests informed NRC that I
l it would be unable to represent a major interest in the BRC consensus process.
j o
At this time, due to the lack of participation in the environmental community, the Commission has decided to abandon its effort to build consensus.
o As an alternate to the consensus process, the Commission initiated an enhanced participatory rulemaking that would provide early access to affected interests on the development of radiological criteria for decommissioning.
o With respect to the Agreement State compatibility issue, the Commission has recently initiated the development of a general policy on compatibility.
The major issues in developing such a policy were discussed in a public workshop on July 26 and 27, 1993.
o The Commission is continuing its moratorium on implementation of the BRC policy and issued a Federal Register Notice on August 24, 1993 (58 FR 44610) that formally withdrew the BRC Policy of 1986 and the BRC Policy of 1990 in accordance with the National Energy Policy Act of 1992.
BP4 (09/93)
RESIDENT INSPECTOR PROGRAM
Background:
The resident inspection program is an important element of the NRC mission to protect public health and safety.
The responsibility for safe operation of a nuclear power plant lies with the licensee.
The NRC inspection program is designed to make selective examinations to ensure that this responsibility is l
being met.
The NRC inspection program is oriented toward audits; l
thus, it does not examine every activity or item, but attempts to verify through carefully selected samples, that the activities are being properly conducted or operated to enhance or ensure safety.
Resident Inspectors:
In 1977, the NRC initiated a program to station resident inspectors at each nuclear power plant under construction and in operation.
Since that time, the program has expanded to the point where normally at least two resident inspectors are assigned to each site with a nuclear power plant.
The onsite resident inspectors live in the area of the nuclear power plant.
They maintain offices at the plant and are normally available during regular business hours.
In addition, resident inspectors spend a portion of their time at the plant during weekends and evenings.
By assigning resident inspectors to reactor sites, the NRC was able to significantly increase the amount of time inspectors spend at the plant.
This increased time provides a greater opportunity to observe and measure licensee activities, verify licensee compliance with NRC requirements, and respond to operational events at the plant.
Since a resident inspector is assigned to a single site, the resident inspector acquires more detailed knowledge of that plant and is able to provide more efficient inspections.
The resident inspector provides a continual inspection and I
regulatory presence, as well as a direct contact between NRC management and the licensee.
The resident inspector is also the key individual in the regional offices's determination of what additional inspection activities need to be accomplished at a specific plant.
The inspection activities of the resident inspector are supplemented by the efforts of engineers and specialists from the regional office staff who perform inspections in a wide variety of engineering and scientific disciplines ranging from civil and structural engineering to health physics and core physics.
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Regulations:
The U.S. Atomic Energy Commission (AEC) was created by an act of Congress in 1946.
In the Atomic Energy Act of 1954, the AEC was vested with developmental and regulatory functions related to peaceful uses of atomic energy.
In 1974, another law was passed (the Energy Reorganization Act) which abolished the Atomic Energy Commission and created two new organizations, the Energy Research and Development Administration (ERDA) and the Nuclear Regulatory Commission (NRC).
The NRC has broad authority under the Atomic Energy Act of 1954, as amended (Act).
This authority is reflected in Section 161o of the Act which provides authority for inspections.
This authority has been implemented through the promulgation of regulations, specifically 10 CFR 50.70, which requires nuclear power plant licensees to permit inspections deemed necessary by the NRC.
Current Status:
i In Fiscal Year 1993, there were 109 operating nuclear power plants and 7 plants with construction permits.
For these plants, there were 167 resident inspectors and 180 regional office inspectors.
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Contact:
i Mark W.
Peranich, Chief, Operating Reactor Inspection Section, Inspection and Licensing Policy Branch 1
US NRC, Washington, DC 20555.
(301) 504-3078 1
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i BP4 (09/93) i HIGHLIGHTS OF RESIDENT INSPECTOR PROGRAM 0
Licensee is primarily responsible for all safety, j
safeguards, and environmental measures necessary to protect the public health and the environment.
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The NRC role is to determine how well the licensee is performing and to ensure that the licensee corrects poor i
performance whenever identified.
o Section 161o of the Atomic Energy Act of 1954, as amended, provides authority for inspections.
This authority has been i
implemented through 10 CFR 50.70, which requires nuclear power plant licensees to permit inspections deemed necessary by the NRC.
0 In 1977, the NRC initiated the resident inspection program o
The resident inspection program has provided:
Increased NRC knowledge of the conditions at licensed nuclear power plants and a better technical base for regulatory actions.
Lessened reliance on the accuracy and completeness of licensee records by improving the inspectorJs ability to independently verify licensee performance.
Additional assurance that licensee management control systems are effective and licensee performance is acceptable.
o As of 1993, there were 109 operating nuclear power plants and 7 plants with construction permits.
For these plants, there were 167 resident inspectors and 180 regional office inspectors.
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BPS (09/93)
ASSISTANCE TO REGULATORY BODIES OF RUSSIA AND UKRAINE
Background:
The NRC has played a leading role during the last five years in cooperative efforts under the Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) to improve safety in the former Soviet Union, mainly through bilateral information exchanges, joint technical working groups, exchanges of safety inspectors and, recently, in specific coordinated research.
Following the breakup of the USSR in late 1991, relationships with the former republics were maintained as they had previously with the Soviet Union.
For the NRC, these relationships devolved to Russia and Ukraine inasmuch as they were the only former republics with operating nuclear power reactors.
(Lithuania, which has a reactor plant, is considered as part of Eastern Europe for assistance purposes.)
Current Status:
The seriousness of safety concerns for Soviet-designed reactors led to a major international effort to organize assistance to the newly independent states.
In May 1992, then-Secretary of State Baker announced in Lisbon, Portugal, a major program of US assistance for Russia and Ukraine.
A significant component of this program was nuclear safety.
Most of the nuclear safety projects proposed derived from the previous experience with the JCCCNRS activities.
The four components of the nuclear safety portion of the Lisbon Initiative are:
Establishment of two regional training centers, one in Russia and the other in Ukraine; Immediate operational safety enhancements for certain reactor types; Risk reduction measures for other reactor types; and, Regulatory assistance in developing consistent and effective safety standards and procedures, as well as training in their use.
Congress approved $25 million in FY 92 funds to initiate work on these proposals, with $3.1 million to support NRC regulatory assistance activities.
BP5 (09/93)
In July of 1992, the heads of the US, Russian and Ukrainian regulatory agencies met in the US to reach agreement on the action needed to implement the regulatory assistance component.
Twenty projects were identified to assist in the development of legislative, regulatory and liability frameworks for effective, independent regulatory programs.
In September 1992, the NRC signed an Inter-Agency Agreement with the United States Agency for International Development to provide direct assistance to the Russian and Ukrainian regulatory bodies.
The United States DOE signed a similar agreement to provide assistance under the Lisbon Nuclear Safety Initiative for the Newly Independent States.
Implementation began in October 1992 and will last for about three years.
In March 1993 at the annual JCCCNRS meeting in Kiev, Ukraine, the JCCCNRS organizational structure was formally recognized to include the Lisbon Initiative.
This meeting was significant in that it marks the first joint meeting of the US-Russia and US-Ukraine JCCCNRS.
The meeting officially established two separate JCCCNRS's, meeting annually on a trilateral basis, the designation of co-Chairmen from each country and the joint participation in cooperative and other selected activities.
In addition all parties agreed to extend the Memorandum of Cooperation in the Field of Civilian Nuclear Reactor Safety for another 5 years.
Currently, activities are in progress to ensure effective and timely implementation of the President's Vancouver Summit Assistance to Russia.
Additional project funding ($5 million) became available with the signing of the FY-93 IAA in June 1993 to support continuation of the NRC activities.
Contact:
Jacob F. Wechselberger, Technical Program Manager for International Activities, Office of the Executive Director for Operations, USNRC, Washington, DC 20555, (301) 504-1730
BP6 (09/93)
EMERGENCY PLANNING AND PREPAREDNESS Backaround:
Following the accident at Three Mile Island in 1979, the Nuclear Regulatory Commission (NRC) reexamined the role of emergency planning for protection of the public in the vicinity of nuclear power plants.
The Commission issued regulations requiring that before a plant could be licensed to operate, the NRC must have
" reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency."
The regulations set forth 16 emergency planning standards and define the responsibilities of licensees and State and local organizations involved in emergency response.
Emeraency Plannina and Preoaredness:
Emergency planning has been adopted as an added conservatism to the NRC's " defense in depth" safety philosophy.
Briefly stated, this philosophy:
(1) requires high quality in the design, construction and operation of nuclear plants to reduce the likelihood of malfunctione in the first instance; (2) recognizes that equipment can fail and operators can make mistakes, therefore requiring safety systems to reduce the chances that malfunctions will lead to accidents that release fission products from the fuel; and (3) recognizes that, in spite of these precautions, serious fuel damage accidents can happen, therefore requiring containment' structures and other safety features to prevent the release of fission products offsite.
The added feature of emergency planning to the defense in depth philosophy provides that, even in the unlikely event of an offsite fission product release, there is reasonable assurance that emergency protective actions can be taken to protect the population around nuclear power plants.
Reculations:
For planning purposes the Commission has defined a plume exposure pathway emergency planning zone (EPZ) consisting of an area about 10 miles in radius and an ingestion pathway EPZ about 50 miles in radius around each nuclear power plant.
EPZ size and configuration may vary in relation to local emergency response needs and capabilities as affected by such conditions as demography, topography, land characteristics, access routes, and jurisdictional boundaries.
The Commission's 16 emergency planning standards are contained in 10 CFR Part 50.47.
They cover the following topics:
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- 1. Assignment of Responsibility
- 2. Onsite Emergency Organization
'j 3.
Emergency Response Support and Resources 4.
Emergency Classification System
- 5. Notification Methods and Procedures
- 6. Emergency Communications 1
- 7. Public Education and Information 8.
Emergency Facility and Equipment
- 9. Accident Assessment
- 10. Protective Response
- 11. Radiological Exposure Control
- 12. Medical and Public Health Support
- 13. Recovery and Reentry Planning and Post-Accident Operations
- 14. Exercises and Drills AJ. Radiological Emergency Response Training
- 16. Responsibility for the Planning Effort:
Development, Periodic Review and Distribution of Emergency Plans Detailed information about emergency planning and preparedness is contained in Appendix E of 10 CFR Part 50 and in NUREG-0654 (FEHA-REP-1), a joint publication of the NRC and the Federal Emergency Management Agency (FEMA) entitled " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."
Current Status In the U.S.,
commercial nuclear power reactors are currently licensed to operate at approximately 70 sites in 33 states.
For each there are onsite and offsite emergency plans to assure that adequate protective measures are taken to protect the public in the event of a radiological emergency.
Federal oversight of emergency planning for licensed nuclear power plants is shared by the NRC and FEMA through a memorandum of understanding.
The memorandum is responsive to the President's decision of December 7, 1979, that FEMA will take the lead in offsite planning and response, his request that NRC assist FEMA in carrying out this role, and the NRC's continuing statutory responsibility for the radiological health and safety of the public.
Each licensee at each site exercises its emergency plan with offsite authorities so that State and local government emergency plans for each operating reactor site are exercised biennially, with participation of State and local governments, within the plume exposure EPZ.
CONTACT:
Robert A. Erickson, Chief, Emergency Preparedness Branch, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, (301) 504-3924
BP6 (09/93)
HIGHLIGHTS OF EMERGENCY PLANNING AND PREPAREDNESS Three Mile Island accident focused attention on emergency planning.
NRC must have reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.
i NRC regulations in 10 CFR 50.47 contain 16 emergency planning standards.
Emergency planning is part of NRC's " defense in depth" safety philosophy.
The plume exposure planning zone (EPZ) extends about 10 miles in radius around each licensed nuclear power plant.
The ingestion pathway EPZ extends about 50 miles in radius.
Details about emergency planning are contcined in Appendix E of 10 CFR Part 50 and in NUREG-0654.
The NRC and the Federal Emergency Management Agency (FEMA) share federal oversight of emergency planning for licensed nuclear power plants through a memorandum of understanding.
Nuclear power reactor licensees exercise their emergency plans with those of offsite authorities biennially.
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BP7 (09/93)
EMERGENCY RESPONSE DATA SYSTEM
Background:
As a result of the accident at Three Mile Island, Unit 2, on March 28, 1979, the Nuclear Regulatory Commission (NRC) and others recognized the need to substantially improve the NRC's ability to acquire accurate and timely data on plant conditions during emergencies.
The NRC's role in the event of an emergency is primarily to monitor the licensee to ensure that appropriate recommendations are made regarding offsite protective actions.
The NRC provides the licensee with technical analysis and logistic support, supports offsite authorities (including confirmation of the licensee's recommendations to offsite authorities), keeps other Federal agencies informed about the i
emergency and related NRC actions, and keeps the news media informed of the NRC's knowledge of the emergency.
To fulfill this emergency response role, the NRC requires reliable real-time data on plant conditions.
In the past, during an emergency, data on plant conditions was transmitted to the NRC by the licensee through the Emergency Notification System (ENS) (voice communication by telephone).
The ENS voice-only emergency communications link required excessive amounts of time for routine transmission of data and for verification or correction of data that appeared questionable.
Errors were also encountered in transcribing and interpreting voice transmitted data.
Therefore, the Emergency Response Data System (ERDS) was designed to supplement ENS.
Emergency Response Data System (ERDS):
ERDS provides the NRC Operations Center with timely and accurate information from installed onsite computer systems in the event of an emergency at a nuclear power plant.
Implementation of ERDS requires each licensee (except for Big Rock Point) to establish and maintain a computer program designed to transmit a set of 30 selected critical plant parameters.
The ERDS is activated by the licensee upon declaration of an alert or higher emergency condition.
Tests with ERDS indicate that a computer-based transmission system is far more accurate and timely than relaying information on plant conditions via telephone.
ERDS will be utilized during (1) emergencies at licensee's facilities, (2) during emergency training exercises, and (3) for periodic testing of the links with the NRC Operations Center.
Licensees will activate their ERDS links to begin data transmission as soon as possible (not to exceed one hour) after declaring an alert or higher emergency classification.
i BP7 (09/93)
ERDS data is available for use by personnel involved in responding to an emergency at the NRC Operations Center, at the Regional Office Incident Response Center, and at the NRC Technical Training Center in Chattanooga, Tennessee.
Upon written request, ERDS data is also made available to State governments within the 10 mile Emergency Planning Zone of that plant for their use in emergency response.
The State must provide its own workstation and enter into a Memorandum of Understanding with the NRC governing the use of the ERDS.
NRC and the States of Michigan, Washington, Alabama, North Carolina, Arizona, Georgia, Massachusetts, Maryland, New Jersey, New York, Tennessee, Pennsylvania, and Ohio have signed an ERDS MOU and four other States have inquired about similar MOU's with NRC.
Regulations:
ERDS was initiated at licensee facilities on a voluntary basis; however, the intent throughout this process was to ensure industry-wide implementation of ERDS.
Consequently, in parallel with the voluntary program, rulemaking was initiated to require the implementation of ERDS by all utilities (except for Big Rock Point).
On August 13, 1991, the final ERDS rule was published in the Federal Register.
The ERDS rule amended 10 CFR Part 50, and requires licensees to initiate data transmissions to the NRC ERDS computer no later than one hour after the declaration of an Alert, Site Area Emergency, or General Emergency.
The licensee is required to provide the necessary software to assemble the data and an output port for each reactor unit in its in-plant computer system.
The required emergency data is transmitted to the NRC via an NRC furnished modem over NRC furnished FTS-2000 telephone lines.
The data points to be included in the transmission are those that, to the greatest extent possible, describe the specific parameters listed in 10 CFR Part 50, Appendix F Section VI.
i These parameters are required to be transmitte6 if they are monitored on plant computers systems.
If the data for a selected plant parameter exists, but cannot be transmitted electronically from a licensee's system, then the licensee will continue to provide that data via the existing Telephone Emergency Notification System.
Each licensee must establish and maintain a configuration control program which will ensure that the NRC is notified of any changes to the ERDS on-site hardware or software.
Any hardware or software changes that affect the transmitted data points identified in the ERDS Data Point Library must be reported to the NRC within 30 days after changes are completed.
Any changes that could affect the transmission format and communication protocol to the ERDS must be provided to the NRC at least 30 days prior to the modification.
i i
BP5 (09/93)
ASSISTANCE TO REGULATORY BODIES OF RUSSIA AND UKRAINE
Background:
The NRC has played a leading role during the last five years in cooperative efforts under the Joint Coordinating Committee for Civilian Nuclear Reactor Safety (JCCCNRS) to improve safety in the former Soviet Union, mainly through bilateral information exchanges, joint technical working groups, exchanges of safety inspectors and, recently, in specific coordinated research.
Following the breakup of the USSR in late 1991, relationships with the former republics were maintained as they had previously with the Soviet Union.
For the NRC, these relationships devolved to Russia and Ukraine inasmuch as they were the only former republics with operating nuclear power reactors.
(Lithuania, which has a reactor plant, is considered as part of Eastern Europe for assistance purposes.)
Current Status:
The seriousness of safety concerns for Soviet-designed reactors led to a major international effort to organize assistance to the newly independent states.
In May 1992, then-Secretary of State Baker announced in Lisbon, Portugal, a major program of US assistance for Russia and Ukraine.
A significant component of this program was nuclear safety.
Most of the nuclear safety projects proposed derived from the previous experience with the JCCCNRS activities.
The four components of the nuclear safety portion of the Lisbon Initiative are:
Establishment of two regional training centers, one in Russia and the other in Ukraine; Immediate operational safety enhancements for certain reactor types; Risk reduction measures for other reactor types; and, Regulatory assistance in developing consistent and effective safety standards and procedures, as well as training in their use.
Congress approved $25 million in F1 92 funds to initiate work on these proposals, with $3.1 million te support NRC regulatory assistance activities.
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BPS (09/93)'
f In July of 1992, the heads of the US, Russian and Ukrainian regulatory agencies met in the US to reach agreement on the action needed to implement the regulatory assistance component.
1 Twenty projects were identified to assist in the development of legislative, regulatory and liability frameworks for effective, i
independent regulatory programs.
~j In September 1992, the NRC signed an Inter-Agency Agreement with the United States Agency for International Development to provide direct assistance to the Russian and Ukrainian regulatory bodies.
The United States DOE signed a similar agreement to provide assistance under the Lisbon Nuclear Safety Initiative for the 1
Newly Independent States.
Implementation began in October 1992
~
and will last for about three years.
In March 1993 at the annual JCCCNRS meeting in Kiev, Ukraine, the JCCCNRS. organizational structure was formally recognized to l
include the Lisbon Initiative.
This meeting was significant in that it marks the first joint meeting of the US-Russia and US-Ukraine JCCCNRS.
The meeting officially established two separate JCCCNRS's, meeting annually on a trilateral basis, the designation of Co-Chairmen from each country and the joint l
participation in cooperative and other selected activities.
In addition all parties agreed to extend the Memorandum of Cooperation in the Field of Civilian Nuclear Reactor Safety for another 5 years.
Currently, activities are in progress to ensure effective.and timely implementation of the President's Vancouver Summit Assistance to Russia.
[
Additional project funding ($5 million) became available with the signing of the FY-93 IAA in June 1993 to support continuation of the NRC activities.
Contact:
)
Jacob F. Wechselberger, Technical Program Manager for International Activities, Office of the Executive Director for 1
Operations, USNRC, Washington, DC 20555, (301) 504-1730 i
BP8 (09/93)
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
Background:
The NRC established the Systematic Assessment of Licensee Performance (SALP) program following the Three Mile Island 2 (TMI-2) accident.
At that time, the NRC relied on periodic NRC inspections to identify problems with plant performance.
With the SALP process, the NRC developed a routine, systematic approach for the review of all inspection and licensing activity.
The NRC uses the SALP process to synthesize its observations of and insights into a licensee's performance and to identify common themes or symptoms.
The primary product of the SALP process is the SALP report.
The SALP report documents NRC senior management's observations and conclusions regarding licensee performance.
The SALP process does not identify or propose resolutions of or solutions to problems.
The licensee is responsible for ensuring plant safety and establishing effective means to measure, monitor, and evaluate the quality of all aspects of plant operations.
The Commission recently approved a number of changes to the SALP program.
These changes include:
reducing the number of functional areas from seven to four, changing the SALP Board membership to Senior Executive Service (SES) members only, focusing on the assessment of the most significant issues in each functional area, emphasizing recent (within the last 6 months) licensee performance when determining the SALP category ratings, and reducing the report length to promote clearer communications with the licensee and the public.
Further, the staff will issue only a single, final, revision of the report.
The staff implemented these program changes for assessment periods ending j
after July 19, 1993.
Program Objectives.
The SALP program has four objectives.
The first objective is to i
conduct an integrated assessment of a licensee's safety
)
performance.
Senior NRC managers meet at the conclusion of the assessment period to assess licensee performance.
A review of the results of NRC inspections and other interactions with the licensee form the basis of the assessment.
The second objective of the SALP program is to provide for meaningful dialogue with the licensee regarding safety performance based on the insights gained from the synthesis of NRC observations.
The SALP process provides a basis for communications between NRC senior management and licensees regarding plant performance.
The third objective of the SALP program is to assist NRC management in making sound decisions regarding allocation of NRC resources used to oversee, inspect, and assess licensee performance.
Plants with superior performance are considered for reduced inspection effort while
I BP8 (09/93) plants with marginal performance will be considered for increased inspection effort.
The fourth objective of the SALP program is to provide a method for informing the public of the NRC's assessment of performance.
The SALP report is available in the local public document room and the public may attend the SALP meeting with the licensee.
SALP Program Requirements:
The SALP program now has four functional areas, instead of the previous seven, to assess licensee performance.
The four 7
functional areas are plant operations, maintenance, engineering, and plant support.
Plant support now includes the previous functional areas of radiation protection, emergency preparedness, and security in addition to chemistry, housekeeping and fire protection.
Safety Assessment / Quality Verification issues, which were previously considered within a separate functional area, are now considered in all functional areas and addressed in the cover letter, as appropriate.
Additional functional areas may be added if the NRC staff believes it is necessary.
The Board assigns a category rating for each of the above functional areas, with emphasis on the last six months of performance.
These category ratings are defined as follows:
Category 1 Licensee attention and involvement have been properly focused on safety and resulted in a superior level of safety performance.
Licensee programs and procedures i
provided effective controls.
The licensee's self-assessment efforts have been effective in the identification of emergent issues.
Corrective actions are technically sound, comprehensive, and thorough.
Recurring problems are eliminated and resolution of issues is timely.
Root cause analysis are thorough.
Cate gory 2 Licensee attention and involvement are normally well focused and resulted a good level of safety performance.
Licensee programs and procedures normally provide the j
necessary control of activities, but deficiencies may exist.
The licensee's self-assessments are normally good, although issues may escape identification.
Corrective actions are usually effective, i
although some may not be complete.
BP8 (09/93)
Root cause analysis are normally thorough.
-Category 3 Licensee attention and involvement have resulted in an acceptable level of safety performance.
However, the licensee's performance may exhibit one or more of the i
following characteristics.
Licensee's programs and procedures l
have not provided sufficient control of activities in important areas.
The licensee's self-assessment efforts may not occur until after a potential problem becomes apparent.
A clear understanding of the safety implications of significant issues may not have been demonstrated.
Numerous minor issues combine to j
indicate that the licensee's corrective action is not thorough.
Root cause analyses do not probe deep 1enough, resulting in the incomplete resolution of issues.
i Because the margin to unacceptable performance in important aspects is small, increased NRC and licensee i
attention is required.
t It should be noted that the lowest SALP rating, Category 3, represents acceptable performance although the margin to
]
unacceptable performance may be small.
The NRC does not rely upon the SALP program to identify unacceptable performance.
SALP Process:
i i
The SALP process starts with the assignment of an assessment period.
The normal length of a SALP assessment period is 18 months (plus or minus two months).
In addition, the NRC regional offices can adjust the length of the SALP period based on plant performance, with shorter assessment periods for plants that need more frequent monitoring.
At the conclusion of the SALP assessment period, a SALP Board convenes in the NRC regional office.
The SALP Board membership is comprised of four NRC regional and headquarters managers.
NRC regional and headquarter's staff responsible for inspection and review activities at the facility make presentations to the Board.
The Board assesses performance and assigns a category rating to each of the four functional areas.
A SALP report is prepared to communicate the results of the assessment.
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i BP8 (09/93)
The Regional Administrator approves the SALP report and transmits it with a cover letter to the licensee (and the public document room) as the SALP report.
The licensee is given the opportunity to review the report.
The NRC and the licensee meet during a special public SALP meeting to discuss the results of the SALP assessment.
This SALP management meeting is normally held near the facility.
SALP Results:
The SALP results are utilized by the NRC in the inspection planning process.
Those facilities receiving low SALP scores l
(Category 3) are considered for increased inspection activity, and those facilities with high SALP scores (Category 1) are considered for reduced inspection activity.
