BREDL Archive Southern Anti-Plutonium Campaign

Preparation of an Environmental Impact Statement for construction, operation and deactivation of a proposed Mixed Oxide Fuel Fabrication Facility to be constructed at the Department of Energy’s Savannah River Site in South Carolina

BLUE RIDGE ENVIRONMENTAL DEFENSE LEAGUE
www.BREDL.org ~ PO Box 88 Glendale Springs, North Carolina 28629 ~ Phone (336) 982-2691 ~ Fax (336) 982-2954 ~ BREDL@skybest.com

May 8, 2001

Mike Lesar, Chief, Rules and Directives Branch

Division of Administrative Services

Office of Administration

Mail Stop T6D59

U.S. Nuclear Regulatory Commission,

Washington DC 20555.

Re: Docket 70-3098, Preparation of an Environmental Impact Statement for construction, operation and deactivation of a proposed Mixed Oxide Fuel Fabrication Facility to be constructed at the Department of Energy’s Savannah River Site in South Carolina

Dear Sir:

I write to provide additional comments on the proposed MFFF. This letter will focus on nuclear reactor safety issues with regards to fuel made from weapons-grade plutonium.

The planned use of mixed oxide (MOX) plutonium fuel is unsafe, uneconomical, and unnecessary. We oppose the use of plutonium fuel in commercial power reactors for the following reasons:

Plutonium fuel derived from dismantled weapons is an experimental program which cannot be compared to European experience with plutonium fuel made from nuclear waste. The mix of isotopes includes 64% higher concentration of Plutonium 239, the heart of a nuclear weapon.
Safety hazards in nuclear plants are a combination of human and technical errors. Both types of error are noted in the Nuclear Regulatory Commission’s Plant Performance Reviews of the McGuire and Catawba reactors (see Attachment). Because of the inherent hazards in these plants, DOE should not move forward with the MOX plan.

The Catawba and McGuire plants operated by Duke have a radiation containment building which depends on blocks of ice to reduce heat and pressure in case of a reactor accident. Duke’s ice condenser system has inherent weaknesses which have resulted in safety problems and lengthy closures of other utility reactors using the same system.

The Department of Energy’s selection of DCS and the planned utilization of Duke Power reactors have not been opened to full public scrutiny. The experimental nature of a weapons-derived fuel project requires a thorough and independent assessment by NRC. Additional information from DOE and DCS is required to fully assess the safety of this program.

The planned use of plutonium fuel in the reactors operated by Duke Power would be a dangerous precedent in the domestic nuclear industry, needlessly exposing many people to the risk of additional radiation exposure from a plutonium fuel-powered plant accident. Furthermore, the use of plutonium fuel in commercial reactors is a break with two decades of American non-proliferation policy and opens a door for other nations to exploit for the purpose of plutonium weapons production.

Plutonium Fueled Reactor Hazards

Commercial Nuclear Reactors Were Not Designed for Plutonium Fuel

Atom splitting in a reactor releases neutrons which split other atoms. This chain reaction is what drives the reactor. The chain reaction must be precisely controlled in order to produce power safely. Compared to neutrons from uranium atoms, plutonium releases more neutrons at a higher speed and energy during the fission process.

“Technical issues that arise in the analysis of risk at plants using MOX focus on the vulnerability of fuel to neutronically induced core disruption and the different inventory of radionuclides available for release from the fuel during accidents. The differences in neutronics and coupling between neutronics and thermal hydraulics result in different responses of MOX and conventional fuel to reactivity transients.”
May 17, 1999 letter to NRC Chairman from the Chair of the Advisory Committee on Reactor Safeguards

Adding plutonium to the reactor in the form of MOX reduces the ability to control the chain reaction:

The rate of fission in plutonium increases with temperature, and the problem is greater with MOX fuel made from weapons-grade plutonium. MOX fuel in a reactor attains higher temperatures than uranium fuel because of the higher quantity of transuranic elements produced during irradiation.
The percentage of delayed neutrons emitted seconds to minutes after a plutonium atom splits is just one-third that of uranium (Pu239=0.2%, U235=0.65%). This means plutonium releases a higher amount of its neutrons in a single burst and adds to reactor control problems.

Plutonium captures more neutrons than uranium, increasing fission and making control measures less effective.

