[Federal Register Volume 64, Number 79 (Monday, April 26, 1999)]
[Notices]
[Pages 20328-20339]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-10357]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-458 and 50-440; License Nos. NPF-47 and NPF-58]


Entergy Operations, Inc. FirstEnergy Nuclear Operating Company 
Notice of Issuance of Director's Decision Under 10 CFR 2.206

    Notice is hereby given that the Director, Office of Nuclear Reactor 
Regulation, has issued a Director's Decision with regard to Petitions 
dated September 25, 1998, and November 9, 1998, filed by Mr. David A. 
Lochbaum on behalf of the Union of Concerned Scientists (UCS), 
hereinafter referred to as the ``Petitioner.'' The Petitions concern 
the operation of the River Bend Station (River Bend) located in St. 
Francisville, Louisiana, and the Perry Nuclear Power Plant (Perry) 
located in Perry, Ohio.
    The Petitions requested that River Bend and Perry should be 
immediately shut down and their respective operating licenses suspended 
or modified until the facilities' design and licensing bases were 
updated to permit operation with failed fuel assemblies, or until all 
failed fuel assemblies were removed from the reactor core. The 
Petitioner also requested that a public hearing be held to discuss this 
matter in the Washington, DC, area.
    As the basis for the September 25, 1998, request, the Petitioner 
raised concerns stemming from the Nuclear Regulatory Commission (NRC) 
Daily Event Report No. 34815, dated September 21, 1998, whereby Entergy 
Operations, Inc. (the licensee for River Bend) reported a possible fuel 
cladding defect. The Petitioner referred to concerns raised in a UCS 
report of April 2, 1998, regarding nuclear plant operation with fuel 
cladding leakage. The UCS considers such operation to be potentially 
unsafe and to be in violation of Federal regulations. In the Petition, 
a number of references to the River Bend Updated Safety Analysis Report 
(USAR) were cited that the UCS believes prohibit operation of the 
facility with known fuel leakage.
    The Petition of November 9, 1998, raises concerns originating from 
the NRC's Weekly Information Report for the week ending October 30, 
1998, in

[[Page 20329]]

which the staff discussed the licensee's findings of possible fuel 
cladding defects. The Perry Petition also referred to concerns raised 
in the UCS report of April 2, 1998.
    In its report of April 2, 1998, the UCS expresses the opinion that 
existing design and licensing requirements for nuclear power plants 
preclude their operation with known fuel cladding leakage. The UCS 
position is based on the assessment of updated final safety analysis 
reports (UFSARs or USARs) of four plants, vendor documentation, 
standard technical specifications, and pertinent NRC correspondence. In 
addition to recommending that the NRC take steps to prohibit nuclear 
power plants from operating with fuel cladding damage, the report 
specifically recommends plant shutdowns upon detection of fuel leakage 
and that safety evaluations be included in plant licensing bases, which 
consider the effects of operating with leaking fuel to justify 
operation under such circumstances.
    Finally, the two Petitions also stated that the licensing basis for 
worker radiation protection was violated whenever the licensee operated 
the plant with potential fuel cladding failures. The Petitions 
references included various USAR Sections and NRC Information Notice 
87-39, ``Control of Hot Particle Contamination at Nuclear Plants,'' and 
stated that industry experience has demonstrated that reactor operation 
with failed fuel cladding increased radiation exposures for plant 
workers.
    On February 22, 1999, the NRC conducted an informal public hearing 
regarding the River Bend Petition as well as a similar petition 
submitted pursuant to 10 CFR 2.206 involving Perry, operated by the 
FirstEnergy Nuclear Operating Company. The hearing gave the Petitioner, 
the licensees, and the public an opportunity to provide additional 
information and to clarify issues raised in the Petitions.
    The Director of the Office of Nuclear Reactor Regulation has 
determined that the two requests, to require River Bend and Perry to be 
immediately shut down and their operating licenses suspended or 
modified until the facilities' design and licensing bases were updated 
to permit operation with failed fuel assemblies, or until all failed 
fuel assemblies were removed from the reactor core, be denied. The 
reasons for this decision are explained in the Director's Decision 
pursuant to 10 CFR 2.206 (DD-99-08), the complete text of which follows 
this notice and is available for public inspection at the Commission's 
Public Document Room, the Gelman Building, 2120 L Street, NW, 
Washington, DC, and at the local public document rooms located at the 
Government Documents Department, Louisiana State University, Baton 
Rouge, Louisiana, and the Perry Public Library, 3753 Main Street, 
Perry, Ohio.
    A copy of the Director's Decision will be filed with the Secretary 
of the Commission for the Commission's review in accordance with 10 CFR 
2.206 of the Commission's regulations. As provided for by this 
regulation, the Decision will constitute the final action of the 
Commission 25 days after the date of issuance, unless the Commission, 
on its own motion, institutes a review of the Decision in that time.

    Dated at Rockville, Maryland, this 18th day of April 1999.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.

Director's Decision Under 10 CFR 2.206

I. Introduction

    By Petitions submitted pursuant to 10 CFR 2.206 on September 25, 
1998, and November 9, 1998, respectively, Mr. David A. Lochbaum, on 
behalf of the Union of Concerned Scientists (UCS or Petitioner), 
requested that the U.S. Nuclear Regulatory Commission (NRC) take 
immediate action with regard to the River Bend Station (River Bend) and 
the Perry Nuclear Power Plant (Perry).
    In the Petitions, the Petitioner requested that the NRC take 
immediate enforcement action by suspending the operating license for 
River Bend and Perry until all leaking fuel rods were removed from the 
reactor core or until the facilities' design and licensing bases were 
updated to permit operation with leaking fuel assemblies. Accompanying 
the Petitions was the UCS report ``Potential Nuclear Safety Hazard--
Reactor Operation With Failed Fuel Cladding,'' dated April 2, 1998. 
Entergy Operations, Inc. (the River Bend licensee), provided the NRC 
with its response to its Petition in a letter dated February 11, 1999. 
FirstEnergy Nuclear Operating Company (the Perry licensee) provided a 
response to its Petition in a letter also dated February 11, 1999. On 
February 22, 1999, the NRC held an informal public hearing at which the 
Petitioner presented information related to the safety concerns in the 
Petitions. The NRC staff has determined that the information presented 
in the Petitions and at the informal public hearing did not support the 
action requested by the Petitioner. The basis for my decision in this 
matter follows.

II. Background

    In support of the requests presented in the Petition dated 
September 25, 1998, the Petitioner raised concerns stemming from NRC 
Daily Event Report No. 34815, filed on September 21, 1998, in which 
Entergy Operations, Inc., reported a possible fuel cladding defect at 
River Bend. The Petitioner repeated the concerns raised in the UCS 
report of April 2, 1998, regarding nuclear plant operation with fuel 
cladding leakage. The UCS considers such operation to be potentially 
unsafe and to be in violation of Federal regulations. In addition, the 
Petitioner cites instances in the licensing basis for River Bend that 
it believes prohibit operation of the facility with leaking fuel.
    In the November 9, 1998, Petition, the Petitioner raised similar 
concerns originating from the NRC Weekly Information Report for the 
week ending October 30, 1998, in which fuel leaks detected at Perry on 
September 2, 1998, and on October 28, 1998, were discussed. The 
Petitioner also repeated the concerns raised in the UCS report of April 
2, 1998. The matters raised in support of the Petitioner's requests are 
discussed herein.

III. Discussion

    The September 25, 1998, Petition presents safety concerns for River 
Bend along with the associated generic concerns addressed in the UCS 
report of April 2, 1998. The plant-specific concerns are based on 
portions of the River Bend Updated Safety Analysis Report (USAR) cited 
in the Petition. The November 9, 1998, Petition presents safety 
concerns for Perry arising essentially from the associated generic 
concerns addressed in the UCS report of April 2, 1998. The Perry 
Petition does not reference plant-specific licensing basis 
documentation.
    Since the generic concerns presented in the UCS report bear upon 
the plant-specific concerns cited in the two Petitions, the staff's 
evaluation first considers the UCS report and follows with a discussion 
of the plant-specific concerns.