Contact:
Gary G.
Zech, Chief, Performance and Quality Evaluation Branch US NRC, Washington, D.C.
20555 (301) 504-1017.
BP8 (09/93)
HIGHLIGHTS OF THE i
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE PROGRAM o
The NRC developed the Systematic Assessment of Licensee Performance (SALP) program following the TMI-2 accident to systematically evaluate observations of and insights into a licensee's performance and to identify common themes or symptoms.
o The SALP program has four objectives: (1) to conduct an integrated assessment of licensee safety performance that focuses on the safety significance of the NRC findings and conclusions during the assessment period, (2) to provide a vehicle for meaningful dialogue with the licensee regarding its safety. performance based on the insights gained from synthesis of NRC observations, (3) to assist NRC management in making sound decisions regarding allocation of NRC resources used to oversee, inspect, and assess licensee performance, and (4) to provide a method for informing the public of the NRC's assessment of licensee performance.
o A SALP Board, comprised of four NRC Senior Executive Service managers, convenes to assess licensee performance.
o The SALP Board recommends to the regional administrator numerical rating to the licensee's performance in each of the four functional areas with emphasis on the last six months of performance.
A Category 1 represents superior performance, Category 2 a good level of performance, and Category 3, acceptable performance.
o The regional administrator assigns the category ratings and issues the SALP report.
The NRC and licensee conduct a public meeting to discuss the o
results of the SALP report.
o Superior performance (Category 1) can result in longer SALP assessment periods and reduced inspection effort for the licensee, and acceptable performance (Category 3) can result in shorter SALP assessment periods and increased inspection effort by the NRC.
BP9 (09/93)
CHERNOBYL STATUS l
Background
- l On April 26, 1986, a major accident, determined to have been a reactivity accident, occurred at Unit 4 of the nuclear power station at Chernobyl, Ukraine, in the former USSR.
The accident destroyed the reactor and released massive amounts of radioactivity into the environment.
After the accident the area in an 18-mile radius around the plant was closed to residential and other access, except for persons requiring official access to tne plant and to the immediate area for evaluating and dealing I
with the consequences of the accident and operation of the undamaged units.
The evacuated population numbered approximately 1
135,000.
Thirty-one people died in the accident and its immediate aftermath, most in fighting the fires that ensued; delayed health effects could be extensive, but there are no generally agreed estimates.
To stop the fire and prevent a j
further criticality accident as well as substantial further release of radioactive fission products, boron and sand were dumped on the reactor from the air.
The three other units of the four-unit Chernobyl nuclear power station were subsequently restarted.
The damaged unit (No. 4) was entombed in a concrete
" sarcophagus," to limit further release of radioactive material.
control measures to reduce the radioactive contamination at and i
near the plant site included cutting down and burying, with the top yard of soil, a pine forest of approximately 1 square mile.
The Soviet nuclear power authorities presented a report on the accident at an International Atomic Energy Agency meeting in Vienna, Austria, on August 25-29, 1986.
i The Chernobyl reactors are of the RBMK type.
These are high-power, pressure-tube reactors, moderated with graphite and cooled with water.
Fifteen RBMKs are still in operation in the former USSR.
The operating RBMK units incorporate safety improvements made since the Chernobyl accident, though important vulnerabilities remain.
U.S. reactors differ in significant ways i
from the RBMKs.
Status at and Around Chernobyl:
There is concern about the long-term safety of the sarcophagus and long-term dependability of the on-site burial set-up for the buried forest and other massive contaminated materials.
Among the sarcophagus concerns is that the faster-than-expected deterioration of structures could lead to rearrangement of materials inside, which could cause an additional release of radioactive material, mainly as dust.
The Ukrainian authorities are studying options for dealing with the sarcophagus problem and have been seeking ideas and help from the international community.
An international competition in late 1992 - early l
BP9 (09/93) 1993 produced a number of proposals, mainly from European countries, but there has so far been no contract award.-
Public concern in the areas near Chernobyl,Ein Ukraine, about the safety of the still operating units has led to a Ukrainian government decision to shut them down permanently in 1993.
It is uncertain at this time whether that decision will be carried out.
Studies of longer-term health and environmental consequences of the accident are continuing.
In 1989-90 decisions were taken to evacuate substantial additional nearby areas, in Ukraine and Belarus (and a small area in Russia).
About 200,000 people have been relocated so far (including the originally evacuated 135,000).
It is not clear at this time to what extent additional relocations will take place.
In an event in October 1991, unrelated to the Chernobyl Unit 4 accident, Unit 2 suffered a fire.
There was no significant radioactive material release, but the plant damage was severe.
Unit 2 is now shut down; restart is not planned.
International Studies:
Soviet and successor-state studies of the condition of the sarcophagus, contamination of land and water bodies, health effects, and other post-accident issues continue with international contributions.
In the fall of 1989, the First International Workshop on Severe Accidents and Their Consequences, devoted to the Chernobyl accident, was held in Sochi, USSR, under joint sponsorship of the Soviet Nuclear Society and the American Nuclear Society.
In mid-1991 an international advisory committee sponsored by the International Atomic Energy Agency completed The International Chernobyl Project: Assessment of Radiological Consequences and Evaluation of Protective Measures.
A fact-finding team on the state of the sarcophagus, operating under the auspices of the Organisation for Economic Co-operation and Development's Nuclear Energy Agency, issued its report in October 1992.
Status of USNRC Follow-up:
The U.S. Nuclear Regulatory Commission's review of the Chernobyl accident was divided into three major phases: determining the facts of the accident, assessing the implications of the accident for safety regulation of commercial nuclear power plants in the United States, and conducting specific further studies suggested by that assessment.
The first phase, fact finding, was a coordinated effort among several U.S. Government agencies and some private groups, with the NRC acting as the coordinating agency.
The work was
BP9 (09/93) essentially completed in January 1987 and updated later that year.
The results are reported in NUREG-1250, " Report on the Accident at the Chernobyl Nuclear Power Station."
The second phase, the implications study, was reported in NUREG-1251, " Implications of the Accident at Chernobyl for Safety Regulation of Commercial Nuclear Power Plants in the United States," issued after public comment in April 1989.
The assessment led to the conclusion that no immediate changes were needed in the NRC's regulations regarding the design or operation of U.S.
commercial nuclear reactors.
Plant design, shutdown margin, containment, and operational controls at U.S. reactors protect them against a combination of lapses such as those experienced at Chernobyl.
Although the NRC has always acknowledged the possibility of major accidents, its regulatory requirements provide adequate protection, subject to continuing vigilance.
The NRC reviews new information that may suggest weaknesses.
Assessments in the light of Chernobyl have indicated that the causes of the accident have been adequately dealt with in the design of U.S.
commercial reactors.
Yet, the assessment went on to conclude, the Chernobyl accident has lessons for us.
The most important lesson is that it reminds us of the continuing importance of safe design in both concept and implementation, of operational controls, of competence and motivation of plant management and operating staff to operate in strict compliance with controls, and of backup features of defense in depth against potential accidents.
Although a large nuclear power plant accident somewhere in the United States is unlikely because of design and operational features, we cannot relax the care and vigilance that have made it so.
Accordingly, the assessment led to the recommendation that certain issues should receive further consideration, to provide a basis for confirming or changing existing regulations or staff guidance.
Those issues include reactivity accidents, accidents at low power or at zero power (when the reactor is shut down), operator training, and emergency planning.
The Chernobyl follow-up studies for U.S. reactors are the third phase of the NRC review.
An overview report on this work, NUREG-1422, " Summary of Chernobyl Followup Research Activities,"
was issued in June 1992.
That report closes out the Chernobyl 1
follow-up research program as such, though certain issues will receive continuing attention in the normal course of NRC work.
For example, the long-term lessons with regard to contamination control -- decontamination, ingestion pathway, relocation of people -- will continue to be followed.
Beyond the recommended specific studies, the NRC assessment i
recognizes that the Chernobyl experience should remain as part of 4
i BP9 (09/93)
-the information to be taken into account when dealing with reactor safety issueE in the future.
CONTACT:
George Sege, Technical Assistant to the Director, Office of Nuclear Regulatory Research, U.S.N.R.C., Washington, DC (301) 492-3904.
BP9 (09/93) t CHERNOBYL STATUS Highlights The accident that destroyed Unit 4 of the Chernobyl Nuclear e
Power Station in the USSR took place on April 26, 1986.
31 people died in the accident and its immediate aftermath.
e Longer-term health effects are being studied.
e 135,000 people were evacuated, from an 18-mile radius exclusion zone around the reactor shortly after the accident.
Additional areas are being evacuated.
200,000 people (including the original 135,000) have been relocated to date.
The damaged Unit 4 reactor is entombed in a " sarcophagus,"
e to prevent further contamination spread.
Scientists in Ukraine and other countries are studying the sarcophagus and considering options for dealing with concerns over its long-term stability.
e The NRC, with other U.S. Government agencies and some private groups, conducted a study to determine the facts of i
the accident.
The results were published in 1987.
(Report No. NUREG-1250).
The NRC assessed the implications of the Chernobyl accident e
for safety regulation of commercial nuclear power plants in the U.S.
The results were reported in 1989.
(Report No.
The assessment concluded that no immediate changes were needed in NRC's regulations, because the causes of the accident have been largely anticipated and i
accommodated in U.S.
- designs, e
The assessment (NUREG-1251) led to recommendation of certain further studies to provide a basis for confirming or changing regulations.
A report on these studies (NUREG-1422) was issued in June 1992.
t
BP10 (09/93)
THREE MILE ISLAND UNIT 2 STATUS Backcround:
General Public Utilities Nuclear Corporation (GPUN or the licensee) is in the final phase of the current cleanup effort at Three Mile Island, Unit 2 (TMI-2).
Since the March 28, 1979, accident, the licensee han conducted a comprehensive cleanup program designed to place the facility in a safe and stable configuration.
Following mitigation of the accident and stabilization of the facility, the major licensee efforts over the past 14 years have included partial facility decontamination, removal of fuel from the reactor vessel and other facilities, offsite shipment of substantial quantities of both high and low level radioactive wastes and the removal, treatment and partial disposal of the accident-generated water.
The licensee has proposed placing the facility into long term storage until Three Mile Island, Unit 1 (TMI-1), also located on the same site as TMI-2, permanently ceases operation, at which time both facilities would be decommissioned.
The long term storage period for TMI-2 is called Post-Defueling Monitored Storage (PDMS) by the licensee.
Recent Accomplishments:
During July and August of 1991, the reactor vessel was drained to make final measurements of the residual fuel remaining in the vessel.
Less than 1390 pounds (609 kilograms) of fuel had been previously estimated by visual observations to remain in the reactor vessel.
The final vessel fuel measurement program is the last phase in a special nuclear materials accountability program at TMI-2.
The measurement technique made use of an array of helium filled detectors to measure fast neutrons produced by the residual fuel.
On February 1, 1993, GPUN notified the NRC staff that their current best estimate of the residual fuel in the reactor vessel is 2040 pounds (925 kilograms) based on the data from the fast neutron measurements.
This estimate was derived from calculations made by onsite staff and an independent review by an I
offsite group headed by Dr. Norman Rasmussen of the Massachusetts Institute of Technology.
This estimate was reviewed and confirmed by three additional independent reviewers from national laboratories.
For the balance of the facility external to the reactor vessel, earlier licensee estimates based on measurements, sample analyses, and visual observations indicated that no more than 385 pounds (174.6 kilograms) of residual fuel remains.
The NRC staff
)
and their consultants from Battelle Pacific Northwest Laboratories have performed independent evaluations and made
i BP10 (09/93) independent measurements of these earlier fuel measurements in the auxiliary and reactor buildings.
On July 6, 1993, the staff issued an analysis which concluded that the fuel remaining in the TMI-2 reactor vessel will remain subcritical, with an adequate margin of safety, during PDMS.
Evaporation of the treated accident-generated water began in January 1991 after a prolonged period of system testing, modification, and repair.
On August 12, 1993, the decontamination and evaporation of 2.23 million gallons of accident-generated water was completed.
Post Defuelina Monitored Storace (PDMS):
In August of 1988, the licensee submitted a Safety Analysis Report (SAR) to document and support their proposal to amend the TMI-2 license to a possession only license and to allow the facility to enter PDMS.
The staff issued final supplement 3 to the Programmatic Environmental Impact Statement for the TMI-2 decontamination and cleanup in August of 1989.
In February 1992, the staff issued a Safety Evaluation regarding the PDMS license amendment and a Technical Evaluation Report regarding PDMS.
These three NRC staff documents form the basis for the staff position on the acceptability of PDMS.
On April 25, 1991, the 1
staff published a notice of opportunity for a prior hearing regarding the licensee request to amend its license.
A member of the public petitioned to intervene in the license amendment proceedings.
The petitioner, the licensee, and the NRC staff reached a settlement agreement on September 25, 1992.
The request to intervene was withdrawn and on October 16, 1992, the Atomic Safety and Licensing Board dismissed the proceeding.
The licensee is currently preparing the facility for placement into the PDMS condition.
The reactor building preparations for PDMS were completed in October 1992 and it is now in a pre-PDMS condition.
The NRC staff and the licensee have prepared a punchlist of remaining hardware and software items to be completed prior to entry into PDMS.
The NRC staff expects to issue a possession only license in the third quarter of 1993 and that TMI-2 will enter PDMS in the fourth quarter of 1993.
Decommissionina Fundina:
On July 26, 1990, the licensee submitted to the NRC their Decommissioning Funding Plan.
The licensee stated in the plan that it will deposit $195 million (1989 dollars) in an escrow account for the radiological decontamination of TMI-2.
The value in 1992 dollars is $229 million.
The $229 million is in excess of the amount required by NRC regulations for a facility the size of TMI-2.
The licensee is required by regulations to submit a preliminary decommissioning plan containing site-specific s
BP10 (09/93) decommissioning cost estimates five years before the expiration date of the TMI-2 license.
Should the preliminary decommissioning plan identify that there are insufficient funds available for decommissioning, the licensee has the remaining five year period to adjust the rate of contributions to the fund to make up any shortfall.
Contacts:
Michael T. Masnik, DORS /NRR (301) 504-1191 Lee H. Thonus, DORS /NRR (717) 948-1164 I
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BP10 (09/93)
HIGHLIGHTS OF THREE MILE ISLAND UNIT 2 STATUS Accident occurred at TMI-2 on March 28, 1979.
e Began processing of contaminated water through EPICOR in e
October 1979.
Vented 43,000 curies of krypton from the reactor building in July 1980.
Performed the first manned entry into the reactor e
building in July 1980.
Obtained first television pictures of the damaged reactor core in July 1982.
Removed reactor vessel head and service structure in July 1984.
Removed the reactor vessel plenum and installed work e
platform in May 1985.
Defueling began in October 1985.
Began core debris offsite shipments in July 1986.
1 Completed defueling in April 1990.
e Began evaporation of the accident generated water in January 1991.
Cost of the cleanup through the end of defueling approximately $980 million.
Licensee plans to escrow $229 million (1992 dollars) for e
radiological decommissioning.
Total occupational exposure 6579 person rem as of the e
end of calendar year 1992.
Annual exposure less than 0.1 millirem to the maximally exposed member of the public during defueling.
Accident-generated water processing completed on August e
12, 1993 - total of 2.23 million gallons.
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LICENSEE PERFORMANCE INDICATORS
Background:
As a part of its mission to protect the public health and safety, the environment, and national security, the Nuclear Regulatory Commission (NRC) monitors the performance of licensees who operate the 109 commercial nuclear power plants currently in operation in the United States.
The findings of this monitoring effort are used by the NRC to adjust its plant-specific regulatory programs.
The NRC staff has developed several tools as inputs to the monitoring function.
One of these tools is a set of performance indicators.
Performance indicators are intended to provide ready information concerning nuclear power plant performance trends and to assist NRC management in identifying poor and/or declining safety performance as well as good and/or improving safety performance.
Performance Indicators:
The Performance Indicator (PI) Program is one aspect of the Commission's efforts to monitor the performance of licensees who operate commercial nuclear power plants in the United States.
The program was started in May of 1986, when an interoffice task group began development of the overall NRC program.
The first PI Report was published in February of 1987 and contained data on six indicators from the first quarter of 1985 through the fourth quarter of 1986.
Reports were provided quarterly to NRC management until June of 1993, when the frequency was changed to twice a year.
Under the direction of the Office for Analysis and Evaluation of Operational Data (AEOD), the PI program has been improved and I
expanded since it was first introduced.
The program currently monitors industry-wide data on eight PIs and evaluates the data to determine performance trends. The eight PIs are:
(1) the I
number of unplanned automatic reactor scrams (trips) while a reactor is critical, (2) the number of selected safety system actuations, (3) the number of significant events, (4) the number of safety system failures, (5) the forced outage rate, (6) the number of equipment-forced outages per 1000 commercial critical hours, (7) the collective radiation exposure, and (8) cause codes.
The AEOD staff provides biannual reports containing i
plant-specific data for these eight PIs to the Commission and to NRC senior managers.
The reports are also placed in the NRC Public Document Room.
In addition, the staff provides plant-specific information and industry average data extracted from each PI report to licensee managers.
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The PI Program is a single, coordinated, overall NRC program that provides.an additional view of operational performance and enhances the NRC's ability to recognize areas of changing safety performance of operating plants. However, it is only a tool that must be used in conjunction with other tools, such as the-results of routine and special inspections and the Systematic Assessment of Licensee Performance (SALP), to provide data to NRC managers who must decide whether any plant-specific regulatory programs need adjusting. It should be recognized that PIs have limitations and are subject to misinterpretation.
Performance indicators alone do not provide a basis for ranking individual power plants or taking regulatory actions and are not used in communications with licensees as a measure of performance level.
The NRC has a sensitivity to pressure which would cause licensee personnel at individual power plants to " manage the indicators" or take any actions that are contrary to plant safety because of performance indicators.
The PIs for a given plant, when viewed as a set, provide additional data for assessing changes in plant i
operational performance. The PIs focus attention on the need to assess and understand underlying causes of identified changes by evaluating other available information.
[
The NRC will continue its review, evaluation, and revision, as needed, of the PI Program.
Development and implementation of risk-based indicators is on-going and will continue.
Purther program revisions will be made as appropriate.
Contact:
Donald E.
Hickman, Chief, Performance Indicator Section, US NRC, Washington, DC 20555.
Phone: (301) 492-4431 1
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1 BP11 (09/93)
HIGHLIGHTS OF LICENSEE PERFORMANCE INDICATORS l
o NRC monitors licensee performance to adjust plant-specific i
regulatory programs o
One monitoring tool is a set of performance indicators (PIs) to identify performance trends o
Quarterly reports provided from 1987 to June 1993, twice a year after that o
Eight indicators evaluated to determine performance trends:
)
(1) automatic scrams while critical, (2) selected safety system actuations, (3) significant events, (4) safety system failures, (5) forced outage rate, (6) equipment forced outages per 1000 commercial critical hours, (7) collective radiation exposure, and (8) cause codes o
Provided to the Commission, NRC senior managers, licensee senior managers, and the public o
PIs have limitations, are subject to misinterpretation, and must be used with other tools to assess regulatory programs o
Not to be used alone for ranking plants, taking regulatory actions, or as a measure of performance level o
Used as a set to focus attention on the need to assess and understand underlying causes o
Review, evaluation, and revision of the PI Program, including risk-based indicators, is on-going I
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NUCLEAR REACTOR RISK TO THE PUBLIC (ACCIDENTS)
Background:
The safe operation of a nuclear power plant requires the consideration of various types of possible accidents which may pose a risk to the public.
The accidents range in terms of severity and likelihood of occurrence.
Risk is considered to be the combination of accident severity (consequences) and the likelihood (frequency) of the accident.
The risk due to operation of nuclear power plants can never be zero, just as the risks we face from other sources such as illness and auto accidents can never be zero.
Regulatory requirements and attention are necessary to assure that the risk from nuclear power plant operation is very low when compared to all other types of risk that we face every day.
The Nuclear Regulatory Commission's (NRC's) responsibility is centered around the application and enforcement of the applicable regulatory requirements as described in Title 10 of the Code of Federal Regulations.
The intent is to assure that the risks are maintained at acceptably low levels.
Nuclear Reactor Accident Risks:
Nuclear power plants are designed to confine the fission products which accumulate within the nuclear fuel.
Part of the overall risk stems from the possibility of accidental release of the fission products into the environment beyond the plant site boundaries.
To reduce this risk to acceptable levels, the concept of " defense-in-depth" is applied to the design, licensing, and operation of nuclear power facilities.
For example, the physical confinement of fission products is implemented by way of multiple barriers such as fuel cladding, reactor coolant system vessel and piping, and a containment building.
The need to maintain the integrity of the reactor core (fuel) and avoid damage requires that an adequate supply of water be provided for cooling it.
Here again, " defense-in-depth" is implemented by providing diverse and multiple backup systems so that there is an adequate assurance of a supply of water for cooling the core.
Due consideration is given to keeping the plant within safe operating limits and conditions (technical specifications).
Other safety measures include paying attention to the availability and reliability of plant equipment, plant i
maintenance, operator training, and plant management in order to minimize the overall risk.
Regulatory reviews, analyses and inspections are used to assure that these measures comply with appropriate NRC regulations and that the estimated risk is acceptably low.
l BP12 (09/93) i Regulations
- Title 10 of the Federal Code of Regulations (10 CFR) contains the NRC's criteria and requirements for assuring an acceptable level of safety with respect to nuclear power plants.
As an adjunct to the regulations, the NRC has developed a set of Regulatory Guides and the Standard Review Plan in order to clarify the regulatory requirements.
The NRC also issues various generic communications that address safety related concerns.
None of the regulations relate directly to quantitative risk measures.
The regulations are viewed in terms of the use of sound engineering concepts to provide what is judged to be an acceptable level of safety.
However, the NRC has issued a policy statement that describes safety goals for the operation of nuclear power plants.
The qualitative safety goals are that nuclear power should pose no significant additional risk to the life and health of individuals and that societal risk should be comparable to other viable competing energy supply technologies with no significant addition to other societal risks.
The quantitative safety goals are that the risk of an individual prompt fatality should not exceed 0.1% of that due to other accidents and that the risk of cancer fatalities to nearby population should not exceed 0.1% of the total cancer risk from all other causes.
Current Status:
Currently, 109 commercial nuclear power plants are licensed to operate in the U.S.
The NRC believes that by meeting the existing regulations these plants pose an acceptably low level of risk to the public.
However, in 1988, the NRC issued a requirement for the utilities licensed to operate nuclear power plants to perform a plant-specific search for vulnerabilities to severe accidents.
In this effort, known as the Individual Plant Examination (IPE) program, virtually all licensees have chosen to use probabilistic risk assessment methods.
This program is directed towards the identification of possible plant weaknesses with respect to safety that stem from considerations beyond those covered by the existing regulations.
The intent is to reduce the risk even further.
The first phase of this program is devoted to investigating accidents which could be initiated within nuclear plants and is expected to take about three years to complete.
Sixty-one IPEs have been submitted to the NRC to date.
Weaknesses identified in this process are being addressed by licensees and reviewed by the staff.
The second phase will address accidents which could be initiated externally, such as earthquakes and floods.
The second phase was started in 1990 and is expected to be completed in 1995.
BP12 (09/93)
In order to improve the means for evaluating risks associated with the operation of nuclear power plants, the NRC is continuing to improve and refine probabilistic risk assessment (PRA) methods.
At the same time, the NRC is monitoring plant operations, equipment failures, operator errors and similar plant performance features in order to identify potential problems before they become serious.
Contact:
Mark P. Rubin, Acting Chief, Probabilistic Safety Assessment Branch US NRC, Washington, D.C.
20555 (301) 504-1089 i
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BP12 (09/93)
HIGHLIGHTS OF NUCLEAR REACTOR RISK TO THE PUBLIC o
As a result of the corrective steps taken by the NRC and the nuclear power industry since the Three Mile Island accident, the estimated probability of a nuclear reactor accident has decreased significantly.
o The NRC has issued a policy statement describing the safety goals with respect to nuclear power plants.
1 o
The risks associated with the operation of nuclear power plants that meet existing safety regulations are acceptably low, o
The NRC has implemented the' Individual Plant Examination program in order to identify any weaknesses to events beyond the design basis of the plant that may exist and to reduce the risks even further.
BP13 (09/93)
ADVANCED REACTORS PROGRAM The NRC's " Statement of Policy for Regulation of Advanced Nuclear Power Plants (Final Statement)", July 8, 1986, encourages early interaction (prior to license application) between NRC and advanced reactor designers to provide licensing guidance applicable to those designs.
In June 1988 the NRC issued NUREG-1226, a report on " Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants".
The report provides guidance on the implementation of the policy, and describes the staff approach to be used in the review of advanced reactor concepts under the policy statement.