~Institute for Energy and Environmental Research, SDA February 1997

Experience With Plutonium Fuel In The U.S. Is Limited

The MOX program is experimental in that no reactor has ever been operated with fuel derived from weapons-grade plutonium. European experience with MOX includes fuel derived from irradiated nuclear fuel, a waste product. Duke Power propose to use something quite different: fuel made from dismantled plutonium weapons. The fuel made from dismantled plutonium weapons would be comprised of a different mix of radioactive isotopes. For example, the plutonium in MOX fuel planned for Catawba and McGuire would be 92% Pu-239, where European reactor MOX contains just 56% of Pu-239. Duke‘s reactors would be fueled with 64% more Plutonium-239; the most dangerous isotope which provides the explosive power of a nuclear weapon.

Reports on Duke Power’s McGuire and Catawba reactors describe human and technical errors which raise questions as to safety and reliability. Without modifications of the plants’ containment vessels, inspection schedules, and maintenance procedures, the increased danger of reactor embrittlement may be hidden by outwardly normal appearance. Safety margins would be reduced if commercial power reactors designed for uranium fuel use plutonium fuel. In her May 17, 1999 letter to Nuclear Regulatory Commission Chairman, the Chair of the Advisory Committee on Reactor Safeguards said,

“The U.S. Department of Energy is proposing to dispose of some fraction of the Nation's excess weapons-grade plutonium by converting this plutonium into MOX for use in commercial nuclear power plants. There is, however, rather limited operational or regulatory experience with the use of MOX in the U.S. Even the experience in other countries is not extensive.”

Reactor Embrittlement

Higher energy neutrons from plutonium are more likely to strike reactor parts such as the stainless steel containment vessel. This neutron bombardment degrades the metal parts of the reactor and the metal becomes brittle. An embrittled reactor may look unchanged, but it will not perform as well under extreme conditions. For example, an event causes the water level in the reactor to drop. Normally, the heated water is replaced by cold water from outside the reactor. However, this cold water bath may cause the embrittled metal part to fail and a minor reactor failure becomes a major one. Embrittlement of reactor parts is a well-known phenomenon and has caused premature closing of commercial power reactors. The additional neutron bombardment caused by MOX fuel’s plutonium will increase the tendency of parts to wear out and fail.

Plutonium Fuel is Unstable

French test results suggest that plutonium fuel is more unstable than uranium fuel. In 1997 a MOX fuel rod violently ruptured when subjected to test conditions designed to simulate an accident. The uranium fuel rod in that test did not rupture. The May 17, 1999 Advisory Committee on Reactor Safeguards letter to Nuclear Regulatory Commission Chairman states,

“We are aware of experimental studies that show there to be enhanced release of fission gases to the fuel-cladding gap during reactor operations with MOX relative to conventional fuels. This may simply be an effect caused by fuel temperature. We are also aware of anecdotal accounts of the results of VERCOURS tests in France dealing with the release of volatile radionuclides such as cesium from MOX under severe accident conditions. Results of these tests revealed that during the early stages of core degradation, releases of volatile radionuclides from MOX are more extensive than from conventional fuels at similar levels of burnup.”

Safety and Reliability Problems at Catawba and McGuire

Hazards in nuclear plants are a combination of human and technical errors. Both types of error are noted in the Nuclear Regulatory Commission’s Plant Performance Reviews of the McGuire, and Catawba reactors.

The Department of Energy’s Environmental Synopsis contains a Nuclear Regulatory Commission Systematic Assessment of Licensee Performance (SALP) for the Catawba, McGuire, and North Anna nuclear power stations. However, the Nuclear Regulatory Commission suspended the SALP program in favor of Plant Performance Reviews (PPR’s). PPR’s were completed in March 1999 for these reactors and rate all three merely “acceptable.” The PPR’s note shortcomings in ice condenser maintenance and inspection in McGuire and Catawba reactors and corrosion of service water pipes and auxiliary feedwater pipes (the only source of water for steam generators when the main feedwater system fails), and examples of poor engineering performance at North Anna and Catawba. I include excerpts from the Catawba PPR:

Catawba NRC Plant Performance Review March 25, 1999:

“Unit 1 experienced a forced outage of approximately three weeks in duration due to blocked flow channels in portions of the ice condenser.”
“Problems in maintenance programs and processes included examples of surveillance deficiencies for ventilation systems and ice condensers.”

“The engineering performance decline was the result of deficiencies in auxiliary building ventilation system testing, an overheating event of the upper surge tank, and degraded conditions in the Unit 1 ice condenser. While the issues were ultimately resolved properly, each had roots in poor engineering performance.”