A. Generic Safety Concerns

    In the UCS report of April 2, 1998, UCS expresses the opinion that 
existing design and licensing requirements for nuclear power plants 
preclude their operation with known fuel cladding leakage. The UCS 
position is based on the assessment of updated final safety analysis 
reports (UFSARs) of four plants, vendor documentation, standard 
technical specifications, and pertinent NRC correspondence. The report 
states

[[Page 20330]]

that the following regulatory and safety concerns exist for plants 
operating with leaking fuel:
     10 CFR 50.59, ``Changes, tests and experiments,'' is 
violated because operation with fuel cladding leakage constitutes an 
unapproved change to the licensing basis for a plant. The report states 
that such operation is an unresolved safety question because the 
criteria of 10 CFR 50.59(a)(2) are satisfied (e.g., probability and 
consequences of an accident may be increased by operating with leaking 
fuel).
     10 CFR 50.71, ``Maintenance of records, making of 
reports,'' is violated because the licensing basis as documented in the 
technical specifications and the analyses contained in the UFSAR for 
the facility do not accommodate operation with leaking fuel.
     Safety analyses for postulated accidents assume intact 
fuel cladding before the event; therefore, plants with known fuel 
leakage could have accidents with more severe consequences than 
predicted as a result of fuel damage. The report further states that no 
information was available showing that operation with leaking fuel has 
been previously evaluated.
     10 CFR 50.34a, ``Design objectives for equipment to 
control releases of radioactive material in effluents--nuclear power 
reactors,'' and other regulations related to the as low as is 
reasonably achievable (ALARA) principle for radioactive materials 
release are violated since plant workers are exposed to a greater risk 
than necessary because of higher coolant activity levels attributable 
to leaking fuel.
    In addition to requesting that the NRC take steps to prohibit 
nuclear power plants from operating with fuel cladding damage, the 
report specifically requests that plants be shut down upon detection of 
fuel leakage, and that safety evaluations be included in plant 
licensing bases that consider the effects of operating with leaking 
fuel to justify operation under such circumstances.
    Before addressing the regulatory concerns raised in the April 1998 
UCS report, the following discussion provides background and bases for 
current NRC guidance and practices with regard to fuel defects.
1. Defense-in-Depth and ALARA Considerations
    In order to protect public health and safety from the consequences 
of potential uncontrolled releases of radioactive fission products 
resulting from the operation of nuclear power plants, plants are 
designed with multiple barriers to fission-product release. This 
traditional ``defense-in-depth'' philosophy is key to assuring that 
radiological doses from normal operation and postulated accidents will 
be acceptably low, as outlined in 10 CFR Part 100, ``Reactor Site 
Criteria.'' Fuel cladding is integral to the defense-in-depth approach 
to plant safety, serving as the first barrier to fission-product 
release.
    The premise of the defense-in-depth philosophy with regard to the 
potential for fission-product release is that plant safety does not 
rely on a single barrier for protection. In this way, a limited amount 
of leakage from each of the barriers--the fuel cladding, the reactor 
coolant system pressure boundary, and the containment--is a design 
consideration and some leakage from each barrier, within prescribed 
limits, is acceptable during operation. These limits, defined within 
the technical specifications, are established as a key component of a 
plant's design and licensing basis. The leakage associated with fuel 
cladding defects is accounted for in plant safety analyses, as 
discussed later in this evaluation under ``Safety Analysis 
Assumptions.''
    Therefore, to meet its defense-in-depth objectives, fuel is not 
required to be leak-free. A limited amount of fuel cladding leakage is 
acceptable during operation since (1) In the event of an accident, 
other fission-product barriers besides the fuel cladding (i.e., the 
reactor coolant system pressure boundary and the containment) help 
prevent uncontrolled releases, (2) limits for reactor coolant system 
activity, as prescribed in the technical specifications, limit the 
level of fuel leakage that is permitted so that the release guidelines 
of 10 CFR Part 100, ``Reactor Site Criteria,'' will not be exceeded 
during accidents, and (3) plant design features and operating 
procedures anticipate leaking fuel and provide means to deal with the 
effects.
    Sources of activity in reactor coolant are fission products 
released from fuel, corrosion products activated in the reactor during 
operation, and fission products released from impurities in fuel 
cladding, tritium produced from the irradiation of water, lithium, and 
boron. Although reactor operators should strive to maintain low levels 
of coolant activity from all of these sources, the staff has long 
recognized that reactor coolant activity cannot be entirely eliminated 
and that some fission products from leaking fuel could be present (see 
Standard Review Plan (SRP), NUREG-0800, Section 4.2, ``Fuel System 
Design''). Thus, plant design considerations, such as reactor coolant 
cleanup systems, shielding, and radwaste controls, have been devised to 
minimize risk to plant workers from exposure to radiation from reactor 
coolant. Plants also implement procedures to respond to leaking fuel 
when leakage is discovered, as was demonstrated by the example of the 
follow-up actions taken by the River Bend and Perry operators to limit 
the production of fission products in the vicinity of the leaking fuel 
rods.
    By containing fuel and fission products, cladding also helps 
maintain radioactive releases to as low a level as is reasonably 
achievable. As previously stated, the technical specifications contain 
limits for the maximum level of coolant activity so that the dose 
guidelines in 10 CFR Part 100 are not exceeded during accidents. These 
are the maximum levels of activity assumed to exist in the reactor 
coolant from normal operating activities. The limits on reactor coolant 
system specific activity are also used for establishing standardization 
in radiation shielding and procedures for protecting plant personnel 
from radiation (see Section B3.4.16 of NUREG-1431, ``Standard Technical 
Specifications, Westinghouse Plants''). Thus, they are consistent with 
NRC regulations requiring licensees to follow an ALARA approach to 
radiation protection.
    The connection between technical specification limits for coolant 
activity and ALARA requirements is key to demonstrating that limited 
fuel leakage during operation is consistent with safe plant operation. 
The ALARA requirement is given in 10 CFR 50.34a and 50.36a. The 
Statement of Considerations for these NRC regulations (35 FR 18385, 
December 3, 1970) contains a discussion of the ``reasonableness'' 
aspect of the ALARA approach. When the Statement of Considerations was 
written, the Commission believed that releases of radioactivity in 
plant effluents were generally within the range of ``as low as 
practicable.'' The Commission also stated, therein, that ``as a result 
of advances in reactor technology, further reduction of those releases 
can be achieved.'' Advances in fuel integrity, design of waste 
treatment systems, and appropriate procedures were cited as areas in 
which the plants had taken steps to meet the reasonableness standard. 
It is important to note that the Commission did not require leak-free 
fuel as a means to satisfy ALARA requirements. In addition to the 
physical barriers to the release cited above, other factors, such as 
radwaste cleanup and plant procedures, provide

[[Page 20331]]

confidence that fission-product release from the fuel can be controlled 
so as to prevent undue risks.
    Later in the same Statement of Considerations, the Commission 
acknowledged the need to allow flexibility of plant operation. 
``Operating flexibility is necessary to take into account some 
variation in the small quantities of radioactivity, as a result of 
expected operational occurrences, which may temporarily result in 
levels of radioactive effluents in excess of the low levels normally 
released'' but still within regulatory limits. The Commission 
recognized that a balance should be maintained between limiting 
exposure to the public and plant operational requirements. Therefore, 
the NRC regulations allow the possibility of increased reactor coolant 
activity levels that might result from limited fuel cladding leaks, but 
require the use of plant equipment to maintain control over radioactive 
materials in gaseous and liquid effluents produced during normal 
reactor operations, including expected operational occurrences. The 
Commission went as far as to define ``as low as practicable'' (the 
phrase later replaced with ``as low as is reasonably achievable'' in 40 
FR 19440, May 5, 1975) in terms of the state of technology, the 
economics of improvements in relation to benefits to public health and 
safety that could be derived by improved technology and methods of 
controlling radioactive materials, and ``in relation to the utilization 
of atomic energy in the public interest.'' This definition appears in 
Section 50.34a itself, mandating that the Commission maintain the 
balance between safety and plant operational requirements.
    By publishing 10 CFR Part 50, Appendix I, ``Numerical Guides for 
Design Objectives and Limiting Conditions for Operation To Meet the 
Criterion `As Low As Is Reasonably Achievable' for Radioactive Material 
in Light-Water-Cooled Nuclear Power Reactor Effluents,'' the Commission 
took steps to provide more definitive guidance for licensees to meet 
the ``as low as practicable'' requirement. Appendix I was published as 
guidance that presented an acceptable method of establishing compliance 
with the ``as low as practicable'' requirement of 10 CFR 50.34a and 
50.36a. In the Statement of Considerations for Appendix I (40 FR 19439, 
May 5, 1975), the Commission characterized the guidance as the 
``quantitative expression of the meaning of the requirement that 
radioactive material in effluents released to unrestricted areas from 
light-water nuclear power reactors be kept `as low as practicable'.'' 
The technical basis for Appendix I contained assumptions for a small 
fraction of leaking fuel rods, as is stated in the Atomic Energy 
Commission's report of July 1973, WASH-1258, ``Final Environmental 
Statement Concerning Proposed Rule Making Action: Numerical Guides for 
Design Objectives and Limiting Conditions for Operation To Meet the 
Criterion ``As Low as Practicable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents.''
2. Associated Regulations and Guidance
    Fuel integrity is explicitly addressed in NRC regulations in 
several instances, and plant licensing bases specifically discuss fuel 
performance limits. To implement NRC regulations, the staff developed a 
number of guidance documents for licensees to use in developing their 
licensing basis. This section outlines the regulatory framework on fuel 
integrity during normal plant operation and discusses instances in 
which the staff has considered the safety implications of fuel 
integrity.
a. Regulatory Requirements
    The General Design Criteria (GDC) of 10 CFR Part 50, Appendix A, 
``General Design Criteria for Nuclear Power Plants,'' contain 
references to fuel design criteria. When fuel performance is used as a 
criterion for a safety function, system, or component, the phrase 
``specified acceptable fuel design limits'' (SAFDLs) appears in the 
following GDC:

 GDC 10, ``Reactor Design''
 GDC 12, ``Suppression of Reactor Power Oscillations''
 GDC 17, ``Electric Power Systems''
 GDC 20, ``Protection System Functions''
 GDC 25, ``Protection System Requirements for Reactivity 
Control Malfunctions''
 GDC 26, ``Reactivity Control System Redundancy and 
Capability''
 GDC 33, ``Reactor Coolant Makeup''
 GDC 34, ``Residual Heat Removal''

    GDC 10, 17, 20, and 26 use this wording in conjunction with 
anticipated operational occurrences and conditions of normal operation. 
For example, GDC 10 requires ``appropriate margin to assure that 
specified acceptable fuel design limits are not exceeded during any 
condition of normal operation, including the effects of anticipated 
operational occurrences.'' As discussed later in this section, SAFDLs 
for a plant are described in plant documentation, typically the UFSAR 
or the FSAR, and are met by operating within technical specifications 
limits.
    NRC regulations also specify that certain conditions beyond steady-
state operation be included in evaluations of the normal operating 
regime for a plant. These are called anticipated operational 
occurrences (AOOs) and are sometimes referred to as ``anticipated 
operating transients.'' In Appendix A to 10 CFR Part 50, the staff 
defines AOOs as ``those conditions of normal operation which are 
expected to occur one or more times during the life of the nuclear 
power unit.'' GDC 29, ``Protection Against Anticipated Operational 
Occurrences,'' gives a general requirement for protection system and 
reactivity control system performance during AOOs, but does not mention 
fuel integrity. Examples of AOOs are the loss of all reactor coolant 
pumps, turbine trip events, and loss of control power. Such occurrences 
are distinct from events termed ``accidents,'' such as a loss-of-
coolant accident (LOCA) or a main steamline break. The references to 
fuel integrity requirements related to accidents and those regarding 
emergency core cooling system (ECCS) performance are beyond conditions 
of normal operation.
    The UCS report relates other regulations beyond the GDC to fuel 
integrity during normal operation as follows:

 10 CFR 50.34a, ``Design objectives for equipment to control 
releases of radioactive material in effluents--nuclear power reactors''
 10 CFR 50.36, ``Technical specifications''
 10 CFR 50.59, ``Changes, tests and experiments''
 10 CFR 50.71, ``Maintenance of records, making of reports''
 Appendix I to 10 CFR Part 50, ``Numerical Guides for Design 
Objectives and Limiting Conditions for Operation To Meet the Criterion 
`As Low As Is Reasonably Achievable' for Radioactive Material in Light-
Water-Cooled Nuclear Power Reactor Effluents''

    Although 10 CFR 50.36a, ``Technical specifications on effluents 
from nuclear power reactors,'' was not directly referenced in the 
report, by citing 10 CFR 50.36, the staff inferred that Section 50.36a 
is linked to fuel integrity when considering the discussion on the UCS 
report.
b. NRC Staff Guidance Documents
    To implement NRC regulations, several NRC staff guidance documents 
are used, including the following:

 Regulatory Guide 1.3, ``Assumptions Used for Evaluating the 
Potential Radiological Consequences of a Loss

[[Page 20332]]

of Coolant Accident for Boiling Water Reactors''
 Regulatory Guide 1.4, ``Assumptions Used for Evaluating the 
Potential Radiological Consequences of a Loss of Coolant Accident for 
Pressurized Water Reactors''
 Regulatory Guide 1.77, ``Assumptions Used for Evaluating a 
Control Rod Ejection Accident for Pressurized Water Reactors''
 Regulatory Guide 1.112, ``Calculation of Releases of 
Radioactive Materials in Gaseous and Liquid Effluents From Light-Water-
Cooled Power Reactors''
 SRP Section 4.2, ``Fuel System Design''
 SRP Section 4.4, ``Thermal and Hydraulic Design''
    Along with the regulations, licensees use the guidance documents 
listed above to form the licensing basis for fuel integrity at their 
plant. The licensing basis for a nuclear power plant, as defined in 10 
CFR Part 54, ``Requirements for Renewal of Operating Licenses for 
Nuclear Power Reactors,'' is ``the set of NRC requirements applicable 
to a specific plant and a licensee's written commitments for ensuring 
compliance with and operation within applicable NRC requirements and 
the plant-specific design basis * * * that are docketed and in 
effect.'' The definition continues by listing elements of the licensing 
basis, such as technical specifications, the FSAR, and licensee 
commitments documented in NRC safety evaluations. Several components 
form the plant's licensing basis for fuel performance: (1) NRC 
regulations that specifically refer to fuel integrity; (2) technical 
specification limits on coolant activity; (3) fuel rod performance 
specifications and analysis assumptions defined in the plant's FSAR and 
referenced topical reports; and (4) commitments to NRC regulatory 
guidance and to generic communications addressing fuel performance.
    Acceptance criteria in the SRP sections, which may be adopted by 
licensees to implement the regulations, are based on meeting the 
requirements of GDC 10 with appropriate margin to ensure that SAFDLs 
are not exceeded during normal operation, including AOOs. Specifically, 
SRP Section 4.2 has as an objective of the safety review ``to provide 
assurance that the fuel system is not damaged as a result of normal 
operation and anticipated operational occurrences.'' The reviewer 
should ensure that fuel does not leak as a result of specific causes 
during normal operation and AOOs, and that leaking fuel is accounted 
for in the dose analyses for postulated design-basis accidents. 
Further, fuel rod failure is defined in SRP Section 4.2 as ``the loss 
of fuel rod hermiticity,'' meaning fuel rod leakage. However, in SRP 
Section 4.2, the staff also states that ``it is not possible to avoid 
all fuel rod failures and that cleanup systems are installed to handle 
a small number of leaking rods.'' Such leaks typically occur as a 
result of manufacturing flaws or loose parts wear. Therefore, on the 
basis of this review guidance, the staff accepts the possibility that 
fuel may leak during normal operation.
    In the case of the Calvert Cliffs Nuclear Plant, a plant cited as 
an example in the UCS report, the plant's licensing basis contains a 
commitment to adhere to the guidance in the SRP. The following four 
objectives for fuel design given in SRP Section 4.2 may be used as fuel 
design objectives within a plant's licensing basis as is done in the 
Calvert Cliffs FSAR:

 Fuel is not damaged as a result of normal operation and AOOs.
 Fuel damage is never so severe as to prevent control rod 
insertion when required.
 The number of fuel rod failures is not underestimated for 
postulated accidents.
 Coolability is always maintained.

    SRP Section 4.4 has as an objective that the thermal and hydraulic 
design of the core should provide acceptable margins of safety from 
conditions that would lead to fuel damage during normal reactor 
operation, including anticipated operational transients. It gives two 
examples of acceptable approaches to meet the acceptance criteria: one 
based on a 95-percent probability at a 95-percent confidence level that 
the hottest rod in the core does not exceed prescribed thermal limits 
during normal operation, including AOOs, and the other using a limiting 
value for thermal limits so that at least 99.9 percent of the fuel rods 
are not expected to exceed thermal limits during normal operation, 
including AOOs. These criteria are limits that strive to maintain a 
very low likelihood of fuel damage during operation; however, they do 
not preclude the possibility that some fuel defects could occur.
    A plant's licensing basis contains fuel performance criteria that 
are specified for normal operation, including AOOs, and analyses are 
conducted to ensure that these criteria will not be exceeded. The 
criteria are related to the SAFDLs mentioned in the GDC and are 
normally presented in terms of prescribed thermal limits, which can be 
calculated and are reliable predictors of the onset of fuel damage. For 
boiling-water reactors (BWRs), critical heat flux or the critical power 
ratio is used as the predictor of fuel damage onset, and for 
pressurized-water reactors (PWRs), the criterion is the departure from 
nucleate boiling (DNB), or the DNB ratio (DNBR).
    An example of fuel design limits given in plant documentation is 
found in the FSAR for Calvert Cliffs Units 1 and 2. Section 3.6 of the 
FSAR presents fuel design and analysis bases. Fuel rod cladding is 
designed to stress and strain limits, considering the operating 
temperature, the cladding material, the expected property changes as a 
result of irradiation, and the predicted life span of the fuel. 
Extensive fuel mechanical analyses are detailed, along with pertinent 
fuel test data, which help to confirm the analysis results. The 
calculations are used to demonstrate that the criteria are satisfied 
for limiting cases under limiting assumptions. Chapter 14 of the 
Calvert Cliffs FSAR gives the fuel behavior acceptance criteria for 
each category of design-basis event analyzed. For AOOs, the minimum 
DNBR is chosen to provide at least a 95-percent probability with a 95-
percent confidence level that DNB will not be experienced along the 
fuel rod with that DNBR (i.e., the SRP Section 4.4 criteria). This 
limit ensures that there is a low probability of fuel rod damage as a 
result of overheated cladding. The fuel temperature SAFDL is set so 
that no significant fuel melting will occur during steady-state 
operation or during a transient. Compliance with the limit offers 
assurance that the fuel rod will not be damaged as a result of material 
property changes and increases in fuel pellet volume, which could be 
associated with fuel melting. Again, as with the limits discussed in 
SRP Section 4.4, these limits are set to prevent fuel damage, but the 
possibility of fuel leakage is recognized.
    The key to plant licensing bases regarding fuel integrity is the 
technical specification limiting the concentration of activity allowed 
in reactor coolant during plant operation. These limits are based on 
maintaining a margin to the dose guidelines in 10 CFR Part 100 for 
steam generator tube rupture (SGTR) accidents in PWRs and main 
steamline break (MSLB) accidents in BWRs. The specific activity limits 
of the reactor coolant system are stated in terms of dose equivalent 
iodine-131, which is attributable solely to fuel leaks. That is 
distinct from gross coolant activity, which is the aggregate activity 
from all sources, including fuel leaks and corrosion product 
activation. The technical basis for these limits can be traced to the 
guidance given in