Other important Commission guidance to be used in the early interaction process with sponsors of advanced designs include the Severe Accident Policy Statement (50 FR 32138; August 8, 1985), the Standardization Policy Statement (52 FR 34884; September 15, 1987), and the Safety Goal Policy Statement (51 FR 30028; August 21, 1986).
Thus the staff is prepared to conduct early, interactive reviews of advanced reactor designs including consideration of the Commission's regulations, regulatory guides and other guidelines.
These include such established and developing criteria as the defense-in-depth philosophy, standardization, the Commission's safety goal and severe accident policies, and applicable industry codes and standards.
NRC carly interaction with a potential applicant is in the context of a preapplication review which takes about two years and culminates in a preapplication safety evaluation report.
The objectives of the preapplication review are (1) to identify major safety issues that could require commission policy guidance to the staff, (2) to identify major technical issues that the staff could resolve in the context of existing regulations or Commission policy, and (3) to identify research and development needed to resolve identified issues.
The early reviews discussed here are done prior to formal submittal of an application for a standard design certification pursuant to 10 CFR Part 52 (54 FR 15372; April 18, 1989).
This regulation was enacted by the Commission to provide a stable procedural framework for the early resolution of licensing issues and finally, certification of a standardized design by rulemaking.
CANDU 3 The CANDU 3 design is a single-loop pressurized heavy water
)
reactor rated at 450 MWe with two steam generators and two heat transport pumps connected in series.
The design utilizes natural uranium fuel, separate heavy water moderator and reactor coolant, computer-controlled operation, and on-line refueling.
The reactor has 232 horizontal pressure tubes supported in a
BP13 (09/93) calandria tank filled with the heavy water moderator.
The tank also supports the reactivity regulating and safety devices which are inserted between and among the pressure tubes.
The CANDU 3 design is an evolution of the CANDU 6 design (600 - 640 MWe).
The CANDU 6 has been approved by the Atomic Energy Control Board (AECB), the Canadian government agency responsible for regulating atomic energy in Canada, for operastion in Canada.
Four CANDU 6 l
units are currently in operation (two in Canada, one each in Korea and Argentina) and 28 other CANDU-type units have been built around the world, but none in the U.S..
The CANDU 3 contains many features and components already in use in the CANDU 6 design.
The CANDU 3 design is being developed by the Atomic Energy of Canada, Limited (AECL), whose design facilities are based in Mississauga, Ontario, and Saskatoon, Saskatchewan, Canada.
AECL Technologies, a United States subsidiary of AECL, Incorporated, informed the NRC of its intent to seek design certification of the CANDU 3 design under the provisions of 10 CFR Part 52 in a i
letter to then Chairman Lando Zech, dated May 25, 1989.
AECL Technologies subsequently submitted, and the NRC is reviewing, a two volume Technical Description and a two volume Conceptual Safety Report describing the CANDU 3.
In addition, AECL Technologies has submitted four technology transfer reports describing various aspects of the CANDU technology.
The NRC is continuing its activities to prepare for a design certification application by becoming familiar with the design, maintaining technical progress on key issues, and work on code development and benchmarking.
In a letter dated August 3,
- 1993, AECL Technologies stated their intent to file a Standard Design certification application within about one year.
The staff is identifying the resources necessary for carrying out a standard design certification review of the CANDU 3 design.
PIUS PIUS is a 640 MWe advanced pressurized water reactor (PWR) design, by ABB Atom of Sweden, that utilizes natural physical phenomena to accomplish control and safety functions usually performed by electromechanical devices.
The PIUS design consists of a vertical pipe, called a reactor module, which contains the reactor core and is submerged in a large pool of highly borated water.
The reactor cure is comprised of fuel elements that are similar to current PWR elements.
The borated pool water is provided to shutdown the reactor and to cool the core by natural circulation while the reactor is shutdown.
The reactor module is open to the borzited pool at the bottom and again at the top of the reactor module.
Devices referred to as density locks are provided at these two openings.
They consisting of a number of open tubes.
During normal operation at design power, the density locks prevent the entry of the highly borated pool water into the
BP13 (09/93) reactor module due to the pressure balance maintained across the tube bundle. In normal operation, the primary loop reactor water
)
enters the reactor module from the steam generator, flows up i
through the core, out of the top of the reactor medule to the steam generator, and is pumped back into the bottom of the reactor module, bypassing both the top and bottom density locks.
Under certain transient conditions, the pressure balance across the density locks is not maintained and the borated pool water flows into the core and shuts down the reactor.
A natural circulation flow path is then established from the borated pool through the lower density lock, up through the core, and back into the borated pool through the upper density lock for long term shutdown cooling.
Unlike most reactors, PIUS does not use control rods for regulating reactivity.
Reactivity is controlled by the boron concentration and temperature of the primary loop reactor water.
The steam generating equipment of the PIUS design is similar to that of a typical U.S.
or European pressurized light water reactor plant.
One important difference in plant design is the very large, by current standards, prestressed concrete reactor vessel.
This vessel holds both the reactor module and the borated pool.
In October 1989, ABB Atom requested the NRC to perform a licensability review of their Process Inherent Ultimate Safety (PIUS) Preliminary Safety Information Document (PSID).
- However, since then, the staff and ABB-CE have agreed that further work on the preapplication review of the PIUS PSID would not be meaningful given the limited resources available.
The staff is in the process of documenting the review of the PIUS design by March 1994 and will terminate all other activities until an application for design is submitted by ABB-CE.
Presently there is no date for a design certification application.
MHTGR The modular high temperature gas-cooled reactor (MHTGR) design is a helium-cooled and graphite-moderated thermal power reactor.
The fuel is millions of ceramic coated microspheres distributed in cylindrical graphite-like rods which are inserted in large hexagonal graphite blocks.
The blocks are stacked vertically within the reactor vessel through which the pressurized helium coolant is circulated.
The plant design consists of four identical reactor modules, each with a thermal output of 350 MWt, which are coupled with two steam turbine-generator sets to produce a total plant electrical output of 540 MWe, with a power conversion of approximately 40%.
This is about half the power output of current single light water reactor units.
The design includes passive reactor shutdown and decay heat removal features to minimize required reactor operator actions.
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BP13 (09/93)
The NRC has licensed one high temperature gas reactor, Fort St.
Vrain in Colorado, which was permanently shutdown in August 1989.
j The advanced MHTGR design, sponsored by the Department of Energy (DOE), has been under review at the NRC since 1986, and a draft
{
Preapplication Safety Evaluation Report (PSER) was issued in March 1989 (NUREG-1338).
Important safety matters now being focused on by the staff are fuel design and performance, containment design and performance, reactor cavity cooling system, accident selection and analysis, accident source terms and analysis, role of the operators, design of the control room and remote shutdown area, emergency preparedness, and quality standards for equipment.
The current review effort to update the draft report and issue a final PSER is tentatively scheduled for completion by the Spring of 1995.
PRISM The design submitted by the Department of Energy (DOE) is a small, modular, pool-type, liquid-sodium cooled reactor producing 471 MWt.
The reactor fuel elements are cylindrical tubes containing pellets of uranium-plutonium-zirconium metal alloy.
The reactor size was selected to permit use of passive shutdown and decay heat removal features.
1 The PRISM standard plant is to consist of nine reactor modules arranged in power blocks of three reactor modules with their steam generators supplying steam to one turbine-generator.
The power output of the standard site would be 1395 MWe.
Each of the reactor modules would be located in a silo below grade.
The steam generator and secondary system hardware would be located in a separate building and would be connected by a below-grade pipeway.
The reactor modules would share a common control center, nuclear island maintenance building, and reactor i
service building.
In general, the PRISM design features have been chosen to prevent core damage events that previova Liquid Metal Reactor (LMR)
I designs have traditionally boon designed to accommodate.
1 DOE submitted the conceptual design to NRC for preapplication review in November 1986.
The NRC published a draft PSER in September 1989 (NUREG-1368).
The staff's preapplication review is to provide guidance to the design's sponsors early in the design process.
The current review effort to update the draft PSER and issue a final PSER is tentatively scheduled for completion by the end of 1993.
1
Contact:
1 Edward D. Thron, Acting Project Director, Advanced Reactors Project Directorate, US NRC, Washington, DC 20555, (301) 504-1111 4
BP13 (09/93)
HIGHLIGHTS - ALL PROJECTS CANDU 3 --
AECL Technologies, a division of AECL, Incorporated, e
submitted a letter in May 1989 stating intent to seek design certification of the CANDU 3 pursuant to 10 CFR Part 52.
A number of documents were submitted to NRC for preapplication review between mid-1990 and early 1991.
Several meetings have been held among the NRC, AECB, AECL, and AECL Technologies to define the scope of the preapplication review and to discuss preliminary technical 1'
issues identified by the NRC.
In an August 3, 1993, letter, AECL Technologies stated their e
intent to submit the CANDU 3 design for standard design certification in about one year.
The staff preparation for design certification review is ongoing.
PIUS --
ABB requested a licensability review of the PIUS design in e
October 1989.
ABB submitted the PIUS Preliminary Safety Information e
Document (PSID) for review in May 1990.
First review of PIUS began in June 1991.
i e
Status of the review to be documented by March 1994.
All e
other activities stopped until a design certification application is received.
MHTGR --
i Preliminary Safety Information Document (PSID) submitted by e
DOE in 1986.
Draft Preapplication Safety Evaluation Report issued by NRC e
in March 1989.
DOE submitted PSID Amendments 11 to 13 responding to issues, December 1991, and March and August, 1992.
NRC tentative schedule for final PSER for the MHTGR design is Spring 1995.
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BP13 (09/93)
PRISM --
Preliminary Safety Information Document (PSID) submitted to o
NRC in November 1986.
[
NRC Draft Preapplication Safety Evalustion Report (PSER) on PRISM (NUREG-1368) released September 1989.
DOE submitted Amendments 12 and 13 responding to issues, i
March and May 1990.
NRC tentative schedule for final PRISM Preapplication Safety-e Evaluation Report (PSER) is December 1993.
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HIGH-LEVEL RADIOACTIVE WASTE BRIEFING UPDATE Backcround:
High-level radioactive waste (HLW) means:
(1) irradiated (spent) reactor fuel, (2) liquid wastes resulting from the operation of the first cycle solvent extraction system, and the concentrated wastes from subsequent extraction cycles, in a facility for reprocessing irradiated reactor fuel, and (1) solids into which such liquid wastes have been converted.
HLU is primarily in the form of spent fuel discharged from commercial nuclear power plants; it also includes some reprocessed HLW from defense activities, and a small quantity of reproce:ssed commercial HLW.
Current plans for management of HLW call for the development of a monitored retrieval storage (MRS) facility by 1998, and a j
permanent HLW repository deep beneath the surface of the earth by the year 2010.
The U.S. Department of Enttrgy (DOE) has the responsibility for disposing of HLW.
The U.S.
Environmental Protection Agency (EPA) is responsible for developing appropriate environmental standards for HLW.
The, Nuclear Regulatory Commission has the licensing authority f.or the disposal and long term storage of HLW.
i Hich-Level Radioactive Waste:
)
This country's policies governing thea permanent disposal of HLW are defined by the Nuclear Waste Policy Act of 1982 (NWPA) and i
the Nuclear Waste Policy Amendments Act (NWPAA) of 1987.
To I
provide the long-term permanent isolation required, the NWPA
)
specifies that HLW will be placed in deep-underground geologic repositories to be built and operated by DOE.
To this end, DOE is developing a waste management system consisting, in part, of a I
geologic repository in which HLW can be permanently isolated deep beneath the surface of the earth, and an MRS in which vaste can be stored prior to permanent disposal.
NRC has the licensing and related regulatory authority for both the MRS and high-level waste geologic repository.
l An MRS facility is an integral part of the waste management system being proposed by DOE for achieving timely acceptance of spent fuel.
NWPAA allows a dual approach to MRS siting:
(1) siting by DOE, through a process of survey and evaluation; and (2) siting through the efforts of the Nuclear Waste Negotiator.
Thrcugh the NWPAA, Congress designated the Yucca Mountain site in Nevada as the single candidate site for characterization as a potential geologic repository.
The Yucca Mountain site has not been selected for a repository; rather, it has been chosen as the only site to be characterized at this time.
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Site characterization is a; program of exploration and research, both in the laboratory and in the field, undertaken to establish the geologic conditior4s and the ranges of those parameters at a particular site.
Sita characterization includes borings, surface excavations, excavation of exploratory shafts or ramps, limited i
l subsurface lateral excevations and borings, and in situ testing at depth to determine the suitability of the site for a geologic repository.
Reculations:
The NRC's requirements governing the disposal of HLW in a geologic repository are contained in Title 10 Code of Federal Regulations, Part 60 (10 CFR Part 60).
These regulations govern prelicensing activities, authorization for DOE to begin construction of the facility, authorization for DOE to receive and place the wastes in the facility, and authorization for DOE to close the facility (license termination).
The NRC's requirements governing the storage of HLW in an MRS facility are contained in Title 10 Code of Federal Regulations, Part 72 (10 CFR Part 72).
These regulations establish requirements, procedures, and criteria for the issuance of licenses to receive, transfer, and possess power reactor spent fuel and other radioactive material associated with spent fuel storage.
j The EPA's standards for the disposal of HLW in a geologic repository are contained in Title 40 Code of Federal Regulations Part 191 (40 CFR Part 191).
These regulations establish generally applicable environmental standards for the management and disposal of spent nuclear fuel and other HLW.
The NRC is responsible for implementing these standards in 10 CFR Part 60.
Current Status:
Currently, the repository program is focused on prelicensing site characterization activities.
In the prelicensing phase, one of NRC's primary responsibilities is to review DOE's site l
characterization plan and associated activities, and to provide comments to DOE identifying any specific concerns.
In addition, NRC staff observes various site characterization activities in the field, such as drilling and tunneling, and also observes DOE quality assurance surveillances and audits.
All prelicensing consultation activities are open to participation by the State of Nevada, affected Indian Tribes, and affected units of local i
governments.
DOE completed its site characterization plan for the Yucca Mountain site in December 1988.
The NRC staff completed its 4
l review of that document in July 1989, and concluded that overall, it was a usable plan for site characterization.
Originally, the
-I BP14 (09/93) staff identified two objections to DOE starting site characterization.
One objection concerned the DOE quality assurance (QA) program, and the other was related to the design process for the Exploratory Studies Facility (ESF).
Additionally, 196 other concerns in the form of comments and questions were raised.
l Regarding the QA objection, NRC notified DOE by letter, dated March 2, 1992, that the objection was removed.
NRC determined that all organizations participating in site characterization activities had developed, and are implementing, a QA program that meets NRC requirements.
NRC continues to monitor QA program implementation through audits and surveillances.
DOE provided information in response to the ESF objection on March 3, 1992.
Based on that information, and information contained in two DOE reports (" Exploratory Studies Facility Alternatives Study: Final Report" and " Risk / Benefit Analysis of Alternative Strategies for Characterizing the Calico Hills Unit l
at Yucca Mountain,") and observations of DOE design reviews, the NRC staff determined that the ESF objection should be lifted.
In a letter dated November 2, 1992, NRC notified DOE that this objection was lifted.
The staff continues to track DOE's activities related to the ESF objection and implementation of the ESF design control process through participation in ESF design reviews, reviews of site characterization Progress Reports, and participation in DOE audits of the ESF design review process.
In 1991, the State of Nevada granted DOE the permits necessary for DOE to proceed with surface based site characterization activities.
These activities include the excavation of test pits and trenches, bore hole drilling, and addressing volcanism, tectonics, and faulting.
DOE continues to actively conduct site l
characterization field work in these areas at the Yucca Mountain Project Site and vicinity.
l On November 30, 1992, DOE began work on the site access road to the main portal for the ESF at Yucca Mountain, and work on the portal pad began in April 1993.
Initial work on the underground opening used the drill and blast method to excavate a 200 foot staging chamber for the tunnel boring machine, which will be used to excavate the main ramp-tunnel of the ESF (about 25,000 feet).
t DOE proposes to complete the excavation of the north ramp portion i
of the main ramp-tunnel of the ESF by July 1994.
In a letter of August 20, 1993, NRC notified DOE of concerns related to the design and design control process based on its observations and reviews of DOE activities.
The letter cites i
deficiencies identified by DOE QA and technical personnel during recent QA audits of DOE's Civilian Radioactive Waste Management l
System Management and Operating Contractor (M&O) and requests that DOE provide a rationale for proceeding with activities i
I i
BP14 (09/93) related to design and construction while the deficiencies are being investigated and corrected.
The deficiencies include inadequate procedures, failure to follow procedures, and inadequate documentation of design bases.
In addition, it is not clear that technical information, identified in DOE reports as necessary for ESF design decisions, will be collected in time to provide input to the ESF design.
Other items requested in the letter include an action plan for corrective actions for the M&O design deficiencies, a controlled baseline ESF design integrated with a geologic repository operations area conceptual design, and a detailed plan for the process DOE will use to keep the staff informed of future design changes.
DOE has been requested to respond to the NRC request within 90 days of the date of the NRC letter or notify the staff within 30 days if that schedule cannot be met.
The NWPA calls for a 1998 date for DOE to accept spent fuel from utilities.
Until recently, the DOE approach has been to use the Nuclear Waste Negotiator for the voluntary siting of an MRS that would begin acceptance of spent nuclear fuel by 1998.
The Office of the Nuclear Waste Negotiator was established by the NWPAA to find a state or Indian Tribe willing to host a repository or MRS at a technically qualified site.
Interest has been expressed by seven groups in evaluating the feasibility of hosting an MRS (Phase IIA).
To date, two of these groups have gone on to Phase IIB where work will focus on detailed environmental and site feasibility studies.
In addition, DOE has recently begun studying the feasibility of locating an MRS on federal land.
DOE's efforts are proceeding in parallel with those of the Nuclear Waste Negotiator in order to meet the 1998 objective established in the NWPA.
In October 1992, DOE initiated a study to evaluate the feasibility of using multi-purpose canisters (MPC) in the waste management system.
The MPC concept is to use a common container that has different overpacks for transportation, storage, and disposal.
The purpose of the MPC is to create first, a compatible approach for the transportation and storage of spent nuclear fuel, and then consider compatibility with final disposal.
DOE expects to issue results from its MPC atudy in 1993.
EPA developed generally applicable environmental standards for a high-level waste repository that were promulgated as 40 CFR Part 191 in 1985.
These standards were remanded in 1987 due to inconsistencies with other EPA standards with respect to individual dose and ground-water protection.
Since that time EPA has been working on revising its standard.
However, in late 1992, Congress passed the Energy Policy Act (EnPA) of 1992 which required EPA to contract with the National Academy of Sciences (NAS) to conduct a study on specific aspects of these standards and issue findings and recommendations.
The NAS plans to
BP14 (09/93) complete this study by the end of 1994.
Among the issues to be included in the Academy's review are:
(1) the reasonableness of a health-based standard based on individual dose; (2) the ability of post-closure oversight to prevent an unreasonable risk of breaching the repository's barriers or_ increasing the exposure of the public to radiation beyond allowable limits; and (3) the capability to make scientifically supportable predictions of the probability of human intrusion for 10,000 years.
The EnPA further requires that the EPA will promulgate by rule, public health and safety standards for protection of the public from releases from radioactive materials stored or disposed of in the repository at the Yucca Mountain site, based upon and consistent with the findings and recommendations of the NAS.
These standards shall be promulgated not later than one year after EPA receives the findings and recommendations of NAS.
Furthermore, the EnPA also requires that NRC, within one year of the promulgation of the EPA standards, amend its technical requirements and criteria to conform with these standards.
CONTACT:
Joseph Holonich, Director Repository Licensing & Quality Assurance Project Directorate Division of High-Level Waste Management (301) 504-3387
BP14 (09/93)
HIGHLIGHTS OF HIGH-LEVEL RADIOACTIVE WASTE (HLW) o NWPA (1982) and NWPAA (1987) lay out a national program for disposal of HLW in a deep geologic repository and possible interim storage in an MRS o
NWPAA designated Yucca Mountain, Nevada, for characterization as a potential repository site o
NRC requirements for the interim storage of HLW are contained in 10 CFR Part 72 o
NRC requirements for the disposal of HLW are contained in 10 CFR Part 60 o
EPA standards for the disposal of HLW are contained in 40 CFR Part 191 o
The EnPA requires that:
NAS conduct a study on specific aspects of the HLW environmental standards and make recommendations on reasonable standards by December 31, 1993; EPA revise its standards based upon, and consistent with, the NAS findings and recommendations within one year after it receives the NAS findings; and NRC amend its technical requirements and criteria to conform to the EPA standard within one year of the promulgation of EPA's standards o
NRC is currently involved in prelicensing interactions and review of DOE HLW repository site characterization activities o
NRC is currently involved in prelicensing interactions and review of DOE MRS activities o
DOE to submit to NRC an MRS application to construct and operate a facility in 1995 o
DOE to begin waste acceptance at an MRS facility in 1998 o
DOE to submit to NRC a HLW repository license application for construction authorization in 2001 o
DOE to begin waste emplacement in a HLW repository in 2010 o
All prelicensing consultation activi'.ies are open to participation by the State of Nevada, affected Indian Tribes, and units of affected local governments
BP15 (09/93)
REMEDIATION OF RESIDUAL RADIOACTIVITY AT NON-ROUTINE CONTAMINATED SITES packoround:
There are about 40 sites contaminated with radioactive material, under the jurisdiction of the Nuclear Regulatory Commission, throughout the country, that are considered to be non-routine decommissioning cases.
To compel timely remediation of these sites, NRC initiated the Site Decommissioning Management Plan (SDMP) in 1990.
NRC emphasis on timely remediation of the sites resulted from former Chairman Carr's August 3, 1989, testimony before the Subcommittee on Environment, Energy, and National Resources, chaired by Congressman Synar, and from continuing NRC concern about lack of progress at some of these sites.
Discussion:
The NRC staff developed the SDMP to identify sites requiring remediation and to provide the Commission with a status report on the actions taken to bring about the remediation of the SDMP sites.
In summary, the SDMP contains the following information:
- 1) definition of the project management plan;
- 2) identification of the sites requiring decommissioning;
- 3) schedule and resources necessary to support NRC actions to regulate the remediation of the contaminated sites;
- 4) resolution of policy and Synar hearing issues for SDMP implementation.
The SDMP not only identifies the sites requiring decommissioning, but also describes:
- 1) the site;
- 2) the wastes and radioactivity remaining on the site;
- 3) the radiological hazard from the remaining wastes;
- 4) the financial assurance required;
- 5) the status of the remediation activities; and
- 6) NRC proposed actions and schedules to ensure timely decommissioning.
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SDMP Criteria:
A site is listed in the SDMP if it meets one or more of the following criteria:
- 1) problems with a viable responsible organization (e.g.,
inability to pay for or unwillingness to perform decommissioning);
- 2) presence of large amounts of soil contamination or unused settling ponds or burial grounds that may be difficult to dispose of;
- 3) long-term presence of contaminated, unused facility buildings;
- 4) license previously terminated; or
- 5) contamination or potential contamination of the groundwater from onsite vastes.
Reaulations:
In 1988, NRC promulgated the final decommissioning regulations, under 10 CFR Parts, 30, 40, 50, 70, and 72.
The regulation defines decommissioning as removing a facility from service, reducing the residual radioactivity to a level that will permit release of the facility for unrestricted use, and termination of the license.
In summary, the new regulation prescribes requirements for decommissioning planning, financial assurance, recordkeeping, and license termination.
SDMP Action Plan:
The SDMP has been effective in ensuring coordination and resolution of some policy and regulatory issues affecting site decommissioning.
Progress on actual site remediation, however, continued to be slow.
Because of that limited progress, prompted the staff developed the SDMP Action Plan, which was approved by the Commission on April 6, 1992.
The Action Plan:
- 1) identifies current criteria to guide remediation of contaminated soils, structures, and equipment and emphasizes site-specific application of the as low as reasonably achievable (ALARA) principle;
- 2) states NRC's position on the finality of decommissioning decisions; I
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BP15 (09/93)
- 3) describes NRC's general expectation that SDMP site remediation will be completed within a 4-year timeframe after operations cease or 3-years after the issuance of an initial decommissioning order;
- 4) identifies currently available guidance on site characterization work in support of decommissioning; and
- 5) describes the process the staff will use to establish and enforce schedules for timely decommissioning on a site-specific basis.
NRC Workshop on the SDMP:
On November 19, 1992, NRC hosted a workshop in Rockville, Maryland, to exchange information and lessons learned in remediating contaminated sites.
The workshop was intended primarily for SDMP site licensees and responsible parties.
However, representatives of state regulatory agencies and private i
consulting firms also attended.
The workshop agenda included i
remarks by Chairman Selin; presentations by NRC on the SDMP, SDMP Action Plan, and future regulations affecting contaminated site remediation; and presentations by four SDMP licensees on their experience in selected aspects of site decommissioning.