Catawba and McGuire utilize ice condensers which absorb energy and allow smaller physical containment structures to contain accidental radioactive releases from the reactors. Ice condensers must work during a reactor emergency-as an air bag must work during an auto accident. The Donald C. Cook nuclear plant uses similar technology was shut down because of ice condenser problems. No nuclear plant should use MOX until these ice condenser problems are solved.

“The NRC has a mandate to protect public health and safety. The findings from D C Cook indicate that both of its units may not have protected the public had there been an accident. The NRC does not know about the adequacy of the other ice condensers. The people living around these plants should be protected by solid designs and functioning safety equipment, not by sheer luck.”
David Lochbaum, Union of Concerned Scientists Backgrounder on Ice Condensers, 5/26/98

Public Health Impacts From Radiation Releases

MOX fuel has a greater quantities of plutonium and other hazardous radioactive isotopes such as Americium 241 and Curium 242--actinide elements which would cause additional harmful radiation exposure to the public during a failure of the reactor containment structure.

“Public attention has been drawn to the higher actinide inventories available for release from MOX than from conventional fuels. Significant releases of actinides during reactor accidents would dominate the accident consequences. Models of actinide release now available to the NRC staff indicate very small releases of actinides from conventional fuels under severe accident conditions.” (emphasis added)
~Letter from Advisory Committee on Reactor Safeguards to Nuclear Regulatory Commission Chairman, May 17, 1999

The release of these more toxic radioactive elements would cause more fatalities immediately following the accident, and would cause more cancers in the years following the breach. A recent study by the Nuclear Control Institute estimates that the risk to the public near McGuire or Catawba of contracting a deadly cancer following a severe accident will increase by nearly 40% when the plants start using plutonium fuel.

A study by Dr. Edwin Lyman estimated the number of cancer deaths that could result from an accident at a plant using MOX fuel:

A reactor using weapons-grade MOX fuel in one-third of its core contains, on average, about three times more plutonium 239, five times more americium 241, and four times more curium 242 than a reactor using only LEU (low enriched uranium) fuel.
Compared to an LEU-fueled reactor, a severe accident at a reactor with a one-third weapons grade MOX core, involving a core meltdown and containment failure or bypass, could cause approximately 30% more cancer fatalities, corresponding to hundreds or even thousands of additional cancer deaths, depending on the type of accident.

The annual risk of contracting a fatal cancer as a result of a severe accident would increase by nearly 40 percent for an average individual living near a nuclear plant if the plant were to load weapons-grade MOX in one-third of its core.

~Nuclear Control Institute MOX Safety Report, March 1999

“Dr. Lyman’s study indicates that the increase in risk associated with the use of weapons-grade MOX in typical U.S. power reactors is so large that, according to NRC staff regulatory guidance, an application for a license amendment to use MOX ‘would not normally be considered.’ See Office of Nuclear Energy Research, Regulatory Guide 1.174.”

~Nuclear Control Institute Background Paper January 21, 1999

Plutonium Fuel Transportation Hazards

Emergency response to rail or highway accidents must be well-prepared and rapid. Delays in response to accidents which involve the release of radioactive material would expose unknown numbers of people to negative health effects. In 1996, a DOE Transport and Safeguards Division Safe Secure Transport (SST) trailer carrying nuclear weapons slid off the road and rolled over in rural Nebraska. Four hours elapsed before DOE headquarters were notified, and it was 20 hours before a Radiological Assistance Program team determined there was no release. A similar delay in response to a MOX fuel accident could make effective emergency response dangerous and clean-up impossible. The following comment by the Georgia Environmental Protection Division cites vehicular tests of powdered materials deposited on roadways and takes issue with the DOE’s approach to emergency response to accidental plutonium fuel releases.

“After passage of about 100 cars only a small fraction of the original contamination remained on the road surface. Unless emergency officials promptly close the accident scene to vehicle traffic (an unlikely situation), emergency responders may face an incident scene that is, unknown to them, extremely hazardous due to respirable plutonium. Post emergency actions may also be complicated due to the enhanced spread of contamination by vehicle traffic.”
~Georgia Environmental Protection Division comments on DOE SPD DEIS

Many rural communities in South Carolina, North Carolina, and Virginia resemble Nebraska in that fire departments and emergency first-responders are entirely volunteer. This does not imply a lack of dedication, but limited resources do not allow volunteers to be prepared for every possible emergency. I served as a volunteer fireman in NC for many years and our experience, training, and equipment did not prepare us for radionuclide transport accidents.