[[Page 20333]]

Appendix I, which is, in turn, based on assumptions that fuel leaks 
would exist during operation. Technical specifications for reactor core 
safety limits, including the reactor protection system setpoints, are 
set so that the SAFDLs are not exceeded during normal operation or 
AOOs. The technical specifications for protection system action are 
intended to prevent fuel damage, but the specifications for coolant 
activity levels recognize that some small amount of fuel leakage is 
allowable during operation. The technical specifications concerning 
coolant activity are based on meeting the dose acceptance criteria in 
the SRP for the limiting design-basis accident (usually SGTR or MSLB 
for PWRs and MSLB for BWRs). These limits are used as assumptions in 
design-basis accident dose analyses to show compliance with dose 
acceptance criteria for the control room operators and the public. By 
maintaining the levels of coolant activity within these limits during 
normal operation, the continued validity of the design-basis analyses 
is maintained.
    The staff has addressed fuel performance problems in several 
generic communications to licensees. Prominent among these were NRC 
Information Notice (IN) 93-82, ``Recent Fuel and Core Performance 
Problems in Operating Reactors,'' and Generic Letter (GL) 90-02, 
``Alternative Requirements for Fuel Assemblies in Design Features 
Section of Technical Specifications.'' In IN 93-82, the staff discussed 
fuel leaks occurring during normal operation from a specific cause--
fretting wear in PWR fuel, which was partly attributed to mixed fuel 
core designs. The staff alerted licensees to the introduction of 
modified fuel designs that requires added attention to ensure that the 
core design basis is not violated. This information notice is an 
example of staff action to use operating information gathered from fuel 
leaks at a few plants to avoid similar problems at other reactors, thus 
reducing the potential for more widespread fuel leakage. In GL 90-02, 
the staff provided licensees with added flexibility to take actions to 
reduce fission-product releases during operation by removing defective 
fuel rods during refueling outages.
    The staff has previously considered the safety implications of 
operation with fuel leakage on a generic basis. Generic Safety Issue 
(GSI) B-22, ``LWR [Light Water Reactor] Fuel,'' which is related to 
fuel leakage, is discussed in NUREG-0933, ``A Prioritization of Generic 
Safety Issues,'' Supplement 22, March 1998. In GSI B-22, the staff 
considered the ability to accurately predict fuel performance under 
normal and accident conditions. The GSI review was conducted to 
determine if predictions of fuel behavior under normal operating and 
accident conditions were sufficient to demonstrate that regulatory 
requirements were being met. In its evaluation of the issue, the staff 
concluded that releases during normal operation would be increased 
because of fuel defects, but would not be increased beyond regulatory 
limits. The staff also stated that, ``additional requirements would not 
decrease the number of fuel defects significantly.'' Furthermore, the 
staff concluded that the release from fuel damaged during design-basis 
accidents and severe accidents would be much larger than the release 
attributed to preexisting fuel defects, and the magnitude of the 
release would not be significantly affected by preexisting fuel 
defects. Thus, the consequence from leaking fuel was determined to be 
very small. The staff concluded that because fuel manufacturers have 
taken an active role to improve fuel performance, fuel leaks are now 
rare, and the significance of the issue has diminished. Therefore, the 
issue was dropped from further consideration.
    In the resolution of GSI B-22, the staff concluded that the 
influence of additional restrictions to operation with fuel leaks on 
core damage frequency and public consequence would be insignificant. 
Thus, operation with a limited number of fuel defects and leaks under 
normal operating conditions is not associated with an excessive level 
of risk, provided that the plant continues to operate within technical 
specifications limits for reactor coolant activity.
3. Evaluation of Generic Concerns
    The staff evaluated the generic concerns associated with fuel 
leakage identified previously by the Petitioner, as follows:
a. 10 CFR 50.59, ``Changes, Tests and Experiments''
    A premise of the UCS report is that 10 CFR 50.59 is violated 
because reactor operation with limited fuel leakage constitutes an 
unapproved change to the licensing basis for a plant. The report states 
that ``Federal regulations require formal NRC approval prior to any 
nuclear plant operating with fuel cladding failures.'' The attachment 
to the report is an assessment of operation with fuel leaks as an 
unreviewed safety question on the basis of the criteria in 10 CFR 
50.59. The report states that such operation is an unreviewed safety 
question because operation with leaking fuel (1) increases the 
probability and consequences of an accident, (2) creates an accident 
different from any in the safety analysis for the plant, and (3) 
reduces safety margins.
    The staff does not agree that operation with leaking fuel 
necessarily constitutes a change to or violation of the licensing basis 
for a plant. A small amount of fuel leakage during operation is 
permitted by NRC staff guidance implementing NRC regulations and is 
accounted for in plant licensing bases. A key component of the 
licensing basis regarding fuel performance is the technical 
specification limiting reactor coolant system activity. The fission-
product release from the level of leaking fuel associated with the 
technical specification limit is included in the design-basis accident 
dose analyses described in the FSAR for a plant to show compliance with 
the dose acceptance criteria in the SRP. Therefore, operating with 
leaking fuel, within the coolant activity technical specification 
limits, does not constitute a change in the plant licensing basis, and 
10 CFR 50.59 does not apply.
b. 10 CFR 50.71, ``Maintenance of Records, Making of Reports''
    The Petitioner states in the report that ``any plant operating with 
fuel cladding failures is violating its design and licensing bases 
requirements, a condition not allowed by Federal safety regulations.'' 
The Petitioner further states that when plants operate with leaking 
fuel, 10 CFR 50.71 is violated since the licensing basis for a plant, 
as documented in the technical specifications and in the analyses 
contained in the FSAR, does not accommodate such operation.
    This concern is closely linked to the previous discussion regarding 
10 CFR 50.59, in that FSARs for plants operating with leaking fuel 
should, in the view of the UCS, include safety analyses accounting for 
the effects of fuel leaks. As previously discussed, plant licensing 
bases do incorporate assumptions for limited levels of fuel leakage 
through technical specifications requirements and designs for plant 
reactor water cleanup systems. Plant FSARs, including the example 
discussed earlier in this evaluation, typically contain information on 
fuel leakage effects, and the safety analyses explicitly allow for 
coolant activity levels attributable to leaking fuel under normal 
operation. Thus, the staff does not consider 10 CFR 50.71 to be 
violated by operation with fuel leakage.
c. Safety Analysis Assumptions
    The UCS report states that ``safety analyses assume that all three 
barriers

[[Page 20334]]