Policy Issues Recuirina Resolution:
i The SDMP contains a series of policy issues, related to the i
remediation of contaminated materials licensee sites, that need to be resolved.
Resolution of these policy issues will provide a regulatory framework for more consistent and efficient licensing actions related to site decommissioning in the future.
Two issues require prompt resolution for more effective implementation of the SDMP:
- 1) developing a national standard for residual radiological contamination that is acceptable for releasing a site or materials for unrestricted use; and 2) developing a rule to require timely decommissioning.
The Commission recently initiated an enhanced participatory rulemaking process to establish radiological criteria for decommissioning.
Representatives of State governments, Tribal governments, local governments, other Federal agencies, environmental groups, citizen groups, and industry groups are j
invited to participate in the workshops associated with the l
rulemaking.
The workshops began in January 1993 and ended on May 7, 1993.
Final requirements are scheduled to be promulgated i
in the Fall of 1994.
A proposed rulemaking on the timeliness of decommissioning was published in the Federal Reaister on January 13, 1993.
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Current Status:
The NRC staff is implementing the SDMP as described in SECY-92-200 dated May 29, 1992.
The SDMP was updated in June 1993 and will be published as NUREG-1444.
The decommissioning of the Allied Signal Aerospace - Bendix Division site was completed, and the site removed from the SDMP list, on February 28, 1992.
Decommissioning of the Budd Company site was completed in late 1992, the license was terminated in April 1993, and the site was removed from the SDMP in April 1993.
Two other sites, Amax, Incorporated and UNC Recovery Systems (Wood River Junction), have completed site remediation activities.
The licenses for these two sites will be terminated, and the sites removed from the SDMP list, after pending administrative or jurisdictional issues are resolved.
In addition, decommissioning actions at the Old Vic and TI Attleboro sites are expected to be completed, and the licenses terminated in 1993.
Remediation of the Pawling and United Technologies (Pratt & Whitney) sites is expected to be completed in 1993 (licenses terminated).
Attached is the current list of SDMP sites.
Contact:
John H. Austin, Chief, Decommissioning and Regulatory Issues Branch, Division of Low-Level Waste Management and Decommissioning, Office of Nuclear Material Safety and Safeguards, 301-504-2560.
i i
CONTAMINATED SITE LIST'
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Advanced Medical Systems, Inc.
AlumgumCompanyofAmerica Amax*
Anne Arundel County /Curtis Bay Army, Department of, Aberdeen Proving Ground Babcock & Wilcox, Apollo, PA Babcock & Wilcox, Parks Township, PA BP Chemicals, America, Inc.
Budd Company Cabot Corporation, Boyerton, PA Cabot Corporation, Reading, PA Cabot Corporation, Revere, PA Chemetron Corporation, Bert Avenue 8
Chemetron Corporation, Harvard Avenue Chevron Corporation (formerly Gulf United Nuclear fuels Corporation)
Dow Chemical Company Elkem Metals, Inc.
Engelhard Corporation Fansteel, Inc.
Hartley and Hartley (Kawkawlin) Landfill Heritage Minerals y
Kerr-McGee, Cimarron Plants Kerr-McGee, Cushing Plant I.ake City Army Amunition Plant (formerly Remington Arms Company)
Magnesium Elektro Minnesota Mining and Manufacturing Co. (3M)
Molycorp, Inc., Washington, PA Molycorp, Inc., York, PA Northeast Ohio Reatonal Sewer District / Southerly Plant Nuclear Metal, Inc.
OldVic,Inc.f Permagrain Products Pesses Company, METC0A Site Pratt and Whitney RMI Titanium Company TRI, Inc. (formerly Process Technology of North Jersey, Inc.)
'The Allied signal Aerospace, Bendix Division site was released for unrestricted use and removed from the SDMP in 1992.
'All Issues resolved; in the administrative process of removing site from the SDMP.
3Site may be removed from SDMP in 1993.
'All required actions to remove the site from the SDMP have been completed and the site has been removed fr6m the SDMP. However, a sumary of actions taken in 1992 leading to the site's removal from the list is included in this
(
Appendix.
('
CONTAMINATED SITE LIST (Continued),
Safety Light Corporation Schott Glass Technologies Sequoyah Fuels Corporation Shieldalloy Metallurgical Corporation, Cambridge, OH Shieldalloy Metallurgicp Corporation, Newfield, NJ TexasInstruments,Ipc UNC Recovery Systems Watertown Arsenal / Mall Watertown GSA Westinghouse Electric Corporation, Waltz Mill Site West Lake Landfill Whittaker Corporation Wyman-Gordon Company i
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'The Allied Signal Aerospace, Bendix Division site was released for unrestricted use and removed from the SDMP in 1992.
2All Issues resolved; in the administrative process of removing site from the SDMP.
i 3Site may be removed from SDMP in 1993.
'All required actions to remov4 the site from the SDMP have been completed and the site has been removed from the SDMP. However, a sumary of actions taken in 1992 leading to the site's removal from the list is included in this I
Appendix.
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HIGHLIGHTS o
over 40 sites contaminated with radioactive material are non-routine decommissioning cases.
o Site Decommissioning Management Plan (SDMP) developed in 1990.
o SDMP includes information on identification of sites, schedule, and resources to support NRC actions.
o Decommissioning rule promulgated in 1988 in Parts 30, 40, 50, 70, and 72.
o Decommissioning defined as removing a facility from service, reducing the residual radioactivity to a level that will permit release of the facility for unrestricted use, and termination of the license.
o SDMP Action Plan approved by the Commission on April 6, 1992.
o Policy issues that require resolution:
- 1) development of residual contamination criteria;
- 2) timeliness of decommissioning.
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BP16 (09/93)
ENFORCEMENT PROGRAM
Background:
As the federal agency responsible for regulating the civilian uses of nuclear materials, the NRC has an extensive program with I
many requirements.
These requirements are imposed on 112 nuclear power plant licensees and approximately 8000 materials licensees.
The requirements are stringent and technically demanding.
Inevitably, with such an elaborate regulatory program, violations of requirements occur, through oversight, negligence, ignorance, confusion, and, in some instances, willful misconduct.
The i
Commission has developed an enforcement program that seeks to promote and protect the public health and safety by ensuring compliance with the Atomic Energy Act, the Energy Reorganization Act, NRC regulations, and license conditions, obtaining prompt correction of violations and adverse quality conditions that may affect safety, oeterring future violations, and encouraging improvement of licensee performance.
Program Operation:
1 The enforcement program starts with inspections and investiga-tions to determine whether licensed activities are being conducted in compliance with regulatory requirements.
All j
violations are subject to civil enforcement action.
Following i
identification of a potential violation, it is assessed in i
accordance with the Commission's Enforcement Policy.
This Policy has been approved by the Commission and is published as i
Appendix C to 10 CFR Part 2 of the Commission's regulations.
As
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a policy and not a regulation, the Commission is able to depart from the Policy if circumstances warrant, but in practice, this happens only rarely.
Violations that are done willfully are subject to criminal enforcement action.
These cases are also investigated by the NRC's Office of Investigations (OI) and, if wrongdoing is found, the case is referred to the Department of j
Justice for consideration for prosecution.
i There are three primary enforcement sanctions available:
Notices of Violation, civil penalties, and orders.
(NOV) summarizes the results of an inspection and formalizes a violation.
It states the requirement and how that requirement was violated.
A civil penalty is a monetary fine issued under authority of section 234 of the Atomic Energy Act.
That section provides for penalties of up to $100,000 per violation per day.
NOVs and civil penalties are issued based on violations.
Orders I
may be issued for violations, or in the absence of a violation, because of a safety issue.
The Commission's order issuing authority is broad and extends to any area of licensed activity that affects the public health and safety.
Orders modify, suspend, or revoke licenses.
As a result
BP16 (09/93) of a recent rulemaking, the Commission can now issue orders to individuals who are not themselves licensed.
In addition to the primary sanctions, the NRC has administrative mechanisms available, including Confirmatory Action Letters to confirm in writing a licensee's commitment to take certain immediate corrective actions, and Notices of Deviation that address violations of non-legally binding commitments, such as an industry good practice or standard, l
The first step in the enforcement process is assessing the severity level of the violation.
Severity levels range from 1
l Severity Level I, for the most significant violations, to I
i Severity Level V for those of minor concern.
Eight supplements to the Enforcement Policy provide guidance in determining these severity levels.
A higher severity level may be assigned for cases involving a group of violations with the same root cause, repetitive violations, or willful violations.
Enforcement conferences are held for violations assessed at Severity Levels I, II, or III, and may be held for violations assessed at Severity Level IV if increased management attention is warranted (e.g., repetitive violations).
An enforcement conference is a meeting between the NRC and the licensee to (1) discuss the apparent violations, their significance, the reason for their occurrence, including the apparent root causes, and the licensee's corrective actions, (2) determine whether there were any aggravating or mitigating circumstances, and (3) obtain other information that will help the NRC determine the appropriate enforcement action.
The decision to hold an enforcement conference does not mean that the agency has determined that a violation has occurred or that enforcement action will be taken.
In accordance with the Enforcement Policy, enforcement conferences are normally closed to the public.
However, on July 10, 1992, the Commission implemented a two-year trial program to allow certain enforcement conferences to be open to public observation.
(See further discussion addressed under
" Current Developments.")
Following the enforcement conference, the regional office prepares the proposed enforcement action.
All Severity Level I and II cases, and some Severity Level III cases are sent to Headquarters for processing and approval.
Routine Severity Level III materials cases, and Severity Level IV and V matters are issued directly from the regional office.
In the absence of mitigating circumstances, civil penalties are normally issued for Severity Level III or higher violations and may be issued for violations at Severity Level IV if the violations are repetitive or similar to previous Severity Level IV violations.
In addition, a civil penalty may be issued for any willful violation.
BP16 (09/93) l If a civil penalty is to be proposed, the base value of the penalty must first be determined.
The base value is based on a combination of the type of licensed activity, the type of licensee, and the severity level of the violation.
Once the base i
value is determined, a number of factors are considered that may aither escalate or mitigate the amount of the civil penalty, depending on the unique circumstances of the case.
These factors l
are: (1) who identified the violation, (2) was the corrective action prompt and extensive or untimely and only marginally j
acceptable, (3) how was the past performance of the licensee, (4) did the licensee have prior notice of similar events or other indications that should have alerted management, (5) were there multiple examples of the violation, and (6) what was the duration of the violation.
If a civil penalty is to be proposed, a written Notice of Violation and Proposed Imposition of Civil Penalty is issued and the licensee has 30 days to respond in writing, by either paying the penalty or contesting it.
The NRC considers the response, I
and if the penalty is contested, may either mitigate the penalty or impose it by an order.
If the civil penalty is to be imposed by order, the order is published in the Federal Register.
Thereafter, the licensee may.
pay the civil penalty or request a hearing.
i In addition.to civil penalties, orders may be used to modify, suspend, or revoke licenses.
Orders that modify a license may require additional corrective actions, such as removing specified individuals from licensed activities or requiring additional controls or outside audits.
The NRC issues a press release with a proposed civil penalty, order, or Confirmatory Action Letter.
Current developments:
On September 16, 1991, the Commission implemented a new rule that would allow orders to be issued to individuals.
The deliberate misconduct rule applies to a licensee, an employee of a licensee, a contractor, subcontractor, or employee of them who knowingly l
provides to a licensee or contractor components or any other goods or services that relate to licensed activities.
This rule prohibits (1) engaging in deliberate misconduct that causes, or but for detection would have caused, a licensee to be in violation of any NRC requirement, or (2) deliberately submitting to NRC, a licensee or contractor, or subcontractor, information known to be incomplete or inaccurate in some respect material to the NRC.
Deliberate misconduct means an intentional act or omission that the person knows would cause or is a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee or contractor, whether or not the person i
knew a resulting violation of NRC requirements would occur.
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i BP16 (09/93)
An order issued under the deliberate misconduct rule might order the wrongdoer to remain out of licensed activities for a specified period, to notify the NRC before resuming involvement in licensed activities, or to inform any prospective employer of the issuance of the order.
The order might require the employer to remove or confirm the removal of an employee from licensed activities, require the employer to advise prospective employers of the existence of the order when they call for reference checks, or require notice to the NRC if a licensee employs or desires to reemploy a wrongdoer in licensed activities.
Parties affected by orders may request a hearing before an Administrative Law Judge or a panel of the Atomic Safety and Licensing Board.
Further appeal to the Commission, and ultimately the Court of Appeals, is possible.
In the recent rulemaking, the Commission also established a clearer mechanism for obtaining information from a licensee when the NRC is considering enforcement action.
A Demand for Information may be issued to a licensee, requiring submission of a response to specific questions.
On July 10, 1992, the Commission implemented a two-year trial program to allow certain enforcement conferences to be open to public observation.
Under the trial program, approximately 25 percent of all eligible enforcement conferences will be open to public observation, at least one conference will be open in each regional office, and open conferences will be conducted with a variety of types of licensees.
Conferences that will be open to public observation will be selected on a random basis by selecting every fourth eligible conference involving a commercial operating reactor, hospital, or other type of licensee.
In j
addition, conferences may also be open in cases where there is an ongoing adjudicatory proceeding with one or more intervenors.
Enforcement conferences will not be open to public observation if the enforcement being action being contemplated (1) may be taken against an individual or, if the action, though not taken against the individual turns on whether an individual has committed wrongdoing; (2) involves significant personnel failures where the NRC has requested that the individual (s) involved be present at the conference; (3) is based on the findings of an NRC OI report; or (4) involves safeguards, Privacy Act, or proprietary information.
Enforcement conferences will also be closed if the conference will be conducted by telephone or at a relatively i
small licensee's facility.
The NRC intends to announce open enforcement conferences to the public normally 10 working days in advance of the conference through the following mechanisms:
(1) notices posted in the Public Document Room; (2) toll-free telephone messages; and (3)
BP16 (09/93) toll-free electronic bulletin board messages.
Pending establishment of the toll-free ma.ssage systems, the public may call (800) 952-9674 to obtain a recording of upcoming open enforcement conferences.
Contact:
James Lieberman, Director, Office of Enforcement US NRC, Washington, DC 20555.
(301) 504-2741.
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BP16 (09/93)
HIGHLIGHTS OF ENFORCEMENT PROGRAM o
Enforcement program seeks to protect public health and safety by ensuring compliance and correction of violations and deterring future violations.
o Violations are detected through inspections and investigations.
o Violations are subject to civil enforcement action and may be subject to criminal prosecution.
o Civil enforcement sanctions include: Notices of Violation, civil penalties, and orders.
o Severity level of a violation reflects the significance of the violation and ranges from the most significant, Severity Level I, to the least, Severity Level V.
o Civil penalties are normally issued for Severity Level III or higher violations.
l 0
The amount of a civil penalty assessed varies with type of licensed activity, type of licensee, severity level, and escalation and mitigation factors.
o If a civil penalty is proposed, licensee may respond by paying or contesting the action.
o If licensee protests action, staff considers response, and either mitigates the penalty or imposes it by order.
o Licensee must then pay or request an administrative hearing.
o orders may be used to modify, suspend, or revoke a license.
O orders may also address deliberate wrongdoing by individual employees of licensees, contractors, or others who provide j
goods or services that relate to licensed activities.
o An order to an individual might remove him or her from licensed activity, require NRC notification of the individual's reemployment in licensed activities, or require notification to prospective employers of the existence of an order.
o NRC may use Demands for Information to obtain information j
when considering enforcement action.
i o
In a two-year trial program starting July 10, 1992, one of every four eligible enforcement conferences involving a commercial operating reactor, hospital, or other type of licensee will be open to public observation.
O A schedule of open enforcement conferences can be obtained by calling (800) 952-9674.
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BP17 (09/93)
INSPECTION OF NUCLEAR POWER PLANTS
Background:
The primary safety consideration in the operation of any nuclear reactor is the control and containment of radioactive material under both normal and accident conditions.
Many controls are established to protect workers and the public from the effects of radiation.
The industry and the NRC both have roles in providing these controls and in ensuring that they are maintained.
The NRC establishes rules, regulations, and guides for the construction and operation of nuclear reactors.
Organizations licensed by the NRC must abide by these regulations and are directly responsible for designing, constructing, testing, and operating their facili-ties in a safe manner.
The NRC, through its licensing and inspection programs, provides assurance that its licensees are meeting their responsibilities.
Inspection Program:
The responsibility for safe operation of a nuclear power plant lies with the licensee.
The NRC inspection program is designed to conduct selective examinations to ensure that the licensee is meeting his prescribed responsibility.
The NRC inspection program is audit oriented, thus, it does not undertake to examine every activity or item, but verifies, through carefully selected samples, that the activities are being properly conducted or operated to enhance or ensure safety.
Areas to sample, the sizes of the samples, and the frequencies of the inspection efforts are based on the importance of the activity or system to overall i
safety and available resources.
The inspection program is preventive in nature and is intended to anticipate and preclude j
significant events and problems by identifying underlying issues.
a The inspection process, from a systems approach, monitors the licensee's activities and provides feedback to the licensee's j
plant management to allow it to take appropriate corrective
)
actions.
However, implementation of the NRC inspection program i
does not supplant the licensce's programs or responsibilities.
Rather, the inspection program provides an independent verification of the effectiveness of the licensee's implementation of its programs, to ensure that operations are being conducted safely and in accordance with applicable NRC requirements.
Inspections are performed on power reactors under construction, in test and in operation.
The inspections are conducted primarily by region-based and onsite resident inspectors.
Resident inspectors are stationed at each reactor under i
construction and in operation.
Region-based inspectors operate i
l
BPl7 (09/93) from the five regional offices located in or near Philadelphia, Atlanta, Chicago, Dallas, and San Francisco.
These programs are supplemented by inspections conducted by special teams comprised of personnel from both headquarters and regional offices.
Inspections are part of NRC's review of applications for licenses as well as NRC's issuance of construction permits and operating licenses.
Inspections continue throughout the operating life of a nuclear facility.
Prior to construction, the inspection program concentrates on the applicant's establishment and implementation of a quality assurance program.
Inspections cover quality assurance activities related to design, procurement and the plans for fabrication and construction.
During construction, a sampling of licensee activities is inspected to confirm that the requirements of the construction permit are followed and that the plant is built according to design and applicable codes and standards.
Construction inspections look for qualified personnel, quality material, conformance to approved design, and for a well-formulated and satisfactorily implemented quality assurance program for ensuring the quality of construction.
As construction nears completion, preoperational testing to demonstrate the operational readiness of the plant and its staff begins.
Inspections during this phase determine whether the licensee has developed adequate test plans, assure that tests are consistent with NRC requirements, and determine that the plant and its staff are prepared for safe operation.
Inspections during the preoperational phase involve (1) reviewing overall test management procedures; (2) examining selected test procedures for technical adequacy; and (3) witnessing and reviewing selected tests to determine their outcomes and the consistency of planned and actual tests.
In addition, inspectors review the qualifications of operating personnel and assure that operating procedures and quality assurance plans are developed and implemented.
About six months before the operating license is issued, a l
startup phase begins in preparation for fuel loading and power ascension.
After issuance of the operating license, fuel is loaded into the reactor and the actual startup test program begins.
As in preoperational testing, NRC inspection emphasis is placed on test management procedures and results.
The licensee's management system for startup testing is examined, test procedures are analyzed, tests are witnessed, and licensee evaluations of test results are reviewed.
I When startup testing is completed satisfactorily, routine operations begin.
Thereafter, NRC continues its inspection 1
l
l BPl7 (09/93) program throughout the operating life of the plant.
The responsibility for safe operation of a nuclear plant lies with the licensee.
The NRC inspection program is designed to conduct selective examinations.
An onsite resident inspector provides a continual inspection and regulatory presence, as well as a direct contact between NRC management and the licensee.
The resident inspector is also the key individual in the regional office's determination of what additional inspections need to be accomplished at a specific plant.
The inspection ectivities of the resident inspector are supplemented by the efforts of engineers and specialists from the regional office staff who perform inspections in a wide variety of engineering and scientific disciplines ranging from civil and structural engineering to health physics and reactor core physics.
The specialist inspectors provide a perspective that is different from, but complementary to, that of the resident who, of necessity, is a generalist.
Since the specialists inspect many different plants, they see many different ways of accomplishing a function, and they develop a more comprehensive view of their specialties.
The inspection program for operating reactors is defined in the NRC Inspection Manual in terms of frequency, scope, and depth.
Detailed inspection procedures provide instructions and guidance for NRC inspectors.
The program consists of three major elements:
core inspections, the minimum done at all plants; area of emphasis inspections, special inspections to focus on a specific issue; and initiative inspections, those which are required to adequately resolve safety issues identified by other inspections or as a result of plant events.
The program is structured to ensure that the finite resources available for inspection are used efficiently and effectively, with increased attention being devoted to those plants where, based on licensee performance, improvements in safety may be needed.
The inspection program is an important element in the NRC's regulatory program.
Its results are factored into NRC's overall evaluation of licensee performance under the Systematic Assessment of Licensee Performance (SALP) program designed to ensure that nuclear power reactors are constructed and operated safely and in compliance with regulatory requirements.
When a safety problem or failure to comply with requirements is identified, the NRC directs prompt corrective action by the licensee, and takes, as necessary, appropriate enforcement action.
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BP17 (09/93)
In February 1989 the Commission announced a policy on cooperation with States which allows States to observe, and in some cases participate in, NRC inspections at reactor facilities.
Contact:
Mark Peranich, Chief, Operating Reactor Inspection Section, Inspection and Licensing Policy Branch US NRC, Washington, DC 20555 (301) 504-3078 7
BP17 (09/93)
HIGHLIGHTS OF INSPECTION PROGRAM o
NRC, through its licensing and inspection programs, ensures licensees are meeting their responsibilities for
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constructing and operating nuclear reactors in a safe manner.
O NRC inspection program is independent, audit oriented, and preventive in nature to anticipate and preclude significant events by identifying underlying issues.
O Inspection process provides feedback to the licensee's plant management to allow it to take appropriate corrective action.
O Inspections performed by resident, regional, and headquarters inspectors.
O Inspections performed prior to construction, during construction, during preoperational testing, and routinely after the plant is in operation.
o Inspection program defined in NRC Inspection Manual and detailed inspections procedures exist for:
1) core inspections, the minimum done at all plants; 2) area of emphasis inspections, special inspections to focus on a specific issue; 3) initiative inspections, those which are required to adequately resolve safety issues.
O Inspection results factored into the overall evaluation of licensee performance under the Systematic Assessment of Licensee Performance (SALP) program.
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BP18 (09/93)
ENVIRONMENTAL EFFECTS OF NUCLEAR POWER PLANTS i
Background:
The discharge of radioactive effluents from the routine operation of nuclear power plants can result in environmental impacts.
These impacts can be on man, and terrestrial and aquatic biota.
In nearly all cases a Final Environmental Statement (FES) was issued by the NRC which details the potential impacts resulting from the routine operation of each nuclear power plant.
The NRC regulation, 10 CFR 50.36a, " Technical Specifications on Effluents from Nuclear Power Reactors," requires that each license authorizing operation of a nuclear power reactor include technical specifications that (1) keep releases of radioactive material to unrestricted areas during normal operation as low as is reasonably achievable, and (2) comply with the applicable provisions of 10 CPR 20.106, " Radioactivity in Effluents in Unrestricted Areas" or Subpart D,
" Radiation Dose Limits for Individual Members of the Public," of the revised Part 20.
These and other NRC regulations require NRC licensees to have in place various effluent and environmental monitoring programs to ensure that the impacts are minimized.
Environmental Effects:
Radioactive effluent releases and their associated doses are reported by licensees in Annual Radioactive Effluent Release Reports (recently changed from Semiannual) while radioactivity levels in various environmental media are reported in the Annual Environmental Operating Reports (AEOR).
The Radioactive Effluents Release Reports include the amount of liquid and airborne radioactive effluents discharged and the calculated doses for the period of release.
Doses which are typically presented for airborne effluents include the beta and gamma air i
doses from noble gases and the maximum organ dose from i
radioiodine and particulates.
These doses are compared to the design objective doses of Appendix I to 10 CFR 50.
For liquid i
effluents, the total body and maximum organ doses are typically presented and are also compared to the design objective doses of Appendix I.
The AEOR provides the results of an environmental sampling and analysis program which is focused on the radiation exposure i
pathways specific to the given plant.
Typical sampling programs include an inner and outer ring of TLDs (thermoluminescent dosimeters) around. the plant; airborne radiciodine and particulate samplers; samples of surface, groundwater, and drinking water and downstream shoreline sediment from existing or potential recreational facilities; and samples of ingestion pathway sources such as milk, fish and invertebrates, and food products including broad leaf vegetation.
The results of the i
i l
l
i BP18 (09/93)
AEOR report are used to supplement the effluent monitoring program to ensure that potential impacts do not go undetected.