Complete Information Has Not Been Made Public

Duke, Cogema, Stone & Webster (DCS) and its subcontractors must be subject to full public scrutiny. The DOE’s Environmental Synopsis is at least two steps removed from the original data which the DOE required prospective contractors to submit in a Request For Proposal (#DE-RP02-98CH10888). Such third-hand information does not provide a sufficient level of detail required for a thorough independent analysis. I hereby repeat our request first made in June 1999 that DOE make all information on the MOX project submitted by DCS (Duke Engineering & Services, COGEMA Inc., and Stone & Webster) available for review to members of the affected public. These data include:

DOE’s Environmental Critique
DCS environmental data and analyses for design, licensing, construction, operation, and eventual decontamination and decommissioning of a MOX facility,

DCS environmental data and analyses for irradiation of MOX fuel in existing domestic, commercial reactors,

DOE projections of populations surrounding the proposed reactor sites and evaluations of air dispersal patterns,

Oak Ridge National Laboratory data on the expected radionuclide activities in MOX fuel compared to that in low enriched uranium fuel used in reactor accident analyses, and

DCS data used in computer models for determining radiation doses from normal operations and accident scenarios.

The NRC must address all the problems outlined above in its pending environmental impact statement. Please find attached 23 pages of documents comprised of largely reports on Duke’s Catawba and McGuire operations from 1999 to 2001.

Respectfully submitted,

Louis Zeller

Attachments

NRC Plant Performance Reviews

Shortcomings, problems, errors, and poor engineering performance

McGuire NRC Plant Performance Review, March 25, 1999

These Duke Power plants in North Carolina began operation in 1981 and 1983. The following excerpts are from the NRC’s PPR:

“...shortcomings in oversight of diesel generator vendors were noted.”

“Several human performance errors during routine plant evolutions were identified...”

“Minor program and procedure problems still indicate room for improvement. In addition to core inspections, a regional initiative inspection is planned for ice condenser inspections during the Unit 2 refueling...”

“An area for improvement was engineering programs and processes such as ... procedures and work instructions for maintenance and calibration of instrumentation....”

“... some fire protection system maintenance material conditions weaknesses have been noted...”

“Self-identified problems with fire barrier penetration seals were reported to the NRC and improvements are being made.”

Catawba NRC Plant Performance Review, March 25, 1999

These Duke Power reactors began operation in 1985 and 1986. The following exerpts are from the NRC’s PPR:

“Unit 1 experienced a forced outage of approximately three weeks in duration due to blocked flow channels in portions of the ice condenser.”

“Engineering performance continued to be acceptable but declined since the last assessment as a result of emergent issues rooted in shortcomings in engineering’s performance.”

“Examples of poorly supported or non-conservative operability or root cause determinations were noted.”

“Problems in maintenance programs and processes included examples of surveillance deficiencies for ventilation systems and ice condensers.”

“The engineering performance decline was the result of deficiencies in auxiliary building ventilation system testing, an overheating event of the upper surge tank, and degraded conditions in the Unit 1 ice condenser. While the issues were ultimately resolved properly, each had roots in poor engineering performance.”

Nuclear Regulatory Commission

Office of Public Affairs -- Region II

61 Forsyth Street, Suite 23T85, Atlanta, GA 30303

Ken Clark (Phone: 404/562-4416, E-mail: kmc2@nrc.gov)

Roger Hannah (Phone 404/562-4417, E-mail: rdh1@nrc.gov)