[to radioactive material release] are intact prior to any accident.'' 
Therefore, according to the UCS, plants with known fuel leakage could 
have accidents with more severe consequences than predicted. The report 
also states the following: ``Pre-existing fuel cladding failures have 
not been considered in the safety analyses for this accident [LOCA], or 
any other accident.''
    In the discussion that follows, the staff explains that preexisting 
fuel cladding leaks are accounted for in plant licensing bases and that 
safety analyses do not assume that all the fission-product barriers are 
fully intact before an accident.
    The analyses of limiting postulated design-basis releases do not 
assume that all the fission-product barriers are fully intact before an 
accident. For the loss-of-coolant accident, which typically yields the 
most limiting postulated releases, all three barriers are assumed to 
allow the release of some fission products. The methodology used to 
analyze this accident is given in Regulatory Guides 1.3 and 1.4, and 
SRP Section 15.6.5, ``Loss-of-Coolant Accidents Resulting From Spectrum 
of Postulated Piping Breaks Within the Reactor Coolant Pressure 
Boundary.''
    For the containment and reactor coolant system (RCS) barriers, 
these assumptions are explicitly given. The containment is assumed to 
leak at the leak rate incorporated in the plant technical 
specifications when the containment is at positive pressure. The RCS 
inside the containment is assumed to completely fail as a fission-
product barrier at the beginning of the accident. Systems outside the 
containment that interface with the RCS are also assumed to experience 
failures.
    The assumption of preexisting leakage for the fuel cladding 
barrier, although not explicitly given, is inherent in the assumption 
of a conservative nonmechanistic release from the fuel. The entire 
iodine and noble gas inventory of the core is assumed to be released to 
the reactor coolant. A conservative fraction of this inventory is 
assumed to be released into the containment and subsequently released 
to the environment. Assuming that this release occurs instantaneously 
further enhances the conservatism of these analyses. This assumption 
disregards the fission-product containment function of the fuel 
cladding at the beginning of the accident.
    Accidents, which may not be bounded by the radiological 
consequences of a LOCA, include the control rod drop accident for BWRs 
and MSLB outside of containment for PWRs. However, the conservatism of 
the source term assumptions for these analyses parallels those for a 
LOCA. Some of the same assumptions used for radiological consequence 
evaluation of a LOCA are used for the analysis of MSLB outside of 
containment. Appendix A to SRP Section 15.1.5, ``Radiological 
Consequences of Main Steam Line Failures Outside Containment of a 
PWR,'' contains an acceptance criterion that references Regulatory 
Guide 1.4. The radiological assumptions for the control rod drop 
analysis are similar to those for a LOCA, as stated in Appendix A to 
SRP Section 15.4.9, ``Radiological Consequences of Control Rod Drop 
Accident (BWR),'' and Regulatory Guide 1.77. For example, the 
guidelines assume that the nuclide inventory in the potentially 
breached fuel elements should be calculated and it should be assumed 
that all gaseous constituents in the fuel cladding gaps are released.
    The radioactivity assumed for release from the LOCA is much greater 
than that associated with preexisting fuel leakage allowed by plant 
technical specifications. The staff has compared releases from 
preexisting defects with the release resulting from fuel damage during 
an accident. In its consideration of GSI B-22, the staff concluded 
that, ``the magnitude of a release from failed fuel during an accident 
is much larger than the release from a preexisting fuel defect'' and 
that ``the resultant consequence from failed fuel was determined to be 
very small'' (NUREG-0933). These assumptions are made despite the 
provisions of 10 CFR 50.46 requiring an ECCS that must be designed to 
prevent exceeding thermal limits that cause such gross fuel failure. In 
addition, for design-basis accidents in which fuel damage is not 
assumed, the preexisting fuel cladding defects are typically assumed to 
serve as release paths facilitating a spike in radioiodine 
concentration in the coolant.
    Additional NRC fuel design requirements complement the conservative 
defense-in-depth assumptions as previously described to prevent an 
unanalyzed large release of fission products. To illustrate its concern 
about fuel leakage influences on accident progression, the UCS report 
describes a LOCA sequence and postulates that hydraulic loads on the 
fuel rods could lead to cladding failures, which would result in a 
large release of fission products into the coolant and prevent control 
rod insertion. Fuel design requirements and guidance specifically 
address the ability to insert control rods, and staff review guidance 
recognizes that preexisting fuel cladding defects could have an effect 
on fuel performance during accidents. In GDC 27, ``Combined Reactivity 
Control Systems Capability,'' the staff requires that reactivity 
control systems, including the control rod system, have the capability 
to control reactivity changes under postulated accident conditions in 
order to assure core cooling. SRP Section 4.2 includes the objective 
that ``fuel system damage is never so severe as to prevent control rod 
insertion when it is required.''
    To ensure that the preceding objective is met, fuel designs 
consider external loads on fuel rods. This is discussed in the appendix 
to SRP Section 4.2, ``Evaluation of Fuel Assembly Structural Response 
to Externally Applied Forces.'' The basis for much of the appendix to 
SRP Section 4.2 is contained in NUREG/CR-1018, ``Review of LWR Fuel 
System Mechanical Response With Recommendations for Component 
Acceptance Criteria,'' prepared by EG&G Idaho in September 1979. This 
report states that ``Cyclic fatigue and material degradation may cause 
a failure [of a fuel system component] at any point in the transient 
[i.e., a LOCA].'' Thus, material degradation that could lead to fuel 
leakage during operation is considered in accident analyses. 
Furthermore, design considerations, such as control guide tubes in PWRs 
and fuel channel boxes in BWRs, help separate control rods from the 
fuel. The separation provided protects control rods from material 
degradation of fuel that might occur in accidents, thus helping to 
prevent control rod obstruction. Such safety analysis assumptions as 
these (which assume preexisting failures of the fission-product 
barriers) provide confidence that the preexisting cladding defects 
allowed by technical specifications limits on coolant activity will not 
erode the safety margin assumed for accident analyses.
d. 10 CFR 50.34a, ``Design Objectives for Equipment To Control Releases 
of Radioactive Material in Effluents--Nuclear Power Reactors''
    In its report, the UCS claims that 10 CFR 50.34a and other 
regulations related to the ALARA principle for radioactive materials 
release are violated since plant workers are exposed to a greater risk 
than necessary because of higher coolant activity levels attributable 
to leaking fuel. The UCS report continues: ``Federal regulations 
require nuclear plant owners to keep the release of radioactive 
materials as low as reasonably achievable. Therefore, it is both an 
illegal activity and a serious health hazard for nuclear plants to 
continue operating with fuel cladding

[[Page 20335]]

damage.'' The UCS report cites Appendix I to 10 CFR Part 50 when 
contending that fuel releases pose an undue risk to plant workers. 
Appendix I contains the numerical dose guidelines for power reactor 
operation to meet the ALARA criterion. These dose values are a small 
fraction of the 10 CFR Part 20 annual public dose limit of 100 millirem 
(i.e., 3 millirem from liquid effluents and 5 millirem from gaseous 
effluents).
    The bases for the guidelines in Appendix I are given in WASH-1258, 
which acknowledges that radioactive material from a number of sources, 
including fission-product leakage to the coolant from defects in the 
fuel cladding, will be present in the primary coolant during normal 
operation. Further, in the ``Bases'' section on RCS specific activity 
in NUREG-1431, ``Standard Technical Specifications, Westinghouse 
Plants,'' April 1995, the limits on specific activity are linked to 
exposure control practices at plants. The section clearly states that 
the limits on RCS specific activity are used in the design of radiation 
shielding and plant personnel radiation protection practices.
    In addition, occupational dose considerations were discussed in the 
resolution of GSI B-22. The staff acknowledged that localized dose 
rates were expected to increase as a result of fuel defects, but 
effects are limited by requirements for plants to operate within their 
technical specifications for coolant activity and releases. In some 
cases, plants will often stay within allowable release limits and 
coolant activity levels by operating at reduced power until the next 
refueling outage allows the problem to be corrected.
    On the basis of the preceding discussion, operation with a limited 
amount of leaking fuel is within a plant's licensing basis and, in 
itself, does not violate ALARA-related regulations. Operation involving 
leaking fuel, however, will likely require plant operators to take 
additional measures in order to ensure that ALARA requirements are 
being met, but these would need to be considered on a case-by-case 
basis.
4. UCS Report Recommendations
    In the report, the UCS recommends that the NRC take steps to 
prohibit nuclear power plants from operating with fuel cladding damage 
until the safety concerns raised by the report are resolved. The 
following steps are specifically recommended: (1) requiring plant 
shutdown upon detection of fuel leakage, and (2) requiring that safety 
evaluations that consider the effects of operating with leaking fuel be 
included in plant licensing bases to justify operation under such 
circumstances. Further, the UCS recommends that UFSARs be revised to 
establish safe operating limits to accommodate operation with leaking 
fuel.
    On the basis of the staff's consideration of the stated safety 
concerns in the report, there is no technical or regulatory basis to 
require that plants operating with leaking fuel be shut down, provided 
they are operating within their technical specifications limits and in 
accordance with their licensing basis. The UCS report, in raising its 
concerns, does not offer any new information to demonstrate that the 
overall risk of operating with fuel defects presents an undue hazard to 
plant workers or the public.
    Further, since the staff does not consider plants operating with 
leaking fuel to be violating 10 CFR 50.59 or 50.71, there is no basis 
for requiring plants to perform additional safety analyses to model the 
effects of fuel defects on accident progressions to update plant safety 
analysis documentation.

B. Plant-Specific Concerns--River Bend Station

    On the basis of the reported fuel leakage at River Bend, the 
Petitioner states that the generic concerns contained in its report 
apply to River Bend. The September 25, 1998, Petition then presents a 
number of references to the River Bend USAR as instances in which, in 
the opinion of the Petitioner, plant licensing bases do not permit 
operation of the plant with known fuel leakage.
    A reference to the USAR in the Petition is the USAR definition of 
unacceptable consequences (USAR Table 15A.2-4), which lists as an 
unacceptable consequence ``Failure of the fuel barrier as a result of 
exceeding mechanical or thermal limits.'' The Petitioner considers this 
criterion violated since a fuel failure exists in advance of any 
design-basis accident that may now occur.
    The Petition then discusses USAR Chapter 15 accident analysis 
descriptions, which state either (1) that fuel cladding integrity will 
be maintained as designed or (2) radioactive material is not released 
from the fuel for the event. The following events cited in the Petition 
have event descriptions in the River Bend USAR, which state that fuel 
cladding will function and maintain its integrity as designed:

 Loss of Feedwater Heating (USAR Section 15.1.1.4)
 Feedwater Controller Failure--Maximum Demand (USAR Section 
15.1.2.4)
 Pressure Regulator Failure--Open (USAR Section 15.1.3.4)
 Pressure Regulator Failure--Closed (USAR Section 15.2.1.4)

    The following two events cited in the Petition have event 
descriptions in the River Bend USAR, which state that ``no radioactive 
material is released from the fuel'' during the event:
 Control Rod Withdrawal Error at Power (USAR Section 15.4.2.5)
 Recirculation Flow Control Failure with Increasing Flow (USAR 
Section 15.4.5.5)
    The Petitioner also states that the River Bend licensing basis for 
worker radiation protection is violated by operation with leaking fuel. 
Again, the Petition cites the USAR (Sections 12.1.1 and 12.1.2.1) as 
the pertinent reference to the licensing basis.
Evaluation of Plant-Specific Concerns
    As discussed in the consideration of generic safety concerns, the 
staff does not agree that preexisting fuel cladding defects and 
resultant fuel leakage violate plant licensing bases. The staff also 
considers that conclusion valid for River Bend. The basis for this 
conclusion is supported in the following discussion.
a. USAR Appendix 15A
    The Petitioner referenced two sections of USAR Appendix 15A, 
``Plant Nuclear Safety Operational Analysis (NSOA)'' (as stated):
    UFSAR 15A.2.8, ``General Nuclear Safety Operational Criteria,'' 
stated:
    The plant shall be operated so as to avoid unacceptable 
consequences.