Regulations:
The regulations, which are presently in place to limit offsite releases and their associated radiation doses, are much more restrictive than those initially issued and those which the first licensed nuclear power plants (during the period of the 1960s) were required to meet.
On December 3, 1970, the Atomic Energy Commission, the predecessor Agency to the NRC, established new regulations, 10 CFR 50.34a and 10 CFR 50.36a, which specified the design and operating requirements for nuclear power reactors to keep levels of radioactivity in effluents "as low as practical."
These regulations provided non-numerical criteria for determining when design objectives and operations met the specified requirements.
On May 5, 1975, the NRC amended regulations 50.34a and 50.36a and added a new Appendix I to Part 50 which provided numerical guides for design objectives and limiting conditions for operation to meet the criteria "as low as is reasonably achievable."
The adoption of these regulations meant that the limiting criterion for nuclear power plant effluents was no longer Part 20, but the design objectives of Appendix I.
An additional regulatory requirement was placed on uranium fuel cycle licensees with the provisions of 40 CFR 190, " Environmental Radiation Protection Standards for Nuclear Power Operations,"
which was promulgated by the Environmental Protection Agency on December 1, 1979.
These standards established total body, thyroid, and other organ dose limits for all effluents and direct radiation except radon and its daughters.
The NRC subsequently incorporated the EPA regulations into 10 CFR Part 20 by reference on March 25, 1981.
Current Status:
In addition to the Annual Effluent Release Reports and the AEOR, the NRC uses other means of verifying that licensees fully evaluate the potential impacts of their operations.
The NRC contracts with some 35 states which perform various environmental monitoring programs including environmental sampling and analysis around nuclear power plants.
The NRC also has a mobile laboratory which is used during plant inspections to confirm, using split samples, the accuracy of the licensee's radiological effluent monitoring program.
Licensees are also required to participate in an Interlaboratory Comparison Program which provides an independent check of the accuracy and the precision of the measurement of radioactive material in environmental
BP18 (09/93) samples.
The NRC also conducts period aerial surveillance of nuclear power plants in which measurements of direct radiation are made.
The NRC documents the results of its independent monitoring and assessment efforts in plant specific inspection reports and in reports such as NUREG/CR-2850, " Dose Commitments Due to Radioactive Releases from Nuclear Power Plant Sites," and NUREG-0837, "NRC TLD Direct Radiation Monitoring Network."
The NRC also compiles licensees' effluent reports and publishes the results in NUREG/CR-2907, " Radioactive Materials Released from Nuclear Power Plants."
With the implementation of the design objective doses of Appendix I and their associated technical specifications, the effluents released from nuclear power plants have decreased significantly.
The collective annual radiation dose to the public has decreased from 1300 person-rem in 1975 (the initial year for which such evaluations were published) to 84 person-rem in 1989.
When normalized to the electric power generation, collective doses 12 have decreased fron 7.6 to 0.16 person-rem /TW-hr (TW=10 watts).
Collective dose represents the sum of doses to individuals within
- 8) km (50 mi) of each facility.
A significant contributor to the reduction in airborne effluents has been the addition of the Augmented Offgas Systems to boiling water reactors.
Reductions have also resulted from improved fuel performance and licensees' improved effluent control programs.
Contact:
LeMoine J.
Cunningham, Chief, Radiation Protection Branch U.S. NRC, Washington, DC 20555 (301) 504-1086 l
l
BP18 (09/93)
EINIRONMENTAL EFFECTS OF NUCLEAR POWER PLANTS
+
Final Environmental Statements on each plant detail the i
potential impacts resulting from routine operation.
t
+
Licensees report radiation releases, including liquid and airborne, and their associated doses in the Annual Effluent Release Report (recently changed from Semiannual).
+
Licensees report radioactivity levels from the environmental sampling of air, ground, water, and ingestion pathways in the Annual Environmental Operating Reports.
?
10 CFR 50.36a requires that each license for a nuclear power reactor include technical specifications that not only keep releases of radioactive material to unrestricted areas as low as is reasonably achievable, but also comply with 10 CFR 20, Subpart D (old 10 CFR 20.106).
10 CFR 20, Subpart D establishes dose limits due to the radioactivity in effluents in unrestricted areas.
+
The NRC documents licensee's effluent releases and the results of the staff's independent monitoring and assessment effort in NUREG-2907, NUREG-2850, NUREG-0837, and plant specific inspection reports.
+
40 CFR 190 sets standards for annual doses to a member of the public for uranium fuel cycle licensees.
+
The staff verifies that licensees evaluate potential radiological impacts through onsite inspections, a mobile lab, state contracts, interlaboratory comparison, and aerial surveillance.
+
10 CFR 50.34a & Appendix I provide numerical guides for design objectives and limiting conditions for operation to meet as low as is reasonably achievable criteria.
4 Effluent releases result in very small doses to members of the public living around nuclear power plants.
BP19 (09/93)
URANIUM MILL TAILINGS l
Backcround:
A lack of orders for new nuclear power plants and the importation of uranium from other countries have now severely eroded the value of uranium and resulted in most U.S. uranium mills shutting down operations or operating on a limited basis.
t Many mills are, or will be, conducting reclamation of tailings piles created in the process of extracting source material (in the form of " yellow cake") from uranium-bearing ore.
These mill tailings wastes, both from unlicensed inactive mills (formerly used in providing uranium primarily for the weapons program) and from licensed active mills regulated by the Nuclear Regulatory Commission or the Agreement States, pose a long-term hazard to the public health and safety.
To provide for the disposal, long-term stabilization, and control of these uranium mill tailings in a safe and environmentally sound manner, Congress enacted the Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA).
In terms of health hazard, the most hazardous radioactive constituent in uranium mill tailings l's radium, which has a very long half-life.
Radium, besides being hazardous itself, produces radon, a radioactive gas whose decay products can cause lung cancer.
This makes mill tailings hazardous for thousands of years.
UMTRCA established two programs to protect public health and the environment from uranium mill tailings.
The Title I program established a joint Federal / State-funded program for remedial i
action at abandoned mill tailings sites, with ultimate Federal I
ownership under license from NRC.
Under Title I, NRC must j
evaluate the Department of Energy's (DOE) and concur that DOE's
{
actions for cleanup and remediation of these inactive tailings sites meet standards set by the Environmental Protection Agency (EPA).
The Title II program is directed towards the active mill tailings sites (those sites under license by NRC or Agreement l
States).
Title II provides:
(1) NRC authority to control radiological and nonradiological hazards; (2) EPA authority to set generally applicable standards for both radiological and nonradiological hazards; and (3) eventual State or Federal ownership, under license from NRC.
For Agreement States, NRC I
also is required to make a determination that all applicable standards and requirements have been met by uranium mills licensed by Agreement States beforr terrinstion of their i
licenses.
1 BP19 (09/93)
Regulations and Standardst UMTRCA charged the EPA with the responsibility for promulgating standards for control of uranium mill tailings.
In 1983, EPA issued standards in 40 CFR Part 192 for both Title I sites 1
(Subparts A, B, and C) and Title II sites.(Subparts D and E).
The portion of the EPA Title I standards dealing with ground water requirements were remanded by the Tenth Circuit Court of Appeals on September 3, 1985.
Based on this court decision, EPA was directed to promulgate new groundwater standards.
EPA proposed these standards in the form of revisions to Subparts A-C of 40 CFR Part 192 in September, 1987, and now is in the process of completing action to promulgate the final groundwater standards.
For Title II sites, NRC conformed its standards in 10 CFR Part 40 to be consistent with the applicable EPA standards.
NRC's final regulations conforming to EPA's requirements for radiological and nonradiological protection and long-term stabilization of the impoundments for the tailings were published on October 16, 1985.
NRC's final regulations addressing EPA's groundwater protection standards were published on November 13, 1987.
EPA, in developing its mill tailings standards, estimated that its standards would significantly reduce radon emissions from tailings and approximately 600 lung cancer deaths per century would be avoided.
Since the EPA standards require that the impoundments for the tailings must be designed to be stable for 1,000 years, to the extent practicable, but in no case less than 200 years, it is assumed that the actual engineered structures will degrade slowly over possibly thousands of years.
Therefore, the use of the standards and NRC's implementing regulations could result in the avoidance of tens of thousands of calculated radon-related lung cancer deaths.
Current Status:
Title I -- Reclamation Work at Inactive Tailings Sites Twenty-four inactive mill tailings piles at 22 sites designated by DOE range in size from about 60 thousand to 4.6 million cubic yards of material.
Except for a site at Canonsburg, PA, the inactive sites are located in the western United States (Arizona, Colorado, Idaho, New Mexico, North Dakota, Oregon, Texas, Utah, and Wyoming).
The DOE surface remediation program is estimated to cost approximately $1.3 billion and is expected to be completed by 1998.
The DOE groundwater cleanup phase was initiated by DOE in 1991.
DOE is in the process of developing a Programmatic Environmental Impact Statement (PEIS) for its groundwater remediation phase.
DOE's initial cost estimates for groundwater remediation were over $500 million with a completion date of 2020.
DOE has completed remedial work, with the
BP19 (09/93) exception of groundwater cleanup, at approximately half of the Title I sites.
It is anticipated that in 1994, DOE will become
)
an NRC licensee for the long-term custodial care of from 3 to 5 of the completed mill tailings sites.
Title II -- Licensed Mill Tailings Sites Of 27 NRC licensed uranium recovery facilities, there are 19 mills; 5 insitu leach facilities; 1 mine-backfill; 1 ion-exchange facility and 1 heap leach.
None of the uranium mills are operating.
Three of the insitu leach facilities are presently operating.
There are four mills and two insitu facilities that have standby licenses which would permit these facilities to go into operation in a relatively short time.
Most of the NRC licensed facilities are either expected to begin, or have already l
started, reclamation activities to provide long-term stabilization and closure of the tailings impoundments.
While i
NRC has terminated insitu leach facility licenses, it has not, as yet, terminated a license for any uranium mill facility.
These NRC-licensed sites are located in Nebraska, New Mexico, Utah, and Wyoming.
There also are 10 uranium mills in Agreement States 4
(Colorado, Texas, and Wasnington) that have similar non-operational tailings impoundments.
In the fall of 1991, NRC, EPA, and the affected mill tailings Agreement States agreed that there was a need to eliminate the dual regulation created by NRC's authority under UMTRCA and the Atomic Energy Act of 1954, as amended, and EPA's authority under the Clean Air Act (CAA).
This interagency consultation resulted
)
in the execution of a Memorandum of Understanding (MOU) to
]
provide the basis for eliminating the dual regulation by EPA under the CAA.
Current activities are addressing the disposal standards in Subpart T of 40 CFR Part 61.
The radon flux standards in Subpart T are the sane as those under UMTRCA.
Consequently, the primary focus of the MOU is to ensure that non-operational piles are closed to comply with the radon standards as expeditiously as practicable, with a goal that such closure occur by the end of 1997.
The MOU specifies that the schedules for closure be enforceable by NRC or the affected Agreement States.
The MOU further provides that the dual regulation of operational sites under Subpart W of Part 61 will be addressed subsequently.
In the late summer of 1992, NRC was requested by EPA, the American Mining Congress, the Natural Resources Defense Council, and the Environmental Defense Fund to sign a settlement agreement related to EPA's Part 61, Subpart T standards and associated stays of the standards.
Although NRC is not a party in the underlying lawsuits, the Commission. recognized that many of the actions outlined in the settlement agreement requiring NRC or Agreement State action were an integral part of the agreement.
i
BP19 (09/93)
While the Commission did not support signing the settlement agreemenc, it did direct staff to prepare a letter to the affected parties in which the Commission would indicate it was in general accord with the substantive provisions of the draft final settlement agreement dated November 24, 1992.
Contact:
John J.
Surmeier, Chief, Uranium Recovery Branch U.S. NRC, Washington, DC 20555 (301) 504-3439
BP19 (09/93)
URANIUM MILL TAILINGS HIGHLIGHTS Uranium Mill Tailings Radiation Control Act of 1978 (UMTRCA) e established a comprehensive regulatory framework for all uranium mill tailings for long-term custody and control.
Title I of UMTRCA established a joint Federal-State funded e
program for remedial action at sites designated by Congress as " inactive" as of 1978, tiith ultimate Federal ownership under license from NRC.
The 24 Title I tailings piles, except for the site at e
Canonsburg, PA, are located in the western United States.
The DOE surface remediation work is estimated to cost approximately $1.3 billion and to be completed by 1996.
DOE anticipates that all of these sites will be under license from NRC by 1998.
The DOE groundwater cleanup phase was initiated in 1991.
DOE is in the process of developing a Programmatic Environmental Impact Statement for this groundwater cleanup phase.
Groundwater cleanup has been estimated by DOE to cost over $500 million.
Title II of UMTRCA provided:
(1) NRC authority to control e
radiological and nonradiologial hazards; (2) EPA authority to set generally applicable standards for both radiological and nonradiological hazards; and (3) eventual State or Federal ownership under license from NRC.
There are 27 uranium recovery facilities licensed by NRC under 10 CFR Part 40 in conformance with EPA generally applicable standards in 40 CFR Part 192.
There are 15 NRC licensed and 8 Agreement State-licensed e
facilities that have non-operating mill tailings impoundments.
An NRC-EPA Agreement State MOU of October 1991 provides the basis for eliminating dual regulation by EPA under the Clean Air Act and establishes a schedule for closure of the tailings impoundments at these 23 non-operating sites by the end of 1997.
1 1
BP20 (09/93)
STATE COMPLIANCE WITH 1993 MILESTONE AND 1996 LEGISLATIVE OBJECTIVE OF THE LOW-LEVEL RADIOACTIVE WASTE POLICY AMENDMENTS ACT OF 1985
Background:
The Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA) (Pub. L.99-240) establishes a series of milestones, penalties, and incentives for regional compacts and States to promote progress toward being able to manage their low-level radioactive waste (LLW) by 1993.
However, slow State progress in developing new LLW disposal facilities has resulted in the storage of LLW at some generator sites, even before January 1, 1993.
This paper includes background information on the LLRWPAA and the status of LLW disposal facility development.
Related regulatory actions are summarized, including a proposed rulemaking that would establish criteria for on-site storage of LLW after January 1, 1996.
LLW Disposal:
Low-level radioactive waste is a general term for a variety of radioactively contaminated wastes generated by nuclear power plants and related industries, hospitals, medical and educational research institutions, private and governmental laboratories, and other commercial activities that use radioactive materials as a part of their normal operations.
Approximately 48,000 cubic i
meters (1.7 million cubic feet) of LLW were disposed of in 1992.
LLW is currently disposed of using shallow land burial at privately operated facilities located in the States of South Carolina and Washington.
The Low-Level Radioactive Waste Policy Act of 1980 (LLRWPA) made the States responsible for the disposal of commercially generated j
and certain Federally generated LLW.
The LLRWPA encouraged the States to form compacts to dispose of LLW on a regional basis.
The LLRWPA also designated January 1, 1986, as the date after which compacts could restrict the use of their disposal facilities by excluding waste generated outside the compact region.
However, by 1983, it had become clear that no new disposal facilities would be operational by the 1986 milestone.
As a result, in January 1986, the LLRWPAA was enacted.
The LLRWPAA extended the January 1, 1986, deadline by seven years, to January 1, 1993.
By that date, new LLW disposal facilities were expected to be operational, and the rights of the LLW generators, to dispose of their LLW at the three sites then operating, would end.
At present, nine compacts have been formed, representing 42 States.
The accompanying figure shows the current arrangements of compacts and unaffiliated States (i.e., those States not in a I
BP20 (09/93) compact).
Legislation to establish the Texas Low-Level Radioactive Waste Disposal Compact (Texas Compact) was signed by Texas Governor Ann Richards on June 9, 1993.
The legislation provides that if either of the other two designated compact member States (Maine and Vermont) ratifies the compact, the agreement will be in full force for Texas and the ratifying State.
Legislation ratifying the compact was enacted in Maine on June 21, 1993.
However Maine law requires that the compact be approved by a majority of Maine voters before the legislation takes effect.
A referendum is scheduled for November 2, 1993.
The Vermont Legislature is currently out of session, and a compact bill is not expected to be introduced until the next regular legislative session in January 1994.
The U.S.
Congress must then consent to the Texas Compact.
The LLRWPAA required the sited States of Nevada, South Carolina, and Washington to make disposal capacity available to LLW generators until December 31, 1992, subject to certain conditions.
Although the Washington facility remains open, serving the Northwest Compact (its compact) and, under contract, the Rocky Mountain Compact, the Nevada facility stopped accepting LLW on December 31, 1992.
The facility in South Carolina will remain open until January 1, 1996, subject to various conditions.
However, the facility will close permanently to out-of-region waste on July 1, 1994.
Widespread storage is expected after this date.
The importation of out-of-region waste to the South Carolina j
facility, during the period of January 1, 1993, to June 30, 1994, will not be approved for States or compact regions that are not making adequate progress toward providing for disposal of their own LLW.
The States of Michigan, New Hampshire, and Rhode Island, the District of Columbia, and Puerto Rico are not eligible for access to the South Carolina facility under these conditions.
Michigan generators began storing LLW in November 1990, when they were denied access because Michigan was no longer in compliance with the LLRWPAA.
New Hampshire's Seabrook Nuclear Power Plant has also been storing LLW since October 1990, because a contract with the Rocky Mountain compact excludai it.
On April 14, 1993, the Southeast Compact commission voted to terminate its contract with the Central Interstate LLW Commission, for access to the South Carolina facility, effective July 1, 1993.
Nebraska is the Host State for the Central Compact.
To help ensure that the States make adequate progress to develop new LLW disposal facilities, the LLRWPAA established six milestones by which the States should make decisions and commit to certain actions.
The majority of the States met the requirements of the three milestone dates that had passed by January 1990.
However, only the Central, Central-Midwest, and Southwestern Compacts met the January 1, 1992, milestone
BP20 (09/93) requirement, where their respective Host States of Nebraska, Illinois, and California submitted applications for disposal facilities.
The State of Texas conformed to this milestone on March 2, 1992, by submitting a disposal facility license application.
However, no State had a new LLW disposal facility operational on January 1, 1993, as envisioned by the LLRWPAA.
The remaining milestone of the LLRWPAA, as it was enacted, is:
January 1, 1996 - The States, upon proper notice by the generator or owner, shall take title to and be obligated to take possession of LLW.
The State will also be liable for all damages directly or indirectly incurred by the generator or owner if it fails to take possession as soon after January 1, 1996, as the generator or owner notifies the State that the waste is available for shipment.
This section of the LLRWPAA, requiring the States to take title to, and possession of, the generated waste by January 1, 1996 (of ten referred to as the "take-title" provision), was held to be unconstitutional by the U.S.
Supreme Court, on June 19, 1992.
Progress to Develop New LLW Disposal Facilities:
Only two new facilities are now scheduled to be operational by January 1996, those in California and North Carolina, and the latter will replace the existing Barnwell facility.
LLW disposal facilities in the Host States of Massachusetts, New Jersey, and Pennsylvania are forecast to be operational between the period 1996 and 1998.
New York's facility is scheduled for operation in 2001.
There are no schedules yet available for the Host States of Connecticut, Illinois, Nebraska, and Ohio.
The unaffiliated States of Michigan, New Hampshire, and Rhode Island, as well as the District of Columbia and Puerto Rico do not have a disposal site under development.
A number of States believe that they may be able to fulfill their responsibilities through the contracting and/or compact process.
The accompanying table shows the dates by which compact Host States and unaffiliated States accomplished, or expect to accomplish, key steps in developing new disposal fscilities.
On October 9, 1992, the Illinois Low-Level Radioactive Waste Disposal Facility Siting Commission voted unanimously to reject the LLW disposal site proposed by the Illinois Department of Nuclear Safety (IDNS).
In support of this decision, the Governor of Illinois directed the IDNS to stop all efforts to locate a facility at the selected site, at Martinsville.
The facility had been scheduled for operation early in 1995.
On December 24, 1992, legislation was enacted that abolished the Siting Commission, repealed the statutory siting criteria, and directed IDNS to recommend a new process for timely and cost-effective establishment of a disposal facility.
Legislation changing the
BP20 (09/93) procedure for siting an LLW disposal facility was signed into law on March 3, 1993, and amended in July 1993.
On December 8, 1992, a straw poll of county residents was conducted to measure the level of public acceptance of the Boyd County site in Nebraska.
Over 90 percent of those voting in the election opposed the construction of the facility.
After the poll results were announced, Governor Nelson stated that he intended to seek a court ruling on whether the vote allows Nebraska to stop the siting process.
In a letter to the Central Interstate LLW Compact Commissioners, dated December 23, 1992, Governor Nelson requested the Commissioners to withdraw Boyd County from consideration.
If the site were not withdrawn from consideration by January 8, 1993, Governor Nelson stated that he would request the Nebraska Attorney General to take the matter to court to determine if there were legal avenues that would allow the State to prevent siting of the facility in Boyd County.
The Governor filed a lawsuit on January 13, 1993, when the Central Compact Commission refused to withdraw the site.
On January 22, 1993, the prospective disposal facility licensee was notified by the appropriate State regulatory agencies that its license application might be denied.
A public hearing was held on March 17, 1993, and a contested case hearing is scheduled for October 4, 1993.
In California, progress has been delayed by two related issues.
The Bureau of Land Management (BLM) has control of the site, and the issue facing the State is how to acquire title to the Federally-owned land.
Before construction of the facility can begin, the site must be transferred from Federal ownership to the State.
Transfer of the land from BLM to the State of California has been delayed because the State Lands Commission (SLC), an independent State agency, refused, in mid-1991, to proceed with the transfer process.
SLC is concerned about safety iscues related to the facility and is concerned that the State taxpayers might have the responsibility for any site remediation.
In a letter to California Governor Pete Wilson, dated August 11, 1993, the Secretary of the Interior recommended that a formal hearing be held, focusing on the issue of migration of radionuclides from the site.
The Secretary stated that the hearing is needed to carry out his responsibilities under Federal law.
The results of the hearing would be used by the Secretary in making a decision to transfer the land.
A final decision is also contingent on the designation of a critical habitat for the desert tortoise, an endangered species under the Endangered Species Act.
In his letter of August 11, 1993, the Secretary of the Interior stated that the Fish and Wildlife Service will shortly release its draft critical habitat designation, with the final designation scheduled for the end of the year.
The Secretary further stated that being in the tortoise's critical habitat would not necessarily prevent the California project from going forward.
However, it would require a more careful look at the project's
l BP20 (09/93) impacts on the tortoise and suitable mitigation measures.
The project may be further impacted by a suit, currently before the California Supreme Court, concerning a California Senate Rules Committee decision requiring a more extensive adjudicatory hearing.
Since no new LLW disposal facilities were operational by
)
January 1, 1993, and the compact commissions that control the existing LLW disposal sites have either closed their facilities or set conditions on receiving LLW from outside their regional compacts, some licensees who generate LLW have been forced to i
store their LLW on-site, until disposal capacity is available.
j Nearly all the Governors' Certifications submitted to meet the i
1990 milestone of the LLRWPAA indicated that the State planned on j
interim storage by waste generators during the 1993 through 1996 j
period.
However, since the South Carolina facility is to be available to many generators until July 1, 1994, there will be a mixed pattern of disposal and storage during this time period.
l Regulatory Actions:
Because of some States' and compacts' slow progress in meeting l
the January 1, 1996, milestone of the LLRWPAA, the Commission is becoming concerned about on-site storage of LLW.
Although the public health and safety can be adequately protected if LLW is stored, the public health and safety will be enhanced by disposal, rather than long-term, indefinite, storage of LLW.
Disposal of LLW in a limited number of facilities licensed under existing regulations (10 CFR Part 61 or compatible Agreement State regulations) will provide better protection of the public health and safety and environment than storage at hundreds of sites around the country.
Permanent disposal of LLW has always been the preferred option for managing LLW, as reflected in the LLRWPAA.
Because of these concerns and as a result of the Commission's consideration of the staff's analysis in SECY-91-306, " Analysis of Comments Received on Title-Transfer and Possession Provisions of the Low-Level Radioactive Waste Policy Amendments Act of 1985," ths U.S. Nuclear Regulatory Commission has proposed to amend its regulations, to establish license condition requirements for on-site storage of LLW, by licensees, after January 1, 1996.
In this proposed rulemaking, the Commission has restated and emphasized its position that it will not look favorably upon on-site storage of LLW by generators after January 1, 1996.
Under the proposed amendments, on-site storage of LLW would not be permitted after January 1, 1996 (other than reasonable short-term storage necessary for decay or for collection or consolidation for shipment off-site), unless the licensee could document that it has exhausted other reasonable waste management options.
Such options include taking all reasonable steps to contract, either directly or through the State, for disposal of the waste.
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BP20 (09/93)
This proposed rulemaking would supplement, but not supersede, the existing regulatory framework applicable to storage of LLW, and the conditions in themselves would not authorize on-site storage.