No: II-98-35

May 11, 1998

NRC OFFICIALS SEND INSPECTION TEAM TO CATAWBA

Augmented Inspection Team Will Inspect and Assess Recent Event

Nuclear Regulatory Commission officials today dispatched an Augmented Inspection Team to the two-unit Catawba nuclear power plant, operated by Duke Energy Company near Rock Hill, South Carolina. The team will assess the circumstances of an event on May 7 which left the Catawba Unit 1 auxiliary feedwater system in a condition different from its design. NRC officials said no accident occurred. Duke engineers told the agency the plant suffered no adverse effects. NRC's interest is in learning how a failure in the unit's non safety-related, secondary condensate system resulted in operators declaring inoperable all trains of the safety-related auxiliary feedwater system. Catawba has a primary and secondary water system. The primary system cools the reactor by circulating water directly through the core. It then passes through thousands of tubes into a large cylindrical tank known as a steam generator. The steam generator is filled with water supplied by the secondary system. This secondary system water serves two functions: (1) it absorbs heat from the primary reactor cooling system, and (2) it produces steam which turns turbines to generate electricity. After turning the turbines, this steam is condensed back into water and normally recirculates through the feedwater system to produce more steam. The auxiliary feedwater system serves as a backup to the feedwater system. On May 7, plant operators determined that, following a planned power reduction, tanks which hold water for use in the auxiliary feedwater system registered a water temperature in excess of system design limits. The operators declared three auxiliary feedwater pumps inoperable due to uncertainty related to their operation under higher water temperatures. Duke attributed the cause to an improper setting on a pump recirculation valve. This erroneous set point, the company believes, resulted in a higher than normal flow of water during the power reduction, diverting some of the hotter water to the auxiliary feedwater system tank. Operators returned water temperatures to normal and declared the auxiliary feedwater system operable. Permanent corrective actions are being evaluated. NRC officials said members of the inspection team will arrive at the site Monday afternoon and Tuesday morning. Team leader Kerry Landis, a branch chief in the NRC Atlanta regional office's Division of Reactor Projects, will be available to discuss preliminary team findings with the public and the press at the conclusion of the inspection.

Nuclear Regulatory Commission

Office of Public Affairs -- Region II

61 Forsyth Street, Suite 23T85, Atlanta, GA 30303

Ken Clark (Phone: 404/562-4416, E-mail: kmc2@nrc.gov)

Roger Hannah (Phone 404/562-4417, E-mail: rdh1@nrc.gov)

No: II-98-47

June 12, 1998

NRC TO MEET WITH DUKE ENERGY ON JULY 8 TO DISCUSS NUCLEAR POWER PLANT ICE CONDENSERS

Status of Systems at McGuire and Catawba to Be Discussed

Officials from the Nuclear Regulatory Commission and Duke Energy Corporation will meet in Atlanta on July 8 to discuss the status of the ice condenser safety system at the McGuire nuclear power plant in North Carolina and the Catawba nuclear power plant in South Carolina. The meeting will be held at 10:00 a.m. (EDT) in NRC offices on the 24th floor of the Atlanta Federal Center, located at 61 Forsyth Street, S.W. The meeting is open to observation by the public and media, and NRC officials will be available at its conclusion to answer questions from observers who attend. Ice condensers are incorporated into some Westinghouse pressurized water reactor containment building designs. They are constructed so that steam released during an accident wil be directed through borated ice where it is cooled and condensed. This serves to mitigate buildup of pressure on the containment building walls.

Nuclear Regulatory Commission

Office of Public Affairs -- Region II

61 Forsyth Street, Suite 23T85, Atlanta, GA 30303

Ken Clark (Phone: 404/562-4416, E-mail: kmc2@nrc.gov)

Roger Hannah (Phone 404/562-4417, E-mail: rdh1@nrc.gov)

No: II-99-43

July 12, 1999

NRC STAFF SETS ENFORCEMENT CONFERENCE WITH DUKE ENERGY

TO DISCUSS APPARENT VIOLATIONS AT CATAWBA NUCLEAR STATION

The Nuclear Regulatory Commission staff has scheduled a predecisional enforcement conference in Atlanta on Tuesday, July 20, to discuss with Duke Energy Corporation apparent violations of NRC requirements related to the Unit 1 and Unit 2 ice condensers at the Catawba Nuclear Station near York, South Carolina. The meeting will be held at 1:00 p.m. in Bridge Conference Room D of the Sam Nunn Atlanta Federal Center at 61 Forsyth Street. It is open to observation by interested members of the public and news media representatives. NRC officials will be available at its conclusion to answer questions from interested observers.

NRC officials said the apparent violations include the potential inoperability of the Unit 2 ice condenser doors due to ice buildup, the failure to promptly identify and correct ice condenser blockage and damaged ice containers in both units, the failure to perform adequate inspections for foreign debris in the ice condensers, and the failure to properly install ice condenser components as designed.

Ice condensers are incorporated into some Westinghouse pressurized water reactor containment building designs. They are constructed so that steam released during an accident will be directed through borated ice where it is cooled and condensed. This serves to mitigate buildup of pressure on the containment building walls.