    UFSAR Table 15A.2-4, ``Unacceptable Consequences Criteria Plant 
Event Category: Design Basis Accidents,'' defined ``unacceptable 
consequences'' as follows:

4-1  Radioactive material release exceeding the guideline values of 
10CFR100.
4-2  Failure of the fuel barrier as a result of exceeding mechanical 
or thermal limits.
4-3  Nuclear system stresses exceeding that allowed for accidents by 
applicable industry codes.
4-4  Containment stresses exceeding that allowed for accidents by 
applicable industry codes when containment is required.
4-5  Overexposure to radiation of plant main control room personnel.

    The current operating condition at the River Bend Station 
apparently violates the spirit, if not the letter, of Criterion 4-2 
since the fuel barrier has already failed, albeit to a limited 
extent. This UFSAR text does not accept a low level of fuel barrier 
failure based on meeting the offsite and onsite radiation protection 
limits. Integrity of the fuel barrier is an explicit criterion in 
addition to the radiation requirements.


[[Page 20336]]


    In the Petition, the UCS highlights the table concerning the 
consequences for the design-basis accident. This plant condition is a 
highly improbable event, and safety analyses ensure that safety limits 
and regulatory requirements are not exceeded as a result of the 
accident occurring. This is why USAR Table 15A.2-4, Item 4-2 states, 
``Failure of a fuel barrier as a result of exceeding mechanical or 
thermal limits'' (emphasis added). The unacceptable consequences of 
this type of event are independent of preexisting fuel cladding 
defects. The unacceptable consequences of this event are additional 
fuel failures as a result of the accident occurring.
    Within the framework of the USAR, ``unacceptable consequences'' are 
specified measures of safety and analytically determinable limits on 
the consequences of different classifications of plant events. They are 
used for performing a nuclear safety operational analysis. Unacceptable 
consequences are described for various plant conditions, including 
``Normal (Planned) Operation,'' ``Anticipated (Expected) Operational 
Transients,'' ``Abnormal (Unexpected) Operational Transients,'' 
``Design Basis (Postulated) Accidents,'' and ``Special (Hypothetical) 
Events.'' USAR Tables 15A.2-1 through 15A.2-5 identify the unacceptable 
consequences for each of the five plant conditions, and are different 
for each of the cases.
    The USAR text clearly documents the acceptability of a low level of 
fuel cladding failures based on meeting the offsite and onsite 
radiation protection limits. For example, USAR Table 15A.2-1 discusses 
the unacceptable consequences for normal operation. This USAR table 
defines unacceptable consequences for normal operation as follows:

4-1  Release of radioactive material to the environs that exceeds 
the limits of either 10 CFR Part 20 or 10 CFR Part 50.
4-2  Fuel failure to such an extent that were the freed fission 
products released to the environs via the normal discharge paths for 
radioactive material, the limits of 10 CFR Part 20 would be 
exceeded.
4-3  Nuclear system stress in excess of that allowed for planned 
operation by applicable industry codes.
4-4  Existence of a plant condition not considered by plant safety 
analysis.

    Item 4-2 in Table 15A.2-1 implies that fuel cladding failures are 
not an unanticipated condition during normal operations and is, 
therefore, consistent with other parts of the River Bend licensing-
basis. Fuel cladding defects are acceptable to the extent that they do 
not jeopardize radiation protection limits established in the plant 
technical specifications and other licensing-basis documents. USAR 
Table 15A.2-4 does not apply for normal operations; only USAR Table 
15A.2-1 applies. Furthermore, the provisions found in USAR Table 15A.2-
4 would continue to be met for postulated design-basis accidents.
    USAR Section 15.0.3.1.1 provides further clarification in its list 
of unacceptable safety consequences for ``moderate frequency'' events, 
which lists: ``Reactor operation induced fuel-cladding failure as a 
direct result of the transient analysis above the minimum critical 
power ratio (MCPR) uncertainty level (0.1 percent).'' Accordingly, 
preexisting cladding defects are considered during some postulated 
transients. In fact, the acceptance criteria for moderate-frequency 
event analyses, based on the GDC (10 CFR Part 50, Appendix A) and the 
Standard Review Plan, and described in the Safety Evaluation Report 
(SER) for River Bend (NUREG-0989), state the following expectations for 
fuel cladding performance: ``An incident of moderate frequency . . . 
should not result in a loss of function of any fission product barrier 
other than the fuel cladding. A limited number of fuel rod cladding 
perforations are acceptable.''
    USAR Chapter 11, ``Radioactive Waste Management,'' Section 11.1, 
``Source Terms,'' details the expected reactor coolant and main steam 
activities to be used to form the basis for estimating the average 
quantity of radioactive material released to the environment during 
normal operations, including operational occurrences. This section 
further addresses that the offgas release rate of 304,000 Ci/
sec at a 30-minute delay time corresponds to design failed fuel 
conditions, that is, maximum acceptable cladding failure for normal 
operation, and is also conservatively based upon 105 percent of rated 
thermal power. This is consistent with limits prescribed in Technical 
Specification 3.7.4, ``Main Condenser Offgas,'' which requires that the 
gross gamma activity rate of the noble gases shall be <290 mCi/sec (or 
<290,000 Ci/sec) after a decay time of 30 minutes.
    In addition, two other parts of the fuel system licensing basis for 
River Bend show that limited fuel leakage during plant operation is a 
design consideration:
    The fuel system design basis for River Bend is given in USAR 
Section 4.2.1 by reference to the generic topical report ``General 
Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A. The 
generic topical report details fuel cladding operating limits to ensure 
that fuel performance is maintained within fuel rod thermal and 
mechanical design and safety analysis criteria. The limits are given 
for normal operating conditions and AOOs in terms of specific 
mechanical and thermal specifications. Evaluations of specific fuel 
failure mechanisms under normal operation and AOOs were discussed, such 
as stress/strain, hydraulic loads, fretting, and internal gas pressure 
to ensure that fuel failure did not result from these causes. The 
design basis did not preclude the possibility that fuel could fail for 
other reasons, such as preexisting cladding flaws leading to leakage.
    The Technical Specifications (Section 3.4.8) for River Bend contain 
a limit for reactor coolant system specific activity. The basis for 
this limit is the same as that discussed in the consideration of the 
generic safety concerns. Section B 3.4.8 of the River Bend Technical 
Specifications ``Bases'' acknowledges that ``the reactor coolant 
acquires radioactive materials due to release of fission products from 
fuel leaks.'' Thus, fission products released during plant operation 
are clearly considered to be contributors to the source term used for 
safety analysis of the MSLB release consequences. The Technical 
Specifications state that the limit is set to ensure that any release 
as a consequence of an MSLB is less than a small fraction of the 10 CFR 
Part 100 guidelines. These portions of the River Bend licensing basis 
are consistent with NRC regulations regarding fuel performance and the 
associated NRC guidance used by licensees to implement those NRC 
regulations that were covered earlier in the discussion regarding 
generic concerns.
    The River Bend licensing-basis items listed by the Petitioner are 
consistent with the parts of the fuel licensing basis discussed above 
with the exception of some minor inconsistencies in documentation (as 
discussed below). That is, fuel leakage during plant operation is not 
precluded by licensing-basis provisions requiring that fuel integrity 
be maintained as designed. The design basis itself allows the 
possibility of leakage while ensuring that cladding damage does not 
result from specific operationally related causes. Fuel is also 
designed to maintain its structural integrity to ensure core 
coolability and to ensure that control rods can be inserted.
b. Chapter 15 Accident Analysis
    The Petitioner also cited references taken from accident analyses 
described in River Bend USAR Chapter 15 (as stated):


[[Page 20337]]