On-site storage of LLW at reactors would continue to be subject to 10-CFR-50.59 evaluations (which allow licensees to store, provided there is no outstanding safety issue), as well as all other regulatory requirements currently in place.
Licensees should continue to use appropriate regulatory guidance for on-site storage of LLW.
t Current Status:
A proposed rulemaking was published in the Federal Recister on February 2, 1993.
The public comment period expired on April 5, 1993.
The staff is currently summarizing and analyzing the comments received.
Final action on this rulemaking is expected in late 1993.
Highlights of this media briefing background paper can be found in the attachment.
Contact:
Michael Bell, Chief, Low-Level Waste-Management Branch, Division of Low-Level Waste Management and Decommissioning, Office of i
Nuclear Material Safety and Safeguards, U.S. NRC, Washington, DC 20555 (301) 504-2553.
=
,~
.. ~
LOW-LEVEL RADIOACTIVE WASTE COMPACT STATUS MARCH 1992 U
M M TES'''
+ 16% Nedonal LLW (10 Steles)
- NY to host site - 7% Nedonal LLW-SLB bonned g
- WA is host State I
- MA to host site-3% Nedonal LLW-SLB bonned
- RAE b host she - <m W RW-M W
= 0% Nedones11W M
\\
8% NadonalLLW
- W to host site - M N RW-SLB W
- stb t=nned
- ^
- w. a R W h stornoe HI 'O r
NN wounEAST yy MN VT
\\
\\
- NJ and CTare party States ID
- NJ and CTselected as host W3 States 1
- 8% Nedonsi LLW t
k gg Y}
J h,-MA
= Disposal tedrictogy to be f
[
ON D - RI determhed by host SIsles UT ND I
g MD i
CA I
DC CENTRAL MIDWEST MMN M
WY
- IL is host State
- PA is host State 1
8% NeGonal LLW
= 18% Nedonal LLW q
L
, NE IL
- EMAGV.SLB bonned
- EMAGV. stb W SOUTHWESTERN pgy CO I
I VA
- CAis host Stat' i
f L
+7% Nedonal RW NC
[TH
- SLs SOUTHEAST y,
1 OK AR 89
- SC currert host State using SLB ROCU MOUNT MS GA
. NC oelected as next host State j
i
(
AL a NVcurrent host State
~_
i
- 21% Nedonal LLW
- Northwest Compact esgitte g
- EMAGVdiosen by NC-SLB bonned b accept LLWIrom aAer 992 TEXAS CENTRAL i
- <1% Nedonal LLW
- m to host sAs
- NE selected as W Siete
.g
=4% Naponal LLW C PR
- 4% NedonalLLW
- BGCC. stb
- EMAGV. SLB bemed
'_t wuwwso-y p6ote: National LLW wohane for 1991 = 1.4 nellflon cube feet tNeposed u,,,,,,,,,,,,,,, %
EMAGV = Earth neounded above grade voedt source: Ome..c st no.,..- --. NRC BGCC = belour ground concrete canisters l
l
BP20 (09/93)
Actual and Estimated Dates for Completing Steps in Facility Development (Estimated Dates Obtained from Compacts / States)
Submit Select License Operate Comoact/ Host State Site Acolication Facility Appalachian / Pennsylvania Early 1997 Early 1997 Mid-1999 Central / Nebraska Dec 1989 Jul 1990 Unscheduled Central Midwest / Illinois Unscheduled Unscheduled Unscheduled Midwest /0hio Unscheduled Unscheduled Unscheduled Northeast / Connecticut &
Unscheduled Unscheduled Unscheduled New Jersey 1994 Mid-1995 Late 1997 Southeast / North Carolina End 1993 Dec 1993 Jan 1996 Southwest / California Mar 1988 Dec 1989 Jul 1994 Unaffiliated States Maine (See note.)
1993 Jan 1994 End 1996 Massachusetts Jun 1994 Nov 1994 Late 1996 New York 1995/1996 1998/1999 2001 Texas Aug 1991 Mar 1992 Jun 1996 Vermont (See note.)
Feb 1994 Oct 1997 Aug 1999 Note: The dates provided for Maine and Vermont reflect the schedule that existed before enactment of Texas Compact legislation.
i
^
}
BP20 (09/93)
{
HIGELIGHTS STATE COMPLIANCE WITH 1993 MILESTONE AND 1996 LEGISLATIVE OBJECTIVE OF THE LOW-LEVEL i
RADIOACTIVE WASTE POLICY AMENDMENTS ACT OF 1985 LLRWPAA established milestones, incentives, and penalties for States to develop new low-level radioactive waste (LLW) i disposal facilities.
Milestones were established in 1986, 1988, 1990, 1992, ee 1993, and 1996.
{
Waste disposal surcharges and take-title and possession ee provisions were penalties for failure to comply.
l l
Partial rebate of surcharges to States provided i
ee incentive in the form of financial assistance.
l ee U.S.
Supreme Court has held the 1995 take-title l
provision to be unconstitutional.
i Majority of States met the first three milestones.
e Only four States (California, Illinois, Nebraska, and Texas) met the 1992 milestone, and only California and North I
Carolina are scheduled to meet the 1996 legislative j
objective of the LLRWPAA.
On-site storage of LLW at some generator sites is occurring because of lack of access to dispost.:1 facilities, especially in the States of Michigan (since November 1990), New Hampshire (Seabrook Nuclear Power Plant since October 1990),
and Rhode Island, and the District of Columbia and Puerto Rico; and more recently the States of the Central Compact.
Existing U.S. Nuclear Regulatory Commission guidance, in e
conjunction with current regulations, provides the l
regulatory and licensing framework for LLW storage.
Although public health and safety can be protected if LLW is stored, public health and safety will be er.hanced by disposal.
NRC does not look favorably upon on-site storage of LLW.
Commission has published a proposed rulemaking that would e
establish criteria for on-site storage of LLW after January 1,
1996.
Licensees would have to exhaust all other reasonable waste management options before storing LLW on-site.
Options include contracting, either directly or indirectly, through the State, for disposal.
1 Proposed rulemaking would supplement, but not supersede, o
existing regulatory. framework.
Conditions of proposed i
rulemaking, in themselves, would not authorize on-site storage.
j Attachment j
1
BP21 (09/93)
INCIDENT INVESTIGATION TEAMS Backcround:
The Incident Investigation Program (IIP) ensures that significant operational events are investigated by the NRC in a systematic, technically sound, and timely manner, to gather information pertaining to the probable root causes of the event, including any NRC contributions or lapses.
The investigation of all significant operational events involving non-reactor and reactor activities licensed by the NRC are within the scope of this program.
By focusing its efforts on uncovering the causes of operating events and on identifying associated corrective actions, the incident investigation process provides for a more complete technical and regulatory understanding of significant events.
This is a major contribution to nuclear safety.
Within the program, appropriate feedback is solicited from the NRC, the nuclear industry, and the public regarding the lessons learned from the events.
Incident Investiaation Team:
For an event of high safety significance, an Incident Investigation Team (IIT) is established by the EDO.
An IIT performs the single NRC incident investigation and reports directly to the EDO.
It is compcsed of headquarters and regional staff and, for events involving power reactors, the team may also include industry representatives.
To maintain independence, the team leader and other technical experts are chosen from those who have had no significant involvement with licensing and inspection activities at the facility at which the event occurred.
For an event of less safety significance, the Regional Administrator appoints an Augmented Inspection Team (AIT).
This team reports directly to the Regional Administrator.
Its members come from the regional staff supplemented by headquarters personnel and, in some cases, by personnel from other regions.
The team investigations are presently governed by two NRC documents:
(1) Management Directive 8.3, "NRC Incident Investigation Program," and (2) NUREG-1303, " Incident Investigation Manual."
The IIT begins its investigation as soon as practicable after the facility has been placed in a safe, secure, and stable condition.
If there is an NRC incident response, the investigation begins after it is deactivated.
BP21 (09/93)
For all IITs a Confirmatory Action Letter (CAL) will be issued as appropriate to the licensee requiring that, within the constraints of plant safety, suspected relevant failed components or systems be quarantined for troubleshooting.
The CAL also ensures that the facility is maintained in a safe shutdown condition until concurrence is received from the NRC to restart.
The IIT investigation emphasizes fact-finding and determination of probable causes for the event.
As the investigation progresses, the IIT issues interim reports outlining the status and plans, and adds relevant new information related to the investigation.
The scope of the investigation is established to ensure that the event is clearly understood, that relevant facts and circumstances are identified and collected, and that probable cause(s) and contributing cause(s) are identified and substantiated by the evidence and information collected.
The team also considers whether the licensee and the NRC took timely and adequate action or if they contributed to the cause before or during the event.
The team also investigates event precursors, event chronology, systems response, human factors considerations, equipment performance, safety significance, and radiological considerations.
The IIT prepares a final report listing detailed findings and conclusions and sends copies to the Commission and the EDO within about 45 days from the time the team is activated.
After receiving the final report, the EDO normally schedules a meeting so the IIT can brief the Commission on the investigation.
Information contained in the report is not released to the public until a copy of the report ic rie:ed in the Public Document Room (PDR).
Following the CommiL_,m. briefing, a copy of the final report is transmitted to the licensee and NRC staff for their review and comment.
In addition to NRC staff and licensee review and comment on the IIT report, the EDO identifies from the investigation findings generic and plant-specific staff actions for evaluation and resolution as well as actions that are safety significant and that warrant additional attention or action.
The EDO assigns these actions to the appropriate NRC office.
Office Directors are asked to issue a status report on the disposition of each assigned staff action.
AEOD will prepare annual closcout reports to identify and document actions taken to close out each assigned staff action item.
In addition, the resolution of each IIT finding is subject to an independent assessment by AEOD of its adequacy and completeness.
Contact:
Stuart Rubin, Chief, Diagnostic Evaluation and Incident Investigation Branch, Division of Operational Assessment, Office for Analysis and Evaluation of Operational Data, U.
S.
Nuclear Regulatory Commission, Washington, DC 20555, (301) 492-4147.
BP21 (09/93)
HIGHLIGHTS OF INCIDENT INVESTIGATION TEAMS The Incident Investigation Program (IIP) ensures that NRC investigations of significant events are timely, thorough, well coordinated, and formally administered.
The scope of the IIP covers the investigation of significant operational events involving non-reactor and reactor activities licensed by the NRC.
The IIP process contributes to nuclear safety by providing for a more complete technical and regulatory understanding of significant events.
Appropriate feedback regarding what is learned from the events is provided to the NRC, the industry, and the j
public.
INCIDENT INVESTIGATION PROGRAM:
For an event of potentially major significance, an Incident Investigation Team (IIT) is established by the EDO.
The investigation of less significant operational events may be conducted by an Augmented Inspection Team (AIT).
The IIT begins its investigation as soon as practicable after the facility has been placed in a safe, secure, and stable condition.
The IIT issues interim status reports at appropriate intervals.
The investigation performed by an IIT emphasizes fact-finding and determination of probable causes for a significant operational event.
The IIT prepares a final report and transmits it to the Commission and the EDO.
Following NRC staff and licensee review and comment on the IIT report, the EDO identifies and assigns NRC Office responsibility for generic and plant-specific staff actions.
j Each resolution of an IIT finding is subject to
=
independent assessment by AEOD as to its adequacy and completeness.
The IIT investigation is governed by two NRC Documents:
(1) Management Directive 8.3, "NRC Incident Investigation Program," and (2) NUREG-1303, " Incident Investigation Manual."
BP22 (09/93)
ROSEMOUNT TRANSMITTERS
Background:
During the late 1980's several utilities were having problems with Rosemount transmitters failing to calibrate, or failing response time testing.
A notification pursuant to 10 CFR Part 21 was issued, and NRC started an investigation.
It was determined that the problem was caused when the transmitters had a very slow leakage of fill oil.
This reduced the travel of the diaphragm within the transmitter, and led to failure.
On April 21, 1989, the NRC issued Information Notice (IN) 89-42, " Failure of Rosemount Models 1153 and 1154 Transmitters," to alert the industry to the reported failures of Models 1153 and 1154 pressure and differential pressure transmitters manufactured by Rosemount Inc.
Rosemount investigated the cause of the failures and confirmed that the failure mode was a gradual loss of fill-oil from the sealed sensing module of the transmitter.
At that time it appeared that the failures were lot related, and that
" suspect lots" had a higher failure rate than non-suspect lots.
On March 9, 1990, the NRC issued Bulletin 90-01.
This Bulletin requested that licensees promptly take corrective actions for Model 1153 Series B, Model 1153 Series D, and Model 1154 transmitters manufactured by Rosemount that were in the suspect lots, and therefore had the potential for leaking fill-oil.
SUPPLEMENT 1 to BULLETIN NO. 90-01:
Since that time, there has been additional informacion collected and analyzed, resulting in the writing of an updated supplement to the original Bulletin.
During the licensee response period to the original Bulletin, NUMARC surveyed all utilities to collect data on all installed Rosemount Model 1153 and 1154 transmitters, and on Rosemount Model 1151 and 1152 transmitters installed in safety-related systems.
The staff, assisted by Brookhaven National Laboratory, reviewed the data collected by NUMARC. The failed transmitters were sorted by operating pressure and time-in-service.
In evaluating this issue, the staff confirmed a i
relationship, as had been previously found by Rosemount and NUMARC, between the likelihood or failure and operating pressure, time-in-service, and the suspect and nonsuspect lot classifications.
A high operating pressure was the most dominant factor leading to a loss of fill-oil, with time-in-service also j
being a significant factor.
Transmitters with an operating 1
pressure of greater than 1500 psi had the highest failure rate, and those with an operating pressure of between 1500 and 500 psi also had an increased failure rate.
Second among these factors was time-in-service, with those transmitters having been in service for less than 60,000 psi-months exhibiting higher failure rates than transmitters that had been in service for more than 60,000 psi-months.
l
BP22 (09/93)
On April 7, 1992, a proposed Supplement 1 to Bulletin 90-01, was published in the Federal Recister.
Twelve replies to this notice were received.
The comments received primarily concerned the scope of coverage for the transmitters to be addressed and clarification of the exact nature of requested actions.
On July 23, 1992, a public meeting was held to discuss the comments l
received and their disposition.
On September 8, 1992, the Committee to Review Generic Communications (CRGR) reviewed the proposed supplement, including the disposition by the staff of comments received, and recommended changes which the staff incorporated into the supplement.
The Bulletin Supplement was issued on December 22, 1992.
The Supplement requests utilities to review the information for applicability to their facilities, perform testing on a transmitter commensurate with its importance to safety and demonstrated failure rate, and modify, as appropriate, their i
actions and enhanced surveillance programs Regulation:
The NRC issued General Design Criterion (GDC) 21, " Protection System Reliability and Testability," in Appendix A to Part 50 of Title 10 of the Code of Federal Reculations (10 CFR Part 50) to 4
require the protection system to be designed with high functional reliability and with sufficient capability to allow periodic testing of its functioning when the reactor is operating.
The
)
NRC established this requirement to ensure that the licensee can readily detect failures of subcomponents and subsystems within the protection system and can readily detect loss of the required protection system redundancy when it occurs.
In 10 CFR 50.55a(h), the NRC requires that protection systems meet the Institute of Electrical and Electronics Engineers Standard,
" Criteria for Protection Systems for Nuclear Power Generating Stations" (IEEE-279).
In IEEE-279, it is stated that means shall be provided for checking, with a high degree of confidence, the operational availability of each system input sensor during reactor operation.
To achieve a high functional reliability, a transmitter must have a low probability of failing while it is operating.
When failures occur, they should be readily detectable, commensurate with the safety function, while the transmitter is operating.
Upon reviewing the analyses, evaluations, and historical data on the loss of fill-oil, the staff concluded that actions requested by the previous bulletin were insufficient to ensure the transmitters achieve the desired high functional reliability.
Current Status:
The Bulletin Supplement, issued on December 22, 1992, also requested that within 60 days after receipt of the bulletin, the licensee p:; ovide a response that includes a statement whether the
BP22 (09/93) licensee will take the actions requested.
This statement includes a list of the specific actions that the licensee will complete and the schedule for completing actions.
When completed, the licensee is requested to provide a statement confirming that the requested actions have been completed.
If the licensee is not completing all actions, a statement is requested identifying those actions that the licensee is not taking and an evaluation which provides the bases for not taking the requested actions.
All responses have been received and review of the responses is ongoing in I&C branch.
Additionally, a review group was formed to conduct an in depth review and evaluation of the Rosemount transmitter issue to determine whether the agency should require licensees to take additional action beyond that specified in the Bulletin Supplement.
The preliminary results of this review indicates that the scope and actions specified in NRC Bulletin 90-01, Supplement 1 are considered appropriate.
Current failure data for Rosemount transmitters confirms that no changes are warranted to the Supplement and licensees need take no additional actions.
Several recommendations which may result from the review include the following:
(1)
Issuing a temporary instruction (TI) to evaluate licensee actions, in response to the supplement, which address the Rosemount transmitter loss-of-fill-oil issue.
(2)
Continuing the periodic dialogue with Rosemount through meetings, telephone discussions and inspections to ascertain performance of the different model transmitters with regard to loss-of-fill-oil.
(3)
Reviewing NPRDS failure data to identify whether a significant increase in the number of reported fill-oil failures has occurred.
)
(4)
Holding a management meeting with NUMARC to discuss several areas of possible lessons learned associated with the i
Rosemount transmitter loss-of-fill-oil issue.
Contact:
Technical
Contact:
Deirdre Spaulding, NRR/HICB (301) 504-2928 Lead Project Manager:
Ngoc Le, NRR (301) 504-1458 l
i I
BP23 (09/93) j HARASSMENT AND INTIMIDATION IN THE NUCLEAR INDUSTRY
Background:
1 The NRC has responsibility for regulating the operation of approximately 110 nuclear power plants and more than 8,000 materials licensees.
The magnitude of the licensed activities is so extensive that the NRC can inspect only a fraction of them.
]
Although the NRC's program for ensuring adequate protection is not structured to be dependent upon allegations of safety deficiencies, the NRC will never have the knowledge possessed by the thousands of employees in the nuclear industry.
The NRC, therefore, expects employees in the nuclear industry to be free to raise potential safety issues.
Recognizing that it is not the NRC, but licensees who have the first responsibility for safety, it is not enough for employees to feel free to come directly to the NRC.
Employees must feel 1
free to raise potential safety issues to their management.
Over the years the NRC, the regulated industry, and the public have benefitted from the issues raised by employees of licensees and their contractors.
If employees are subjected to retaliation by their employers for reporting safety concerns, there is a potential " chilling effect" for both allegers and co-workers who may have additional safety concerns to report.
The current regulatory process seeks to provide protection against retaliation for employees engaged in protected activities, e.g.,
raising of potential safety concerns to a licensee or the NRC.
Discrimination against an employee for engaging in protected activities is prohibited by the Commission's regulations (10 CFR 19.20, 30.7, 40.7, 50.7, 60.9, 61.9, 70.7, and 72.10).
Discrimination for purposes of the Commission's regulations includes discharge and other actions that relate to compensation or terms, conditions, and privileges of employment.
A licensee is subject to enforcement action under the Atomic Energy Act (AEA) of 1954, as amended, for violations of these prohibitions by the licensee or its contractors and subcontractors.
The AEA does not give the NRC the authority to provide a personal remedy, such as reinstatement or back pay, to an employee who has been subject to discrimination.
The NRC's authority is directed at the licensee.
NRC regulations and Federal statutes allow enforcement actions against licensees that could include denying, revoking, or suspending a license; imposing civil penalties; and initiating criminal proceedings.
An employee who believes that discrimination has occurred may seek a personal remedy by filing a complaint within 180 days with the Department of Labor (DOL) in accordance with Section 211 of
1 l
BP23 (09/93) the Energy Reorganization Act (ERA) of 1974, as amended.
However, the DOL process for providing personal remedies for those retaliated against (and the NRC process for taking enforcement action against the licensees who have retaliated) is time consuming.
Although the NRu can initiate an investigation at any time during the DOL process, the agencies agreed in a 1982 Memorandum of Understanding that the NRC would not normally conduct a parallel investigation.
To a large degree, the NRC relies on the DOL for investigating allegations of discrimination.
However, the NRC does respond to technical aspects of allegations.
Between October 1988 and April 1993, the NRC received a total of 609 retaliation complaints and initiated full-scale investigations for 44 of them.
Three hundred-sixty nine of these complaints were also filed with DOL.
Based on the complaints received, seven NRC enforcement actions were taken against licensees during this period and others are pending.
On July 9, 1993, the NRC Inspector General (IG) issued a report that addressed the staff's handling of retaliation complaints received from whistleblowers employed by NRC licensees.
The IG, after interviewing various whistleblowers and NRC staff members, concluded that the NRC process for handling allegations of retaliation does not provide an adequate level of protection for whistleblowers.
On July 15, 1993, the NRC testified in a hearing conducted by the Senate Subcommittee on Clean Air and Nuclear Regulation on the NRC's handling of intimidation and harassment allegations by employees within the nuclear industry.
Current Developments:
On July 6, 1993, the Executive Director for Operations established a Review Team for Reassessment of the NRC Program for Protecting Allegers Against Retaliation.
The Review Team is to determine whether the Commission has taken sufficient steps within its authority to create an atmosphere within the regulated community where individuals with safety concerns feel free to engage in protected activities without fear of retaliation.
Because Section 211 of the ERA was recently amended as part of the Energy Policy Act of 1992, Pub.
L.
102-486, this review is to focus primarily on the existing statutory framework, in which the DOL provides the personal remedy to the employee, and NRC is responsible for regulating licensees such that licensees will foster an atmosphere where individuals will be encouraged to come forward with safety information without fear of retaliation.
In accordance with its charter, the Review Team is to consider:
BP23 (09/93)
(a)
Whether the NRC has taken sufficient action through issuance of regulations, policy statements, and inspections to ensure i
that NRC licensees encourage their employees and contractors to raise safety concerns without fear of reprisal; I
(b)
Whether the current NRC process for handling allegations is appropriate from the perspective of allegers' feeling free to bring safety concerns to the NRC; and (c) Where discrimination may have occurred --
(1)
Whether there are NRC actions that can assist in a speedier resolution of issues within the DOL process; (2)
Whether NRC should be F]re proactive in conducting investigations during tne pendency of DOL proceedings; (3)
Whether the NRC takes sufficient follow up action to determine if the licensee has taken action to remove the potential chilling effect arising from the discrimination; (4)
Whether the NRC can and should use civil penalties and orders more vigorously to emphasize the need for licensees actively to encourage employees to raise safety concerns without fear of discrimination; and (5)
Whether the NRC can and should use orders and demands for information more vigorously, where individuals are found to have caused discrimination; and (d)
Whether the NRC is sufficiently proactive in cases where employees raise concerns with the NRC and express fear that they may become subject to retaliation for raising safety concerns.
On August 2, 1993 (58 FR 41108), the Review Team requested public comment on whistleblower protection issues, and provided a list of questions to aid in focusing discussion on these issues.
On August 25, 1993, the Review Team held a public meeting with attorneys of employees seeking remedies pursuant to Section 211 of the ERA to obtain views on whether the NRC has taken sufficient steps within its statutory authority to create an atmosphere within licensees' organizations where individuals feel l
free to engage in protected activities without fear of retaliation.
The Review Team held a similar public meeting with industry attorneys on process and remedies pursuant to Section 211 of the l
ERA on August 26, 1993.
The Review Team intends to hold public meetings at four locations 4
(Bay City, Texas; New London, Connecticut; Phoenix, Arizona; and Cleveland, Tennessee) to discuss current whistleblower protection activities.
The NRC will invite participation in these meetings i
BP23 (09/93) by any interested persons, including individuals who have made safety allegations, other employees, licensee and contractor representatives, and the public.
This action is intended to assist the NRC in evaluating current whistleblower protection I
activities and in recommending improvements in the regulatory process.
A report from the Review Team is due to the Commission in January 1994.
Contact:
James Lieberman, Director, Office of Enforcement US NRC, Washington, DC 20555.
(301) 504-2741.
i i
j 1
1 1
l
BP23 (09/93)
HIGHLIGHTS OF H & I ISSUES IN THE NUCLEAR INDUSTRY NRC expects employees in nuclear industry to be free to raise potential safety concerns.
If employees are subjected to retaliation by their employers for reporting safety concerns, there is a potenti.al
" chilling effect" for both whistleblowers and co-workers who may have additional safety concerns to report.
Discrimination against an employee for engaging in protected activities is prohibited by the Commission's regulations (10 CFR 19.20, 30.7, 40.7, 50.7, 60.9, 61.9, 70.7, and 72.10).
Discrimination for purposes of the Commission's regulations includes discharge and other actions that relate to compensation or terms, conditions, and privileges of employment.
AEA doesn't give NRC authority to provide personal remedy (reinstatement or back pay).