The decision to hold a predecisional enforcement conference does not mean that a determination has been made that violations have occurred or that enforcement action will be taken. The purpose is to discuss the apparent violations, their causes and safety significance; to provide the licensee with an opportunity to point out errors that may have been made in NRC inspection reports; and to enable the licensee to outline its proposed corrective actions.

No decision on the apparent violations or any contemplated enforcement action, such as a civil penalty, will be made at the conference. Those decisions will be made by NRC officials at a later time.

Nuclear Regulatory Commission

Office of Public Affairs -- Region II

61 Forsyth Street, Suite 23T85, Atlanta, GA 30303

Ken Clark (Phone: 404/562-4416, E-mail: kmc2@nrc.gov)

Roger Hannah (Phone: 404/562-4417, E-mail: rdh1@nrc.gov)

No: II-97-70

September 23, 1997

>NRC STAFF TO HOLD CONFERENCE WITH DUKE POWER COMPANY TO

DISCUSS APPARENT VIOLATIONS AT McGUIRE NUCLEAR PLANT

The Nuclear Regulatory Commission staff will hold a predecisional enforcement conference with Duke Power Company on Wednesday, October 1, to discuss apparent violations of NRC regulations involving ice condenser doors at the McGuire nuclear power plant near Huntersville, North Carolina.

The apparent violations involve the company's failure to ensure that ice condenser inlet doors on Unit 2 would be able to open if needed and a failure to perform adequate corrective actions based on industry experience and operational events at McGuire.

The ice condenser is a passive accident mitigation system containing about two and one-half million pounds of borated ice. If an accident were to occur, the ice condenser system would condense steam and lower pressure in the plant's containment structure. The ice is located behind a number of doors designed to open when the pressure in containment reaches a certain level above the pressure inside in the ice condenser area.

In July, McGuire plant employees determined that 10 of the 48 ice condenser inlet doors in lower containment were incapable of opening with less force than specified in the plant's technical specifications and may not have opened in an accident situation.

The meeting will be held in the NRC Atlanta office, located at 61 Forsyth Street, Room 24T20, at 10:00 a. m. It will be open to observation by the public.

NRC officials said the decision to hold a predecisional enforcement conference does not mean that a determination has been made that a violation has occurred or that enforcement action will be taken.

The purpose is to discuss apparent violations, their causes, and safety significance; to provide the licensee with an opportunity to point out any errors that may have been made in NRC inspection reports; and to enable the company to outline its proposed corrective actions.

No decision on the apparent violations or any contemplated enforcement action, such as a civil penalty, will be made at the conference. Those decisions will be made by NRC officials at a later time.

Nuclear Regulatory Commission

Office of Public Affairs

Washington DC 20555

Telephone: 301/415-8200 -- E-mail: opa@nrc.gov

No. 99-219

October 15, 1999

NRC TO ALLOW DUKE ENERGY TO SUBMIT

EARLY LICENSE RENEWAL APPLICATIONS

The Nuclear Regulatory Commission has granted a request from the Duke Energy Corporation toallow the company to submit applications to renew the licenses for its McGuire Unit 2 and the two Catawba nuclear power plants earlier than usually permitted.

NRC regulations specify that license renewal applications may not be submitted to the Commission earlier than 20 years before the expiration of the current 40-year operating license. This limit is designed to ensure that sufficient operating experience is accumulated to identify any plant-specific aging concerns. However, in amending this license renewal rule in 1995, the Commission indicated it would consider an exemption to this requirement if sufficient information was available on a plant-specific basis to justify it.

By June 2001, the earliest date the NRC said it will accept a license renewal application from Duke, McGuire Unit 1 will have achieved the required 20 years of operation; Unit 2 will have 18.3 years; Catawba Unit 1 will have 16.5 years and 15.3 years for Unit 2. In a safety evaluation, the NRC determined that the operating experience of McGuire Unit 1, in conjunction with the substantial number of years for the other three units, should be sufficient to identify any aging concerns applicable to all four units.

McGuire, 17 miles south of Charlotte, N.C., and Catawba, six miles northwest of Rock Hill, S.C., are two-unit stations utilizing pressurized water reactors with ice-condenser containments having a rated power output of about 1130 megawatts each. The four plants are sufficiently similar in design, operation and maintenance that the operating experience of McGuire Unit 1 should apply to the other three units, according to the NRC staff.