    UCS reviewed the UFSAR Chapter 15 description of accident 
analyses performed for the River Bend Station. UFSAR Section 
15.1.1.4, ``Barrier Performance,'' for the loss of feedwater heating 
event stated:
    The consequences of this event do not result in any temperature 
or pressure transient in excess of the criteria for which the fuel, 
pressure vessel, or containment are designed; therefore, these 
barriers maintain their integrity and function as designed.
    UFSAR Sections 15.1.2.4 for the feedwater controller failure--
maximum event, 15.1.3.4 for the pressure regulator failure--open 
event, and 15.2.1.4 for the pressure regulator failure--closed event 
all contain comparable statements that barrier performance was not 
performed because the fuel remained intact.
    These analyzed events appear to be valid only when the River 
Bend Station is operated with no failed fuel assemblies. Operation 
with pre-existing fuel failures (i.e., the current plant 
configuration) appear to be outside of the design and licensing 
bases for these design bases events.
    UFSAR Section 15.4.2.5, ``Radiological Consequences,'' for the 
control rod withdrawal error at power event stated:
    An evaluation of the radiological consequences was not made for 
this event since no radioactive material is released from the fuel.
    UFSAR Section 15.4.5.5, ``Radiological Consequences,'' for the 
recirculation flow control failure with increasing flow event 
stated:
    An evaluation of the radiological consequences is not required 
for this event since no radioactive material is released from the 
fuel.
    These analyzed events also appear valid only when the River Bend 
Station is operated with no failed fuel assemblies. Operation with 
pre-existing fuel failures (i.e., the current plant configuration) 
appear to be outside of the design and licensing bases for these 
design bases events.
    The effect from pre-existing fuel failures was considered, at 
least partially, for one design bases event. UFSAR Section 
15.2.4.5.1, ``Fission Product Release from Fuel,'' for the main 
steam isolation valve closure event stated:
    While no fuel rods are damaged as a consequence of this event, 
fission product activity associated with normal coolant activity 
levels as well as that released from previously defective rods is 
released to the suppression pool as a consequence of SRV [safety 
relief valve] actuation and vessel depressurization.
    The aforementioned design bases events (e.g., control rod 
withdrawal error at power, loss of feedwater heating, et al.) are 
not bound by these results because the radioactive material is not 
``scrubbed'' by the suppression pool water as it is in the MSIV 
[main steam isolation valve] closure event.

    As previously stated, the Petitioner cited four references to the 
USAR accident analysis section entitled ``Barrier Performance.'' At 
issue are essentially equivalent statements made where the USAR stated, 
in part, that the defense-in-depth ``barriers maintain their integrity 
and function as designed.'' The UCS concluded that operation with 
preexisting fuel failures is, therefore, outside the River Bend design 
and licensing bases. In stating that barriers are ``maintained,'' the 
USAR clearly implies that the events themselves do not result in 
additional fuel cladding failures. To further support this conclusion, 
the radiological consequences described for three of the four events 
(Section 15.1.2, ``Feedwater Controller Failure--Maximum Demand''; 
Section 15.1.3, ``Pressure Regulator Failure--Open''; and Section 
15.2.1, ``Pressure Regulator Failure--Closed'') are, indeed, bounded by 
an event that takes into consideration the effects of preexisting 
cladding failures. The three preceding events all result in actuation 
of the safety relief valves (SRVs) to the suppression pool. The USAR 
discussion (see USAR section titled ``Radiological Consequences'') 
notes that radioactivity is discharged to the suppression pool, and 
that the activity discharged is much less than those consequences 
identified in USAR Section 15.2.4.5 (for the MSIV closure event).
    The MSIV closure event, as described in the USAR, clearly considers 
the activity released from ``previously defective rods'' in determining 
dose consequences. The source term used in these calculations assumes 
the same iodine and noble gas activity as an initial condition as is 
used in the basis for determining RCS activity technical specifications 
limits. USAR Section 15.2.4.5.1, ``Fission Product Release from Fuel,'' 
also explains, ``Since each of those transients identified previously 
which cause SRV actuation results in various vessel depressurization 
and steam blowdown rates, the transient evaluated in this section [the 
MSIV closure event] is that one which maximizes the radiological 
consequences for all transients of this nature.'' Thus, the USAR 
explicitly describes how ``the aforementioned design-basis events'' are 
bounded by the results for the MSIV closure event, for those events 
resulting in an SRV actuation. Furthermore, USAR Section 15.1.1.5 
describing the fourth event, the loss of feedwater heating, also states 
that ``this event does not result in any additional fuel failures,'' 
further reinforcing the staff's position.
    The quotation taken from the control rod withdrawal error from 
power and recirculation flow control error event descriptions--``[a]n 
evaluation of the radiological consequences was not made for this event 
since no radioactive material is released from the fuel''--appears to 
be taken out of context. Considering the many references ostensibly 
permitting operation with preexisting fuel cladding failures found 
within the USAR, technical specifications, NRC regulations, staff 
implementing guidelines, and other licensing-basis documents, the 
intent of this statement is clearly that no additional radioactive 
material is released from the fuel as a consequence of the event.
    Finally, in each of the accident analysis cases listed in the 
Petition, the event is classified as a ``moderate frequency'' event (or 
an ``anticipated operational transient''). Specific criteria for 
unacceptable consequences are delineated in USAR Table 15A.2-2. For 
this type of anticipated transient, unacceptable performance of the 
fuel is described as, ``[r]eactor operation induced fuel cladding 
failure as a direct result of the transient analysis above the MCPR 
[Minimum Critical Power Ratio] uncertainty level (0.1%)'' (emphasis 
added). Therefore, fuel cladding defects existing before the accident 
are not precluded from consideration.
c. Fuel Cladding Defect Propagation
    The Petition then raised concerns regarding the possibility that 
preexisting fuel cladding defects could propagate under design-basis 
transients (as stated):

    As detailed in UCS's April 1998 report on reactor operation with 
failed fuel cladding, it has not been demonstrated that the effects 
from design basis transients and accidents (i.e., hydrodynamic 
loads, fuel enthalpy changes, etc.) prevent pre-existing fuel 
failures from propagating. It is therefore possible that 
significantly more radioactive material will be released to the 
reactor coolant system during a transient or accident than that 
experienced during steady state operation. Thus, the existing design 
bases accident analyses for River Bend Station do not bound its 
current operation with known fuel cladding failures.

    As previously stated in the evaluation of generic issues raised by 
the April 1998 UCS report, the staff has previously considered the 
safety implications of operation with fuel leakage on a generic basis. 
In GSI B-22, the staff considered the ability to accurately predict 
fuel performance under normal and accident conditions. In its 
evaluation of the issue, the staff concluded that releases during 
normal operation would be increased because of fuel defects, but would 
not be increased beyond regulatory limits. The staff also concluded 
that the release from fuel damage during design-basis accidents and 
severe accidents would be much larger than the release attributed to 
preexisting fuel defects, and the magnitude of the release would not be

[[Page 20338]]

significantly affected by preexisting fuel defects. Therefore, the 
consequence from leaking fuel was determined to be very small.
    The Petitioner has, however, noted some apparent inconsistencies in 
documentation of the licensing basis as found in the USAR for River 
Bend that could be taken out of context. The statements cited for two 
events--the control rod withdrawal error from power and recirculation 
flow control error--are not consistent with the other parts of the 
River Bend licensing basis discussed in this evaluation. The technical 
basis for coolant activity limits clearly permits operation with a 
limited amount of fuel leakage and, as discussed, the design basis does 
not preclude the possibility of limited fuel leakage during operation. 
Therefore, although these events should not cause fuel damage, 
preexisting leakage could still be a consideration, and only the 
activity in the reactor system coolant up to the technical 
specification limit would be available for release. The MSLB is 
considered the limiting event with respect to release of coolant 
activity from leaking fuel. The staff expects that the consequences of 
the MSLB would bound those that would be predicted for the control rod 
withdrawal error from power or the recirculation flow control error 
events. Thus, the minor discrepancies uncovered by the Petitioner in 
the documentation of the plant licensing basis do not constitute a 
safety concern requiring NRC action.
    The licensee has taken actions to limit the effects of the minor 
fuel rod defects at River Bend reported on September 21, 1998. The 
control rod pattern has been altered to achieve a depressed flux 
profile in the vicinity of the leaking rods, thereby suppressing the 
production of fission products as the plant continues operation at 
slightly less than full power. Following the initial detection of a 
leaking rod, the licensee reduced the activity in the pretreatment 
offgas sample from 22.5 mCi/sec to 1.8 mCi/sec, which was very close to 
the prefuel-leak level of 1 mCi/sec. The peak value was never more than 
a small fraction of the technical specification limit of 290 mCi/sec. 
The offgas treatment system has been effectively eliminating any 
detectable radioactivity in offgas effluent, and only small dose rate 
increases were observed in areas of the plant in which offgas system 
components are located. Since work is not normally performed in those 
areas, the licensee did not institute any additional exposure controls. 
However, the licensee is continuing to closely monitor the offgas 
system to ensure that the coolant activity concentration remains within 
technical specifications limits.
d. ALARA Concerns
    The Petitioner further stated that Entergy Operations, Inc., was 
violating its licensing basis with regard to the ALARA worker 
protection program (as stated):
    In addition to operating with non-bounding design bases accident 
analyses, it appears that the River Bend licensee is also violating 
its licensing basis for worker radiation protection. UFSAR Section 
12.1.1, ``Policy Consideration,'' stated:

    The purpose of the ALARA [as low as reasonably achievable] 
program is to maintain the radiation exposure of plant personnel as 
far below the regulatory limits as is reasonably achievable.