NRC's authority is directed at licensee--NRC could deny, revoke, or srapend a license; impose civil penalties; and refer to DOJ for criminal proceedings.
DOL has authority to provide personal remedy, when appropri-ate--but process has been time consuming.
In accordance with 1982 MOU with DOL, agencies agreed that NRC would not normally conduct parallel investigation.
While NRC responds to technical aspects of allegations, NRC, to a large extent, relies on the DOL for investigating allegations of discrimination.
NRC IG issued report on July 9, 1993, that concluded that the NRC process for handling allegations of retaliation does not provide an adequate level of protection for whistleblowers.
NRC testified in hearing en July 15, 1993, to the Senate Subcommittee on Clean Air and Nuclear Regulation on the NRC's handling of intimidation and harassment allegations by employees within the nuclear industry.
EDO established Review Team on July 6, 1993, for reassessment of the NRC program for protecting allegers against retaliation.
Review Team is soliciting public comment and will be conducting public meetings to obtain information to assist in evaluating current program and process.
Review Team response to Commission is due January 14, 1994.
BP24 (09/93)
DECOMMISSIONING NUCLEAR POWER PLANTS Backaround:
Recently, several licensees announced their decisions to permanently cease power operation at their_ nuclear power generating facilities.
The reasons for their decisions include both economic and technical considerations.
Thus, these facilities and several others entered the decommissioning process earlier than originally anticipated.
Recent decommissioning highlights are presented in Table 1 and a status for decommissioning plants is presented in Table 2.
Decommissionina:
Title 10 of the Code of Federal Reaulations, Section 50.2 (10 CFR 50.2), defines decommissioning to mean removal of a facility safely from service and reduction of residual radioactivity to a level that permits release of the property for unrestricted use and termination of the license.
Decommissioning can involve three different methods:
DECON, SAFSTOR, or ENTOMB.
Under DECON, equipment, structures and portions of a facility and site containing radioactive contaminants are removed or decontaminated to a level that permits the property to be released for unrestricted use relatively soon (a few years) after cessation of operations.
Under SAFSTOR, often considered " delayed DECON," a nuclear facility is placed and maintained in a condition that allows safe storage for a period of time to allow radioactive decay to reduce radiation levels at the facility.
The plant is subsequently decontaminated to levels that permit release of the site for unrestricted 1se.
Under ENTOMB, radioactive contaminants are encased in a structurally long-lived material such as concrete and the entombed structure is appropriately maintained and monitored until the radioactivity decays to a level permitting unrestricted release of the property.
The ENTOMB alternative has generally not been selected by nuclear power plant licensees.
These three decommissioning alternatives are not totally discrete actions inasmuch as some decontamination and other preparatory activities such as component disassembly may be performed under both the SAFSTOR and ENTOMB alternatives.
Finally, to be acceptable, the method selected must provide for completion of decommissioning within 60 years and a time beyond 60 years will be considered only when necessary to protect public health and safety, in accordance with NRC regulations.
Reculations:
The procedure for decommissioning a nuclear power plant is set out principally in 10 CFR 50.75, 50.82, 51.53, and 51.95.
(An underlying, original assumption embodied in the current regulations is that decommissioning would occur after the time of
BP24 (09/93) expiration of the facility operating license.)
The formal process begins with the filing of an application by the licensee, normally after the plant has ceased permanent operations, for authority to surrender its license and to decommission the facility.
The regulations contemplate, however, that the actual process begins even before the licensee makes a decision to permanently cease operations.
Five years before the licensee expects to end operation of the plant, the licensee is obligated to submit a preliminary decommissioning plan containing a cost estimate for decommissioning and an up-to-date assessment of the major technical factors that could affect planning for decommissioning.
Then, within two years following permanent cessation of operations, but no later than one year prior to expiration of the license, a licensee must submit to the NRC an application for authority to surrender a license voluntarily and to decommission that facility, together with an environmental report covering the proposed decommissioning activities.
The i
application must also be accompanied, or preceded, by a proposed decommissioning plan that includes:
(1) A description of the decommissioning alternative chosen and activities involved.
(2) A financial plan showing a cost estimate for decommissioning, the amount of funds currently available for decommissioning, and plans for assuring the availability of adequate funds for completion of decommissioning.
The NRC reviews the decommissioning plan, prepares an environmental impact statement (EIS) or environmental assessment (EA), as appropriate, and gives notice to interested persons.
If the NRC finds the proposed decommissioning plan to be satisfactory (that is, the plan is_in accordance with the regulations and is not inimical to the common defense and security or the health and safety of the public), the NRC issues a decommissioning order that approves the proposed decommissioning plan and authorizes decommissioning.
Upon completion of decommissioning activities, including the completion of a termination radiation survey, the NRC will issue an order that terminates the license.
Although not a necessary step in the decommissioning process, a licensee may request to have its authority to operate the facility rencvv Once an operating license is amended to remove a licensee's authority te operate, the facility license is called a possession only license (POL).
The POL serves two principal purposes.
It confirms the nonoperating status of the plant and more significantly, it provides a basis to remove requirements that are not necessary for a plant that will no longer operate.
Thus, a POL provides a basis for a licensee to begin eliminating 1
personnel, equipment, and activities which are unnecessary to maintain a permanently shut down plant.
These actions can be
l BP24 (09/93) taken as long as they do not affect the public health and safety, or materially affect costs and/or methods for decommissioning the facility.
Currently, licensees of permanently shut down plants often request regulatory relief from the NRC in a number of areas where compliance with operating reactor requirements is unnecessary to ensure public health and safety.
The NRC reviews each relief request individually.
Rulemaking changes may be initiated to reduce expenditure of both NRC and licensee resources required to process each regulatory relief request, while still ensuring the protection of the public health and safety.
Prematurelv Shutdown Plants:
Subsequent to the publication of the final decommissioning rule six power reactor facilities have been shut down prematurely:
the Fort St. Vrain Nuclear Generating Station, the Shoreham Nuclear Power Station, the Rancho Seco Nuclear Generating Station, the Yankee Rowe Nuclear Station, San Onofre Nuclear Generating Station, Unit 1, and Trojan Nuclear Plant.
Three Mile Island Nuclear Station, Unit 2 also ceased operation following the March 28, 1979, accident.
Current Status:
During CY-1992, the NRC completed the review efforts necessary to support decommissioning of the Shoreham and Fort St. Vrain facilities.
On June 11, 1992, the NRC issued an order to the current Shoreham licensee, Long Island Power Authority, approving the Shoreham decommissioning plan.
The Long Island Power Authority is in the process of dismantling that facility.
On November 23, 1992, the NRC issued an order approving the Fort St.
Vrain decommissioning plan and dismantlement activities are now ongoing.
Issuance of the Rancho Seco decommissioning plan has been delayed due to the Commission remanding to the Atomic Safety Licensing Board three issues (loss of offsite power, decommissioning funding plan, and decommissioning environmental i
assessment) raised by the Environmental and Resources Conservation Organization (ECO).
On June 16, 1993, the NRC staff issued its safety evaluation and its environmental assessment on the proposed Rancho Seco decommissioning plan.
This matter is now before the Licensing Board.
During fiscal year 1992, Yankee Atomic Electric Company and Southern California Edison Company announced their decisions to
]
prematurely shut down and decommission the Yankee-Rowe and San Onofre 1 facilities, respectively.
In January 1993, the Portland General Electric Company decided to terminate operations at the j
Trojan plant.
All three of these facilities now have been i
permanently shut down.
The NRC is working to support the decommissioning of these facilities, j
BP24 (09/93)
In January 1993, the Commission issued guidance regarding activities which may be permitted prior to approval of a decommissioning plan.
Licensees of plants which have possession only licenses or shutdown orders should be allowed to undertake any decommissioning activity that does not (a) foreclose the release of the site for possible unrestricted use, (b) significantly increase decommissioning costs, (c) cause any significant environmental impact not previously reviewed, or (d) violate the terms of the existing license.
Also, licensees may be permitted to use their decommissioning funds for approved decommissioning activities, notwithstanding the fact that their decommissioning plans have not yet been approved by the NRC.
In accordance with this guidance, Yankee Atomic Electric Company is conducting early removal of the four steam generators, the pressurizer, and reactor vessel internals from the Yankee Rowe plant.
In a July 15, 1993, letter to Yankee, the staff stated that it has no objection to these activities.
Contact:
Seymour H. Weiss, Non-Power Reactors & Decommissioning Project, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, (301) 504-2170.
1 i
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BP24 (09/93)
TABLE 1 DECOMMISSIONING HIGHLIGHTS FORT ST. VRAIN
- August 18, 1989, plant was permanently shut down because of failure of the control rod drives and degradation of the steam generator ring header.
- May 21, 1991, license is amended to a POL status.
1
- June 11, 1992, all fuel placed in an onsite independent spent fuel storage installation (ISFSI).
- November 23, 1992, the Commission issued an order approving licensee decommissioning plan.
- September 1, 1993, the Prestressed Concrete Reactor Vessel top head has been successfully removed.
SHOREHAM
- June 28, 1989, licensee's shareholders approve an agreement with the New York state to not operate.the facility.
- August 24, 1989, reactor vessel defueled.
- June 14, 1991, license is amended to a POL status.
February 29, 1992, the license was transferred to the Long Island Power Authority for decommissioning.
- June 11, 1992, Commission issued an order approving licensee decommissioning plan.
September 1, 1993, decommissioning of the facility is 75 percent complete.
September 1993, transfer of fuel to Limerick began.
Fuel transfer is to be completed by September 1994.
RANCHO SECO
- June 7, 1989, plant shut down as a result of approval by voters of a non-binding referendum prohibiting the licensee from operating the facility.
- December 8, 1989, reactor vessel defueled.
- March 17, 1992, license is amended to a POL status.
- Active intervention on proposed decommissioning plan by Environmental and Resources Conservation Organization (ECO).
- June 3,
- 1993, U.S.
Court of Appeals (Ninth Circuit) rules in favor of the NRC in the matter of the issuance of the Ranch Seco POL.
- June 16, 1993, safety evaluation and environmental assessment of proposed decommissioning plan issued by NRC staff.
i BP24 (09/93)
YANKEE ROWE February 27, 1992, the licensee announced permanent cessation of operations because of inability to address uncertainties associated with the safety margin of the reactor vessel.
The vessel was previous 3v defueled.
- August 5, 1992, license is amended to a POL status.
- July 15, 1993, NRC staff states "no objection to early component removal activities."
THREE MILE ISLAND, UNIT 2
- March 28, 1979, accident occurred in the plant that caused permanent cessation of operations.
- January 30, 1990, reactor was defueled.
- August 12, 1993, processing of Accident-Generated Water is complete.
SAN ONOFRE, UNIT 1
- November 30, 1992, plant was permanently shut down rather than bring it into compliance with current NRC safety requirements.
- October 23, 1992, a POL license amendment was issued.
POL amendment became effective March 9, 1993, when the reactor vessel was certified as completely defueled.
TROJAN
- January 4, 1993, the licensee announced permanent cessation of operations.
- January 27, 1993, reactor was defueled.
- May 5, 1993, license is amended to POL status.
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TABLE 2 REACTOR DECOMMISSIONING STATUS SHUTDOWN POWER REACTORS DOCKET N0.
THERMAL SHUT PRESENT FUEL i
REACTOR POWER LOCATION DOWN STATUS ONSITE?
50-3 Indian Point-1 615 MW Buchanan 10/31/74 Possession Yes (PWR)
New York Only Lic.
50-10 Dresden 1 700 MW Morris 10/31/78 Possession Yes (BWR)
Illinois Only lic.
50-16 Fermi 1 200 MW Monroe Co.
9/22/72 SAFSTOR No (Fast Breeder)
12/9/63 SAFSTOR No (BWR)
- California Approved 50-29 Yankee Rowe 600 MW Franklin Co.
10/1/91 Possession Yes (PWR)
Hassachusetts Only Lic.
50-114 CVTR (Pressure 65 MW Parr Jan. 67 Byproduct No Tube,Heavywater)
S. Carolina Lic. (St.)
50-130 Pathfinder 190 MW Sioux 9/16/67 DECON No (Nuclear Superheat BWR)
- Falls S.D.
NRC Part 30 50-133 Humboldt Bay-3 200 MW Eureka 7/2/76 SAFSTOR Yes (BWR)
- California Approved 50-171 Peach Bottom 1 115 MW York Co.
10/31/74 SAFSTOR No (HTGR)
- Pennsylvania Approved 50-206 San Onofre 1 1347 MW San Clemente 11/30/92 Possession Yes (PWR)
California Only Lic.
50-267 Fort St. Vrain 842 MW Platteville 8/18/89 DECON Yes (HTGR)
California Only Lic.
l 50-320 Three Mlle 2772 MW Middletown 3/28/79 Shutdown No Island 2 (PWR)
Pennsylvania Defueled 50-322 Shoreham 2436 MW Suffolk Co.
6/28/89 DECON Yes i
(BWR)
Oregon Only Lic.
50-409 Lacrosse 165 MW Lacrosse 4/30/87 SAFSTOR Yes (BWR)
- Wisconsin Approved
- Project management assigned to NMSS
l BP25 (09/93)
BWR WATER LEVEL INSTRUMENT ERRORS DUE TO NONCONDENSIBLE GAS
Background:
During the past year, anomalies have been observed in reactor vessel water level indication at several boiling water reactors (BWRs) during controlled depressurization while commencing plant outages or following reactor trips.
These anomalies consisted of
" spiking" or " notching" of level indication, and in one instance, a sustained error in level indication.
The root cause of these level indication anomalies has been determined to be the effects of non-condensible gas in the condensate chamber and reference leg of " cold-leg" type water level instruments.
Under rapid depressurization conditions non-condensible gases could cause significant errors in the level indication.
Discussion of Technical Issue:
Cold reference leg water level instruments measure reactor vessel water level by measuring the differential pressure of two columns of water, the variable leg and the constant height reference leg.
The reference leg is maintained filled to a constant height of water by the condensate chamber.
Steam is condensed in the condensate chamber and keeps the reference leg full.
Excess condensate is returned to the vessel through the steam supply line of the condensate chamber.
Non-condensible gases, such as hydrogen and oxygen, formed by radiolysis in the reactor vessel, are present in the steam supplied to the condensate chamber.
The gases can collect in the condensate chamber and can accumulate to high partial pressures.
The non-condensible gases become dissolved in the water at the top of the reference leg.
For the dissolved gases to present a problem, they must be dissolved throughout the length of the reference leg.
There are three mechanisms that may result in this condition.
One mechanism is diffusion.
Diffusion is a very slow process, however, and it is thought to be unlikely to cause significant amounts of gas to move down the reference leg.
The second potential mechanism is thermal convection which could cause currents in the reference leg that carry the dissolved gas.
The feasibility of this mechanism is dependent upon the geometry of the reference-leg and the proximity of a heat source.
The third and most likely mechanism for dissolved gases to be carried down the reference leg is leakage in the valves and fittings at the instrumentation racks that would result in a flow path down the reference leg.
Very small, barely detectable leaks, less l
than 0.01 gpm, can cause significant amounts of gas to be transported throughout the instrument.
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BP25 (09/93)
Dissolved gases in the reference leg do not present a problem unless the instrument is depressurized.
When depressurized, the gases come out of solution and form bubbles that travel up the reference leg.
During slow depressurization, level indication l
has been seen to temporarily " spike" or " notch" while a bubble moves through the vertical sections of the piping.
Significant I
spiking may automatically actuate such systems as the primary containment isolation system (PCIS); this occurred at the Pilgrim plant.
After spiking, which is of short duration, the indicated water level returns to actual level.
Level spiking is of lesser safety significance.
Bubbling of the gases may eject a significant amount of water from the reference leg.
Loss of reference leg inventory will cause level indication to be erroneously high.
This occurred during a normal plant cooldown on January 21, 1993 at WNP-2, resulting in a 32 inch error in level indication that gradually recovered over a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
If the reactor is rapidly depressurized, as would occur during a design-basis loss-of-coolant (LOCA) or opening of the automatic depressurization system (ADS) valves, even larger errors in the level indication could result.
However, analyses presented by the industry indicated that significant errors would not be expected until the reactor is depressurized below approximately 450 psi.
The staff has taken several actions to address this problem.
The BWR Ohners Group (BWROG) Regulatory Response Group (RRG) was activated during July 1992, and the staff has met several times with the BWROG to discuss short-and long-term plans for resolving this concern.
The staff also issued Information Notice 92-54 in July 1992, Generic Letter 92-04 in August 1992, and Information Notice 93-27 in March 1993 to alert licensees to the potential problem and to request information concerning actions taken or planned by licensees in response to potenti;l errors in level indication.
The BWROG conducted a test program to support their efforts to resolve this issue.
The results of the BWROG reference leg de-gas test program confirmed that no significant errors in level indicaticn will occur until the reactor is depressurized below 450 psig, and that large errors in level indication are possible once the reactor is depressurized to lower pressures.
The staff received additional information from the'BWROG pertaining to reactor vessel water level instrumentation inaccuracies during normal depressurization due to the effects of noncondensible gas.
At the staff's request the BWROG submitted a report on May 20, 1993, discussing the impact of level errors on automatic safety system response and operator actions during transients and accidents initiated from reduced' pressure conditions during plant cooldown (Mode 3).
Based on this information, in addition to the January 21, 1993 WNP-2 event and data from the reference leg de-gas testi.sg that was conducted by the BWROG, the staff concluded that additional
t BP25 (09/93) short-term actions needed to be taken for protection against potential events occurring during normal cooldown.
On May 28, 1993, NRC Bulletin (NRCB) 93-03 was issued, in which the staff requested each BWR licensee to implement additional short term compensatory actions, and to implement a hardware modification to resolve this issue at the next cold shutdown after July 30, 1993.
Current Status:
The staff has received responses to NRC Bulletin 93-03
" Resolution of Issues Related to Reactor Vessel Water Level Instrumentation," from all licensees.
Three licensees submitted responses to NRCB 93-03 within 15 days.
These licensees discussed minor deviations from the actions and schedule requested in the bulletin; the staff found these deviations acceptable.
The majority of licensees have decided to install a backfill modification which would constantly purge the reference leg with a very low flow rate (0.008 gpm) of water supplied by the Control Rod Drive (CRD) system.
The constant flow of water up the reference leg would prevent dissolved gases from migrating down the reference leg.
This modification would be installed during the next planned refueling outage, or if an unplanned shutdown should occur, they plan to install the modification at that time, provided engineering and procurement have been completed.
The majority of plants will enter cold shutdown for refueling either this fall or during the spring of 1994.
The staff recognizes that it may be difficnit for some licensees (i.e. those plants shutting down very soon after July 30, 1993) to design and install a hardware modification in the time frame requested in NRCB 93-03.
On August 17, 1993, the staff issued SECY-93-223.
The Commission paper indicated that the staff will consider exercising flexibility on a plant-specific basis to prevent undue hardship or adverse consequences, consistent with Commission guidance.
On the basis of its initial review of licensee responses to the bulletin and meetings with several licensees, the staff believes that, if a delay is warranted, a schedule extension cn the order of 60 days would be appropriate (if a plant should have an unplanned cold shutdown very soon after July 30, 1993).
The staff has also concluded, however, that the schedule submitted by some licensees is not consistent with the intent of NRCB 93-03, and is considering the need for additional regulatory action to accelerate their schedules.
Contact:
Amy E.
Cubbage, BWR Reactor Systems Section, Reactor Systems Branch, Division of Systems Safety and Analysis, US NRC, Washington, DC 20555.
(301) 504-2875
BP25 (09/93)
HIGHLIGHTS OF BWR WATER LEVEL INSTRUMENT ERRORS DUE TO NONCONDENSIBLE 3AS Potential for significant errors in level indication when e
reactor is rapidly depressurized below approximately 450 psig, and during normal plant depressurization belov 450 psig.
Information Notice No. 92-54, Generic Letter No. 92-04, Information Notice No 93-27, and NRC Bulletin No. 93-03 were issued by the staff.
Staff has concluded that interim plant operation is e
acceptable, In the long-term, however, the staff expects each licensee e
to implement hardware modifications to ensure that its level instrumentation system design is of high functional reliability, e
All affected licensees have committed to install hardware modifications to resolve this issue.
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BP26 (09/93)
FIRE BARRIER ISSUES Regulatory
Background:
NRC-approved plant fire protection programs as referenced by the Plant Operating License Conditions and Appendix R to 10 CFR Part 50, Section III G.1.a,
" Fire Protection of Safe Shutdown capability," require one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control stations to be free from fire damage.
To ensure that electrical cables and components are free from fire damage,Section III G.2 of Appendix R requires that the safe shutdown trains be separated by one of the following:
(1) separating cables and equipment and associated circuits of redundant trains by a fire barrier having a 3-hour rating, (2) separating cables and equipment and associated non-safety circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazard, or (3) enclosing cable and equipment and associated non-safety circuits of one redundant train in a fire barrier having a 1-hour rating.
If the licensee chooses to rely on separation or to install a 1-hour barrier, it must also install fire detection and automatic fire suppression systems in the fire area.
Thermo-Lag Fire barriors:
Approximately 80 licensees use Thermo-Lag fire barriers to satisfy the NRC's fire protection requirements.
Thermo-Lag 330-1 is a proprietary fire barrier material manufactured and supplied by Thermal Science, Incorporated.
The vendor manufactures Thermo-Lag in nominal 1/2-inch and 1-inch thicknesses which provide fire endurance ratings of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, respectively.
Thermo-Lag is a sacrificial subliming material that is consumed when it is exposed to a fire.
In a fire, the solid material sublimes, the subliming gases are decomposed by the fire, and the virgin Thermo-Lag material is replaced by a char layer.
The sublimation process and the insulating effects of the resulting char layer protect the equipment located within the confines of the fire barrier from the effects of the fire.
Conversely, more traditional fire barriers, such as concrete block walls, provide j
fire endurance by maintaining structural integrity during the i
fire exposure and limiting heat transfer through the barrier.
Staff Concerns:
The staff has three principal concerns with the Thermo-Lag fire barriers:
fire endurance, combustibility, and ampacity derating.
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1 BP26 (09,93) l Qualification fire tests of cable tray and conduit barriers conducted by industry and small-scale panel tests performed by the NRC staff at the National Institute of Standards and I
Technology (NIST) demonstrated that Thermo-Lag fire barriers may not provide the fire endurance needed to satisfy the NRC's requirements.
The test results led to the issuance of NRC Information Notice (IN) 92-55, NRC Bulletin (NRCB) 92-01, and NRCD 92-01, Supplement 1.
The NRC is also concerned that some Thermo-Lag barriers used by some licensees, such as walls and ceilings, have not been qualified as fire barriers by test.
The NRC is also concerned that Thermo-Lag may burn more readily i
when exposed to fire than originally believed by the NRC and j
licensees.
The staff is concerned about the following possible situations (1) a licensee has not considered the combustibility of the Thermo-Lag materials in its fire hazards analyses, or (2) 1 a licensee is using Thermo-Lag where a noncombustible material should be used.
The NRC issued IN 92-82 on December 15, 1992, to inform the licensees of this issue.
The staff is considering long-range actions.
Finally, the NRC is concerned that the ampacity derating factors used by the licensees to dorate their power cables may not be great enough to actually account for the insulating effects of i
the Thermo-Lag material.
Therefore, cable temperatures may exceed their temperature ratings, possibly resulting in an accelerated aging of the cable insulation.
While the staff is concerned that Thermo-Lag fire barriers may not, in all cases, meet NRC requirements, it believes the i
relative safety significance of these concerns is low for several reasons, including:
(1)
In response to generic communications, licensees have established fire watches to compensate for possibly inoperable fire barriers.
This measure provides an adequate level of fire protection until licensees develop and implement permanent corrective actions.
l (2)
Licensees rely on a defense-in-depth concept where multiple safety measures are incorporated.
Automatic fire detection and sprinkler systems are provided in areas which have safe shutdown equipment.
Trained fire brigades are required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day at all plants.
Fuels that can feed a fire and ignition sources to start a fire are controlled.
Because of these measures, it is unlikely that a fire significant enough to challenge a fire barrier will occur.
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BP26 (09/93)
Actions Taken by the Staff:
In June 1991, NRR established a special team to review the safety significance and generic applicability of the technical issues regarding the use of Thermo-Lag.
The special review team issued its final report, which identified specific concerns along with their technical bases, in April 1992.
Subsequently, the staff prepared an action plan to address the issues associated with Thermo-Lag and the NRC fire protection program.
The scope of the action plan includes coordination with industry and testing by the staff.
The staff has issued the following generic communication documents on Thermo-Lag:
IN 91-47, " Failure of Thermo-Lag Fire Barrier Material To Pass Fire Endurance Test," August 6, 1991.
IN 91-79, " Deficiencies in the Procedures for Installing Thermo-Lag Fire Barrier Materials,"
December 6, 1991.