In its request for early license renewal, Duke Energy explained that regular and systematic exchanges of information among its nuclear stations provide a means to continually improve plant programs. Duke Energy plans to submit license renewal requests for all four units simultaneously, to expedite processing and reduce costs.

The current operating license for McGuire Unit 1 expires in 2021, and for McGuire Unit 2, in 2023. The current operating license for Catawba Unit 1 expires in 2024, and for Unit 2, in 2026.

Once submitted, the license renewal applications will have to meet the same requirements NRC is using in evaluating other license renewal applications. If granted, the renewed licenses will have the effect of extending the operating life of each plant by as many as 20 years.

The exemption was published in the Federal Register October 8.

Nuclear Regulatory Commission

Office of Public Affairs -- Region II

61 Forsyth Street, Suite 23T85, Atlanta, GA 30303

Ken Clark (Phone: 404/562-4416, E-mail: kmc2@nrc.gov)

Roger Hannah (Phone 404/562-4417, E-mail: rdh1@nrc.gov)

No: II-99-26

April 5, 1999

NRC FINDS PERFORMANCE 'ACCEPTABLE'

AT MCGUIRE IN LATEST REVIEW

The Nuclear Regulatory Commission staff has found that safety performance remains acceptable in the NRC's latest plant performance review at the McGuire nuclear power plant, operated by Duke Power Company near Huntersville, North Carolina. Charlotte In a letter to Duke Power which outlined the results of the review, which ran from March 1997 through January 1999, Charles R. Ogle, an official in the agency's Atlanta regional office, said "overall performance at McGuire was acceptable" and that "strong management involvement resulted in improvements" in the area of plant operations. He said performance also improved in maintenance and plant support and that engineering performance was "consistent."

Ogle said that the NRC plans to conduct inspections from the Atlanta regional office on the plant's ice condensers, fire protection system, and development of an Independent Spent Fuel (Dry Cask) Storage Installation, in addition to its normal inspection program, during the next assessment period at McGuire

The text of the plant performance review letter is available from the NRC Region II Office of Public Affairs and on the NRC web site at: http://www.nrc.gov/OPA/ppr.

NRC reviews safety performance twice a year at every licensed nuclear power plant in the nation. These reviews give the agency staff an integrated assessment of plant performance and provide a basis for planning inspection activities.

Plant performance reviews are being used by the NRC as an interim measure to monitor nuclear power plant safety. The agency began using it for this purpose after suspending the Systematic Assessment of License Performance (SALP) process until a new assessment program is developed. Previously, SALP reports were issued every 12 to 24 months.

The new reactor oversight and assessment program being developed will provide quarterly performance reports, based on a number of performance indicators and on inspection findings. This program will be tested at eight sites beginning in June and will be extended to all plants next January.

NRC Performance Summaries

Catawba 1

Initiating Events

Significance: TBD Feb 16, 2001
Identified By: NRC
Item Type: AV Apparent Violation
Failure to Promptly Identify and Correct the Unit 1 Residual Heat Removal System Water Hammer Condition
An apparent violation of 10 CFR 50, Appendix B, Criterion XVI was identified for the failure to identify a root cause and establish effective corrective actions to prevent repetitive water hammer events in the Unit 1 residual heat removal (ND) system which have caused the repeated failure of snubbers on supports 1-R-ND-0226 and 1-R-ND-0596. (Section 40A2.b.(2).2)
Inspection Report# : 2001003

Mitigating Systems

Significance: G Mar 30 2001
Identified By: NRC
Item Type: FIN Finding
Failed to Demonstrate Performance of the Station Drinking Water System as Backup Cooling Water to the Unit 1 and 2 A Train Charging Pumps
The licensee failed to demonstrate that the performance or condition of the station drinking water system, a risk-important system that provides backup cooling water to the Unit 1 and 2 A train charging pump motors and bearing oil coolers, was being effectively controlled through the performance of appropriate preventive maintenance (including surveillance activities). This resulted in a failure to recognize and correct a degraded system pressure condition, until it was identified by the inspectors. The degraded pressure condition was determined to be of very low safety significance because an analysis performed by the licensee demonstrated that the backup function to cool the charging pumps and motors would have been provided at the degraded pressure (Section 1R12.2).
Inspection Report# : 2000006