    UFSAR Section 12.1.2.1, ``General Design Considerations for 
ALARA Exposures,'' stated that River Bend's efforts to maintain in-
plant radiation exposure as low as is reasonably achievable 
included:

    Minimizing radiation levels in routinely occupied plant areas 
and in vicinity of plant equipment expected to require the attention 
of plant personnel.

    According to the NRC Information Notice No. 87-39, ``Control of 
Hot Particle Contamination at Nuclear Plants:''

    A plant operating with 0.125 percent pin-hole fuel cladding 
defects showed a five-fold increase in whole-body radiation exposure 
rates in some areas of the plant when compared to a sister plant 
with high-integrity fuel (<0.01 percent leakers). Around certain 
plant systems the degraded fuel may elevate radiation exposure even 
more.

    Industry experience demonstrated that reactor operation with 
failed fuel cladding increased radiation exposures for plant 
workers. The River Bend licensee has a licensing basis requirement 
to maintain radiation exposures for plant workers as low as is 
reasonably achievable. The River Bend licensee informed the NRC 
about potential fuel cladding failures. It could shut down the 
facility and remove the failed fuel assemblies from the reactor 
core. Instead, it continues to operate the facility with higher 
radiation levels.

    In its letter to the NRC dated February 11, 1999, the River Bend 
licensee stated that if the plant were to shut down solely to remove 
leaking fuel bundles, worker exposure would be increased since 
additional exposure would later be incurred for normal shutdown and 
maintenance activities. Also, during the February 22, 1999, informal 
public hearing on the Petition, the River Bend licensee stated that 
dose rates in the general plant areas are essentially unchanged and 
that the average daily dose to plant workers has remained at the 
historical level of approximately 0.14 person-rem per day during normal 
operations. River Bend has seen some increased levels in dose rates in 
isolated areas, such as in rooms containing offgas system equipment; 
however, these areas are not routinely occupied and access to the rooms 
are controlled by the health physics department. The licensee stated 
that if a 14-day outage were conducted to remove defective fuel 
bundles, the outage would incur a worker dose on the order of 9 person-
rem for reactor disassembly, reassembly, and refueling activities. This 
exposure would be in addition to that incurred from activities planned 
for the scheduled refueling outage. The licensee contends that shutting 
down in this situation to replace leaking fuel would be an action 
contrary to ALARA. The staff agrees that conducting plant shutdown only 
to address the current situation at River Bend would be contrary to the 
ALARA principle for plant workers, provided exposure levels remain at 
their current values.
    River Bend has two independent radiation-detection systems capable 
of sensing fission-product release from leaking fuel rods--main steam 
line radiation monitors and offgas system radiation monitors. The main 
steam line radiation monitors are used to detect high radiation levels 
from gross fuel failure. The offgas system radiation monitors can 
detect low-level emissions of noble gases, which are indicative of 
minor fuel damage. The offgas system monitor indication signaled the 
recent fuel damage found at River Bend.
    The actions taken by the licensee to limit further fuel damage, as 
well as the continued attention to reactor coolant activity and offgas 
radiation levels, provide confidence that River Bend can continue safe 
operation, within its licensing basis, with the limited fuel leakage 
recently detected.

C. Plant-Specific Concerns--Perry Nuclear Power Plant

    On the basis of the reported fuel leakage at Perry, the Petitioner 
states that the generic concerns contained in the UCS report apply to 
the Perry plant. In the opinion of the Petitioner, plant licensing 
bases do not permit operation of the plant with known fuel leakage.
    As discussed in the consideration of generic safety concerns, the 
staff does not agree that pre-existing fuel cladding defects and 
resultant fuel leakage violate plant licensing bases. The staff also 
considers that conclusion valid for Perry. Fuel leakage during plant 
operation is not precluded by licensing basis provisions requiring that 
fuel integrity be maintained as designed. The Perry design basis itself 
allows the possibility of leakage while ensuring that cladding damage 
does not result because of specific operationally related causes. Fuel 
is also designed to

[[Page 20339]]

maintain its structural integrity to ensure core coolability and to 
ensure that control rods can be inserted.
    The Updated Safety Analysis report (USAR) for Perry contains 
unacceptable consequences criteria for different event categories (USAR 
Tables 15A.2-1 through 15A.2-4). The unacceptable consequences for 
normal operation do not preclude fuel leakage. The second criterion 
listed precludes fuel failure to the extent that the limits of 10 CFR 
Part 20 would be exceeded. The unacceptable consequences for 
anticipated operational transients prohibit fuel failure predicted as a 
direct result of transient analysis. For abnormal transients and 
design-basis accidents, widespread fuel cladding perforations and fuel 
cladding fragmentation are prohibited.
    Two parts of the fuel system licensing basis for Perry show that 
limited fuel leakage during plant operation is a design consideration. 
The fuel system design basis for Perry is given in the USAR Section 15B 
by reference to the generic topical report ``General Electric Standard 
Application for Reactor Fuel,'' NEDE-24011-P-A. The generic topical 
report details fuel cladding operating limits to ensure that fuel 
performance is maintained within fuel rod thermal and mechanical design 
and safety analysis criteria. The limits are given for normal operating 
conditions and AOOs in terms of specific mechanical and thermal 
specifications. Evaluations of specific fuel failure mechanisms under 
normal operation and AOOs were discussed, such as stress and strain, 
hydraulic loads, fretting, and internal gas pressure, to ensure that 
fuel failure did not result from these causes. The design bases did not 
preclude the possibility that fuel failure could occur for other 
reasons, such as pre-existing cladding flaws leading to leakage.
    The Technical Specifications for Perry (Section 3.4.8) contain a 
limit for RCS specific activity. The basis for this limit is the same 
as that discussed in the consideration of the generic safety concerns. 
Section B3.4.8 of the Perry Technical Specification ``Bases'' 
acknowledges that ``the reactor coolant acquires radioactive materials 
due to release of fission products from fuel leaks.'' Thus, fission 
products released during plant operation are clearly considered to be 
contributors to the source term used for safety analysis of the main 
steamline break release consequences. The technical specifications 
state that the limit is set to ensure that any release as a consequence 
of a main steamline break is less than a small fraction of the 10 CFR 
Part 100 guidelines. These portions of the Perry licensing basis are 
consistent with NRC regulations regarding fuel performance and the 
associated NRC guidance used by licensees to implement those NRC 
regulations that were covered earlier in the discussion regarding 
generic concerns.
    The licensee has taken actions to limit the effects of the existing 
minor fuel leaks at Perry. The control rod pattern has been altered to 
achieve a depressed flux profile in the vicinity of the leaking rods, 
thereby suppressing the production of fission products as the plant 
continues operation. The off-gas treatment system has been effectively 
eliminating radioactivity in off-gas effluent, and there has been no 
change in general radiation area dose rates. However, the licensee is 
continuing to closely monitor the off-gas system pre-treatment 
radiation levels and is ensuring that the coolant activity 
concentration remains within technical specifications limits.
    Perry has two independent radiation detection systems capable of 
sensing fission product release from leaking fuel rods: main steamline 
radiation monitors and off-gas system radiation monitors. The main 
steamline radiation monitors are used to detect high radiation levels 
from gross fuel failure. The off-gas system radiation monitors can 
detect low-level emissions of noble gases, which are indicative of 
minor fuel damage.
    In its letter to the NRC dated February 11, 1999, the Perry 
licensee stated that if the plant were to shut down solely to remove 
fuel bundles exhibiting leakage, plant worker exposure would be 
increased since additional exposure would later be incurred for normal 
shutdown and maintenance activities. The licensee contends that 
shutting down in this situation to replace leaking fuel would be an 
action contrary to ALARA. The staff agrees that conducting plant 
shutdown only to address the current situation at Perry would be 
contrary to the ALARA principle for plant workers, provided exposure 
levels remain at their current values.
    The actions taken by the licensee to limit further fuel damage, as 
well as the continued attention to reactor coolant activity and off-gas 
radiation levels, provide confidence that Perry can continue safe 
operation, within its licensing basis, with the limited fuel leakage 
detected.

IV. Conclusion

    The Petitioner's requests are denied for the reasons specified in 
the preceding sections that discuss the Petitioner's information 
supporting the request. The Petitioner did not submit any significant 
new information about safety issues. Neither the information presented 
in the Petition nor any other subsequent information of which the NRC 
is aware warrants the actions requested by the Petitioner.
    A copy of this Director's Decision will be filed with the Secretary 
of the Commission for review in accordance with 10 CFR 2.206(c). This 
Decision will become the final action of the Commission 25 days after 
its issuance unless the Commission, on its own motion, institutes a 
review of the Decision within that time.

    Dated at Rockville, Maryland, this 18th day of April 1999.

    For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 99-10357 Filed 4-23-99; 8:45 am]
BILLING CODE 7590-01-P