IN 92-46, "Thermo-Lag Fire Barrier Material Special Review Team Final Findings, Current Fire Endurance Tests, and Ampacity calculation Errors," June 23, 1992.
NRCB 92-01, " Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage," June 24, 1992, and reviewed licensee responses.
IN 92-55, " Current Fire Endurance Test Results for Thermo-Lag Fire Barrier Material," July 27, 1992.
NRCB 92-01, Supplement 1,
" Failure of Thermo-Lag 330 Fire Barrier System to Perform its Specified Fire Endurance Function," August 28, 1992.
Generic Letter (GL) 92-08, "Thermo-Lag 330-1 Fire Barriers," December 11, 1992.
IN 92-82, "Results of Thermo-Lag 330-1 Combustibility Testing," December 15, 1992.
In addition, the staff revised previously developed fire barrier test acceptance criteria to clarify them and incorporate current technical information.
The staff provided the improved criteria, which will be applicable to all new fire barrier testing, to the Nuclear Management and Resources Council (NUMARC) on November 19, 1992.
The staff will provide these criteria to licensees in a GL, " Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains Within
BP26 (09/93) the Same Fire Area (Supplement 1 to GL 86-10, 'Impl aentation of Fire Protection Requirements')".
The proposed GL F.upplement was published in the Federal Register to solicit public comments on July 23, 1993.
The staff completed a small-scale fire test program at NIST and witnessed full-scale fire endurance qualification tests and ampacity derating tests performed by Texas Utilities Electric Company for the Comanche Peak Steam Electric Station and by the Tennessee Valley Authority for the Watts Bar Nuclear Plant at Omega Point Laboratories (OPL) in San Antonio, Texas.
NUMARC began constructing specimens for the generic industry
]
testing of Thermo-Lag installations at OPL in July 1993.
The test program includes 2 phases.
Phase 1 includes 7 upgraded test configurations which are funded by Thermal Science, Inc. (TSI, the vendor), and Phase 2 includes 10 test configurations (existing industry installations and additional upgrades) which are funded by NUMARC.
The staff is monitoring NUMARC's test program closely by observing activities at OPL and communicating frequently through letters and meetings.
l During February 1993, the staff completed a reassement of the NRC fire protection review and inspection programs.
The staff has prepared a Fire Protection Task Action Plan to implement the recommendations made as a result of the fire protection program reassessment and is working to resolve the tasks in a timely 1
manner.
l The staff is also evaluating fire barriers other than Thermo-Lag.
To date, the staff has sent a generic set of questions to all five vendors, other than TSI, believed to supply fire barriers to commercial nuclear power plants.
The staff is also reviewing available vendor and licensee test reports and other documents in an effort to verify the adequacy of these fire barriers.
Additionally, the staff is in the process of testing these other fire barriers at the National Institute of Standards and j
Technology.
Based on information obtained to date, the staff has issued the following generic communication documents on these other fire barriers:
)
IN 93-40, " Fire Endurance Test Results for Thermal Ceramics FP-60 Fire Barrier Material," May 26, 1993.
IN 93-41, "One Hour Fire Endurance Test Results for Thermal Ceramics Kaowool, 3M Company FS-195 and 3M Company Interam E-50 Fire Barrier Systems," May 28, i
1993.
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BP26 (09/93)
Actions Planned by the Staff:
4 The NRC staff plans to issue the supplement to GL 86-10 as soon as public comments are resolved.
The staff will also continue to work with NUMARC and individual licensees to resolve the generic and plant specific technical concerns.
The staff will complete its evaluation of the other fire barrier systems used by the licensees to satisfy the NRC's fire protection requirements.
A supplement to Generic Letter 92-08 that will request-licensees to address issues relating to other fire barrier systems is being considered for issuance in 1994.
I
Contact:
K. Steven West, Chief, Special Projects Section Plant Systems Branch, US NRC Washington, DC 20555.
(301) 504-1220 i
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BP26 (09/93)
HIGHLIGHTS OF FIRE BARRIER ISSUES Appendix R requires separation of safe shutdown trains by a e
3-hour fire barrier or a 1-hour barrier with fire detection and suppression.
Approximately 80 licensees use Thermo-Iag 330-1 fire barriers to comply with Appendix R.
NRC staff has concerns regarding the fire endurance, e
combustibility, and ampacity derating of the Thermo-Lag barriers.
Applicable generic communications on Thermo-Lag include:
IN 91 Failure of Thermo-Lag Fire Barrier Material To Pass Fire Endurance Test (8/6/91)
IN 91 Deficiencies in the Procedures for Installing Thermo-Lag Fire Barrier Materials (12/6/91)
IN 92 Thermo-Lag Fire Barrier Material Special Review Team Final Findings, Current Fire Endurance Tests, and Ampacity Calculation Errors (6/23/92)
NRCB 92 Failure of Thermo-Lag 330 Fire Barrier System to Maintain Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage (6/24/92)
IN 92 Current Fire Endurance Test Results for Thermo-Lag Fire Barrier Material (7/27/92)
NRCB 92-01, Sup. 1 - Failure of Thermo-Lag 330 Fire Barrier System to Perform its Specified Fire Endurance Function (8/28/92)
GL 92 Thermo-Lag 330-1 Fire Barriers (12/11/92)
IN 92 Results of Thermo-Lag 330-1 Combustibility Testing (12/15/92) j NRC staff has published a proposed generic letter supplement e
(GL 86-10, Supplement 1) that provides fire endurance test acceptance criteria for public comment.
NRC staff is reviewing other fire barriers used by licensees e
to satisfy NRC fire protection requirements.
Applicable generic communications include:
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BP26 (09/93)
IN 93 Fire Endurance Test Results for Thermal Ceramics FP-60 Fire Barrier Material (5/26/93)
IN 93 One Hour Fire Endurance Test Results for Thermal Ceramics Kaowool, 3M Company FS-195 and 3M Company Interam E-50 Fire Barrier Systems (5/28/93)
NRC staff plans to issue a generic letter supplement to e
include fire barriers manufactured by other vendors.
BP27 (09/93)
STEAM GENERATOR TUBE PLUGGING CRITERIA
Background:
The thin-walled tubing of the steam generator (SG) constitutes j
over 50% of the reactor coolant pressure boundary (RCPB) for Pressurized Water Reactors.
The integrity of this boundary is particularly important in minimizing the release of radioactive fission products to the environment.
The Nuclear Regulatory Commission (NRC) has a regulatory program designed to ensure that the structural and leakage integrity of the steam generator tubing is maintained at nuclear power plants in the U.S.
In the early to mid 1970s when phosphate water chemistry was primarily used for secondary water chemistry control, the main cause removal from service of SG tubes by plugging was wastage.
Wastage is characterized by large area thinning of the tube wall.
Therefore, SG tube plugging criteria for most plants were based on uniform thinning of the steam generator tube walls.
- However, the current dominant degradation mechanism is stress corrosion cracking (SCC).
Tube Plugging Criteria:
l Because of the potential consequences of the loss of SG tube integrity, the agency has measures for ensuring that the integrity of the SG tubing is maintained.
The traditional tube plugging criteria have typically been based on a minimum wall thickness requirement which assumes that the degradation involves uniform thinning of the tube wall in the axial and circumferential directions.
The assumption of uniform thinning i
conservatively bounds the degrading effects of all flaw types occurring in the field, and is the basis of the standard 40%
depth-based plugging limit.
However, the 40% plugging limit is very conservative for highly localized flaws such as pits and short cracks.
The staff has approved higher depth-based limits, ranging up to 64%, for specific types of flaws at specific plants; for instance, for pitting at Indian Point Unit 3 in the early 1980s.
In recent years, utilities have proposed various flaw-specific plugging limits which do not incorporate a minimum wall thickness requirement.
These proposals are often referred to as alternate plugging criteria (APC) since they would be an alternative to the depth-based criterion.
Each of these proposals would permit tubes with up to 100% through-wall cracks to remain in service, subject to certain restrictions.
These restrictions are intended to ensure adequate structural and leakage integrity.
Certain of these proposed limits, known as F* (F-star) and L* (L-star) limits, have been approved by the NRC and have been in use for several years.
The F* and L* limits apply to primary water
BP27 (09/93) stress corrosion cracking (PWSCC) flaws located within the thickness of the tubesheet for tubes which have been hardroll expanded.
Rupture of the tubing within the tubesheet region is precluded by the constraint against tube wall deformation provided by the tubesheet.
The specified F* and L* limits have been set to preclude tube pullout from the tubesheet during postulated accidents and to ensure the leaktight integrity between the fully expanded tubing and the tubesheet.
Another type of flaw-specific plugging limit which has been proposed by various utilities involves the use of a bobbin voltage amplitude based limit for axially oriented outside diameter stress corrosion cracking (ODSCC) confined within the tube support plate (TSP) intersections.
This proposal includes commitments to the use of enhanced inspection methods, enhanced sampling plans, and reduced primary-to-secondary leak rate limits.
The voltage-based plugging limit is established from a burst pressure / voltage amplitude correlation which is adjusted for allowances in voltage amplitude uncertainty and for projected voltage growth during the next operating cycle.
This limit is intended to ensure adequate structural and leakage integrity of the tubing.
Another flaw-specific plugging limit includes length-based limits for PWSCC at roll transition locations, similar to limits currently being implemented in some European countries.
As is the case for voltage amplitude based limits, proposals for length-based limits are expected to be programmatic, involving commitments to specific inspection methods, inspection sampling plans, and reduced primary-to-secondary leak rate limits as well as revised plugging limits.
The Electric Power Research Institute (EPRI) has issued a technical support document to support length-based plugging limit proposals from U.S.
utilities.
Current Status:
The NRC staff has approved the use of restricted versions of the voltage-amplitude based limits.
The restricted versions incorporated much smaller voltage limits than those proposed and limited the use to one operating cycle for several plants (Farley 1
Units 1 and 2, D.C.
Cook Unit 1, and Catawba, Unit 1).
This restricted version of the APC is referred to as the Interim Plugging Criteria (IPC).
A special task force has reviewed the technical bases for the voltage based IPC.
The conclusions and recommendations of the Task Group have been issued as a draft report for comment, NUREC-1477, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes."
The staff is considering a generic approach to defect-specific steam generator repair criteria based on a performance-based rule that would incorporate the task group's
5 BP27 (09/93) recommendations and the technical work being-done by the industry on APC to encourage the development of better inspection technology.
The idea being that if defects can be more accurately characterized, there would be less need for large conservatism in the repair criteria.
With respect to the other proposals for alternate plugging criteria, all plant specific APC reviews have been sunpended pending the generic review of the alternate plugging criteria.
The NRC staff expects to complete the project in 18 to 24 months.
Requests from plants for renewal of approved interim plugging criteria.are being considered.
As topical reports are reviewed and approved by the staff, the staff may consider plant specific applications to adopt the criteria that are covered by those topical reports.
Contact:
Jack R. Strosnider, Chief, Materials and Chemical Engineering Branch, Division of Engineering, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission, Washington, DC 20555, (301) 504-2796.
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BP27 (09/93)
HIGHLIGHTS OF STEAM GENERATOR TUBE PLUGGING CRITERIA Steam generator (SG) tubing constitutes a large percentage of the reactor coolant pressure boundary (RCPB) for PWR plants.
Traditional plugging criteria developed for uniform thinning (i.e., wastage) resulted in a depth-based limit.
Defects extending 40% through-wall require repair / plugging.
The 40% depth-based limit applies to all degradation mechanisms, not just wastage.
The 40% plugging limit tends to be overconservative for other flaw mechanisms.
Certain defect-specific plugging limits have been approved in the past, primarily for degradation within the tubesheet (e.g., F* and L* criteria).
The industry is proposing various forms of alternate plugging criteria (APC) to the traditional depth-based limit for plugging / repair of SG tubes.
These include voltage-based and length-based limits.
A degradation-specific approach to managing steam generator tube integrity has advantages (i.e., it requires using more appropriate inspection and repair criteria for the specific flaw type that is encountered).
A voltage-based alternate tube repair limit more conservative than that proposed by the industry has been approved (referred to as the interim plugging criteria or IPC) while review of the industry proposal proceeds (scheduled completion in late 1994).
Significant issues being considered in the staff's review include:
Maintaining a low probability of SG tube rupture Evaluating the potential for and consequences of increased leakage under postulated accident conditions Implementation of the IPC has actually resulted in more restrictive operating leakage limits.
A special task force has reviewed the technical bases for the voltage based IPC.
The conclusions and recommendations of the Task Group has been issued as a draft report for comment, NUREG-1477, " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes."
BP27 (09/93)
The staff is initiating a generic approach to defect specific steam generator repair criteria based on a performance-based rule that will incorporate the task group's recommendations and the technical work being done by the industry on APC to encourage the development of better inspection technology.
The idea being that if defects can be more accurately characterized, there would be less need for large conservatism in the repair criteria.
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BP28 (09/93)
MEDICAL ISSUES
Background:
The Nuclear Regulatory Commission has statutory authority for the regulation of byproduct material as a result of the Atomic Energy Act of 1954.
NRC's mandate to protect public health and safety includes the regulation of the medical use of byproduct material in the fields of nuclear medicine, radiation therapy, and research.
Consequently, NRC's involvement with, and interest in, this field is not recent, and in fact, the NRC and its predecessor, the Atomic Energy Commission, have regulated the medical use of radioisotopes since 1946.
On February 9,
- 1979, NRC issued a Medical Policy Statement entitled " Regulation of the Medical Uses of Isotopes; Statement of General Policy" (44FR8242).
In it, the Commission stated that it will:
(1) continue to regulate the medical uses of radioisotopes as necessary to provide for the radiation safety of workers and the general public; (2) regulate the radiation safety of patients where justified by the risk to patients and where voluntary standards, or compliance with these standards, are inadequate; and (3) minimize its intrusion into medical judgments affecting patients, and into other areas traditionally considered to be a part of the practice of medicine.
NRC's oversight of the medical use of byproduct material is focused on ensuring its safe medical use to protect public health and safety.
The NRC's emphasis includes protecting workers and visitors from unwanted radiation exposure, as well as assuring that administration of, or radiation from, byproduct material are in accordance with the direction of a physician qualified to use
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such materials or radiation.
j Byproduct material or radiation from byproduct material is regulated by either State or Federal laws.
Twenty-nine States, i
known as Agreement States, have entered into an agreement with the NRC to regulate the use of byproduct material (as authorized by section 274 of the Atomic Energy Act).
These States issue licenses and currently regulate approximately 4,500 institutions I
for medical use, that is, university medical centers, hospitals, clinics, and physicians in private practice.
The NRC regulates the medical use of byproduct material in 21 States, the District of Columbia, the commonwealth of Puerto Rico, and various territories of the United States and administers approximately 2000 licenses for medical use.
Several major issues in the regulation of the medical use of byproduct material, currently under review by NRC staff, were j
raised in a series of articles, dated December 13-17, 1992, in the Cleveland Plain Dealer.
These include:
(1) effectiveness of NRC's oversight of the national medical use program; (2) followup of patients subject to misadministrations; and (3) i
BP28 (09/93) expansion of NRC's regulatory purview to include sources of radiation other than byproduct material.
The articles allege j
that NRC was unaware of the extent of some misadministrations (and/or failed to take appropriate actions in response) in which deaths and serious injuries resulted from radiation exposure.
NRC staff continues to review the issues raised in the articles relative to the reassessment initiatives for the medical use program.
Medical Management Plan:
Several recent rulemakings by NRC in the area of medical use have prompted criticism and opposition by certain elements within the regulated community.
These include:
(1) the " Interim Final Rule" (effective August 23, 1990) which amended regulations related to the preparation of radiopharmaceuticals and the therapeutic use of radiopharmaceuticals; and (2) the " Quality Management Program (QMP) and Misadministrations" Rule (effective January 27, 1992) which required licensees to submit a copy of their QMP and a written certification that a QMP had been implemented.
The Commission has been working to effectively resolve these issues and alleviate the associated concerns of the medical community while maintaining communication with the involved parties.
Consequently, during September 1992, a
" Medical Issues" paper was developed by the staff.
This paper was discussed with the Advisory Committee on the Medical Uses of Isotopes, and representatives from the Agreement States during October 1992, and with NRC regional management in November 1992.
As a result, the staff has developed a management plan to include reassessment of the medical use program and initiated a number of actions to address the more pressing problems.
Two key components of the management plan are: (1) continuation of staff initiatives, not only those previously identified to senior management and the Commission, but others that have emerged as the result of recent events; and (2) continuation of interaction with the regulated community.
The current initiatives focus on nine major areas of the medical use regulatory program and include such activities as the development of more comprehensive licensing and inspection guidance, and rulemaking.
The staff briefed the commission on the management plan on September 10, 1993, and has subsequently received Commission approval for implementation.
The medical management plan is publicly available and is described in SECY-93-244, issued August 31, 1993.
Oncology Services, Indiana, Pennsylvania Misadministration:
The_ licensee committed to provide a written detailed response to the March 5, 1993, NRC deficiency letter to address weaknesses in radiation safety training, RSO oversight, normal emergency
BP28 (09/93) operating procedures, and internal and independent audit program.
Region I received the licensee's submittal on April 8, 1993.
I Upon review, Region I issued a letter to OSC dated April 22, 1993, to acknowledge the submitted radiation safety program upgrades, inform OSC that the upgraded program appeared to meet NRC requirements, and inform OSC of planned future NRC inspections.
Inspections were performed during the period of April 22 through May 5, 1993, and an inspection report was issued on May 28, 1993.
The inspection findings confirmed that appropriate supervision was exercised by the Medical Director and authorized physician users, and the physicists were cognizant of and performed their duties in accordance with the applicable requirements.
Additionally, the inspectors verified that the corporate oversight of licensed activities had improved.
On June 3,
1993, NRC issued general relaxation Orders to OSC which allowed them to resume brachytherapy procedures, without specific relaxation requests, at the Greater Harrisburg and Greater Pittsburgh Cancer Centers.
A hearing by the Atomic Safety and Licensing Board (ASLB) to review NRC's suspension of licensed activities is expected to occur during October 1993.
Congressional Hearings:
U.S. Senate Committee on Governmental Affairs On May 6, 1993, the NRC testified before the U.S. Senate Committee on Governmental Affairs on NRC's national program for regulation of the medical use of radioactive material.
One area of interest for the Committee was possible options for NRC's regulatory jurisdiction for medical use of radioactive material to ensure a more efficient national program.
As a result of this hearing, an NRC Task Force was established to prepare a report for Commission review and submittal to the Senate Committee on this subject.
The Task Force briefed the Commission during late July 1993 on various regulatory options.
Subsequently, the Task Force received additional guidance from the Commission to expand their efforts and identify the type of data needed by NRC to further evaluate the impact of implementing various options presented.
This revised report was submitted to the Commission and Senate Committee during September 1993.
The staff awaits further direction from the Commission.
U.S. House of ReDresentatives Subcommittee on Environment.
Enerov, and Natural Resources During late July 1993, NRC staff representing the Offices of the Nuclear Material Safety and Safeguards, and State Programs testified before the U.S. House of Representatives Subcommittee on Environment, Energy, and Natural Resources regarding NRC's oversight of the national medical use regulatory program.
In i
particular, NRC's oversight of Agreement State medical use regulatory programs and their compatibility with NRC's regulatory
BP28 (09/93) program were the focus.
The staff is scheduled to testify on other aspects of the NRC's materials program either during late calendar year 1993 or early 1994.
Reevaluation Initiatives:
The staff has taken several steps to re-evaluate the medical use program which include those areas discussed below.
o Risk Analysis and Human Factors:
NRC currently has several contracts in place with national laboratories and private entities to evaluate the risks associated with new technologies and the human error component of misadministrations.
These include:
(1) a contract for the investigation of certain therapy misadministrations, to analyze the root cause(s); (2) contracts to evaluate the contribution of the human-machine interface in brachytherapy and teletherapy performance errors; and (3) contracts to evaluate quality assurance and risk associated with brachytherapy procedures and devices, and gamma stereotactic surgery.
These efforts could result in revised equipment and operating procedures, and form the basis for revised regulations and revision of inspection procedures and frequency.
Contract work has yet to be completed.
o Medical Use Program Audits:
As directed by the Commission, independent audits have or will be conducted of NRC's medical use program, in addition to the staff medical management plan.
An NRC senior manager conducted a review of the medical use regulatory program and submitted a final report to the Commission during June 1993.
The findings of ae senior manager's review were considered and incorporated $nto the staff's proposed medical management plan submitted to the Commission during September 1993.
The external review of the program will be conducted by the National Academy of Sciences.
The goal of the external review is to develop an assessment of the adequacy and appropriateness of the current framework for medical use of byproduct material.
It will include an in-depth review of the basic regulatory rules, policies, practices, and procedures.
i Update on Quality Management Rule:
On July 25, 1991, the NRC amended its regulations governing therapeutic administrations of byproduct material and certain uses of radioactive sodium iodide to require implementation of policies and procedures, as part of a quality management program (QMP), to provide high confidence that the byproduct material or
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BP28 (09/93) radiation from byproduct material will be administered as l
directed by an authorized user physician.
The final rule became effective January 27, 1992.
A review panel. consisting of Headquarters and Regional staff, including an NRC Medical Visiting Fellow, meets weekly to review proposed enforcement i
actions that involve QMP procedures which may or may not have resulted in a misadministration.
The purpose of this review is to assess the inspection procedure.and ensure that the enforcement policy is applied consistently for violations of QM procedures or policies.
Several inspections have occurred as a result of misadministrations that are due, in part, to violations of QMP policies or procedures; some of which have resulted in civil penalties.
NRC has contracted with Lawrence Livermore National Laboratory to review QM programs submitted by licensees.
The purpose of the reviews is to identify deficiencies in the i
submitted programs and provide this information to the licensees for modification to the QMPs.
Implementation of a licensee's QMP will be evaluated during routine inspection.
After the contractor has reviewed a majority of the submitted programs, NRC staff will re-evaluate the QM rule and corresponding guidance to determine whether modifications ara needed.
Update on ACNP/SNM Petition for Rulemaking:
On June 17, 1993, published a proposed rule (58FR33396) to amend its regulations for the medical use of byproduct material.
The proposed rule provides greater flexibility by allowing properly qualified nuclear pharmacists and authorized users, who are physicians, greater discretion to prepare radioactive drugs containing byproduct material for medical use.
The proposed rule would also allow research involving human subjects using byproduct material and the medical une of radiolabeled biologics.
The comment peried expires October 15, 1993, after which, the 3
staff will develop a final rule for Commission approval.
i Additionally, NRC published a final rule on July 22, 1993, (58FR39130) to extend the expiration'date of the interim final rule related to preparation and therapeutic use of radiopharmaceuticals from August 23, 1993, to December 31, 1994.
This action allows licensees to continue to use byproduct material until NRC completes the related rulemaking described above.
-Contact: Larry Camper, Section Leader, Office of Nuclear Material Safety and Safeguards, US Nuclear Regulatory Commission, Washington, DC, 20555, (301) 504-3417 l
BP28 (09/93)
HIGHLIGHTS ON RE-EVALUATION INITIATIVES o
Staff medical management plan was forwarded to the Commission during September 1993.
The management plan contains: 1) action items resulting from the findings of the Incident Investigation Team initiated to investigate the patient fatality in November 1992, in Indiana, Pennsylvania;
- 2) items identified in the senior manager review; and 3) items previously identified by the staff and discussed in the staff medical issues paper developed during September 1992.
The staff is scheduled to brief the Commission on this subject on September 10, 1993.
o Several contracts with national laboratories and private entities involve evaluating risks associated with new technologies and the human error component of misadministrations.
Some final reports will be submitted within the next three months.
o An internal review by an NRC senior management representative was performed and a report submitted to the Commission during June 1993.
An external review will be conducted by the National Academy of Sciences and is in the initial stages of coordination with NRC staff.
The final report will not be submitted until early 1996.
o As a result of the May 6, 1993, hearings before the U.S.
Senate Committee on Governmental Affairs, an NRC Task Force was established to prepare a report for Commission review that discussed possible options for NRC's regulatory jurisdiction for medical use of radioactive material.
The Task Force briefed the Commission during late July 1993 on this subject.
Subsequently, the Task Force received 1
additional direction from the Commission to expand their efforts and identify the type of data needed by NRC to evaluate the impact of implementing various options presented.
This revised report was submitted to the Commission and U.S.
Senate during September 1993.
The staff awaits further direction from the Commission.
i o
During late July 1993, NRC staff representing the Offices of Nuclear Material Safety and Safeguards, and State Programs testified before the U.S. House of Representatives Subcommittee on Environment, Energy, and Natural Resources regarding NRC's oversight of the national medical use regulatory program.
In particular, NRC's oversight of Agreement State medical use regulatory programs, and the compatibility of these programs with the NRC's regulatory program were the focus.
The staff is scheduled to testify before this committee on other aspects of the NRC's materials program either during late calendar year 1993 or early 1994.