Significance: G Mar 30, 2001
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Adequately Perform TS SR 3.4.9.3 for Pressurizer Heaters
A non-cited violation was identified regarding the licensees failure to properly perform Technical Specification Surveillance Requirement 3.4.9.3, which verifies that pressurizer heaters can be automatically transferred from their normal power supplies to their emergency power supplies. Once identified, the portion of the automatic circuit that had been omitted from the test was properly tested on February 5, 2001, and was verified to be functional. This finding had a credible impact on safety because the licensee had never demonstrated the full automatic capability of the power supply transfer circuitry for the pressurizer heaters, which are important for maintaining pressurizer pressure control during a loss of offsite power event. The finding was also the latest in a number of missed surveillance requirements identified at Catawba over the last two to three years. This finding was of very low safety significance because the circuit was functional when tested and because of provisions in the licensee's emergency procedures for manually aligning the heaters to their emergency power source had the automatic transfer failed during a loss of normal power event (Section 1R22).
Inspection Report# : 2000006

Significance: G Feb 16, 2001
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Identify Conditions Adverse to Quality - two examples
The first example of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified for a failure to identify a condition adverse to quality which contributed to a Unit 1 reactor vessel level instrument system (RVLIS) channel being inoperable. A quality control inspector did not initiate a Problem Investigation Process report after identifying that a RVLIS system terminal board was not reconnected (wired) in accordance with electrical drawings. Because of an electrical drawing error, the terminal board was then wired incorrectly and resulted in a failure to meet Technical Specification 3.3.3. Function 4 requirements for an inoperable RVLIS channel from June 1999 to November 4, 2000. Because other indications would have been available to the operators to mitigate the consequences of an accident, and based on the probability that the operators would have used the conservative indication of decreasing reactor vessel level from the operable RVLIS channel, the inspectors determined that this issue was of very low safety significance. (Section 40A2.a.(2).2) The second example of a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI was identified for a failure to identify a condition adverse to quality which contributed to not recognizing that four post accident monitoring control room recorders in Unit 1 were inoperable from September 24 through September 29, 2000, and degraded from September 29 through October 19, 2000. Specifically, operators did not review applicable electrical drawings in order to identify which components were supplied from a failed electrical breaker. Consequently, they did not recognize that post accident monitoring control room recorders, which are used in the emergency operating procedures to determine mitigation strategies, were no longer operable. Because other indications would have been available to the operators to use in lieu of these accident monitoring recorders and because the Technical Specification Limiting Condition for Operation requirements were not exceeded, the inspectors determined that this issue was of very low safety significance. (Section 40A2.a.(2).3)
Inspection Report# : 2001003

Significance:G Feb 16, 2001
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Meet 10 CFR 50, Appendix B, Criterion III and XI for Unit 1 RIVLIS
10 CFR 50, Appendix B, Criterion III, requires in part that the design bases is correctly translated into drawings. 10 CFR 50, Appendix B, Criterion XI, requires in part that all testing required to demonstrate that components will perform satisfactorily in service is identified and performed. To the contrary, an error in the electrical drawings for the Unit 1 reactor vessel level indication system (RVLIS) circuitry was introduced during a previous drawing revision on July 1, 1985, which led to the improper wiring of the RVLIS instrumentation in a June 1999 modification. Following the modification activities, the licensee did not develop an adequate post modification testing plan for the RVLIS electrical circuitry, resulting in one channel of RVLIS being inoperable for 18 months. This finding was determined to have very low safety significnace and is captured in the licensee's corrective action program under PIP C-00-05558 (Section 4OA7).
Inspection Report# : 2001003

Significance: G Jun 24, 2000
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Scope an Accident Mitigating Function Associated with ECCS Leak Detection in the Maintenance Rule
The licensee failed to include in its maintenance rule scope an accident mitigating function for a control room alarm associated with emergency core cooling system post-accident leak detection capability. The alarm was tied to residual heat removal and containment spray pump room sump levels and was identified in 1998 as a mitigating function, as described in the Catawba Updated Final Safety Analysis Report. As a result, two functional failures were not properly classified in February 2000. This issue was characterized as a non-cited violation of 10 CFR 50.65 (b)(2) and was determined to have very low safety significance because the licensee's scoping and functional failure determination errors did not directly result in additional unavailability of the alarm function (Section 1R12.2).
Inspection Report# : 2000003

Significance: G Jun 24, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Provide Adequate Procedures for Performing Maintenance on Safety-Related Sump Pump Level Switches
Residual heat removal and containment spray pump room sump level alarm function was lost for several months up to February 2000 due to inadequate maintenance procedures associated with sump level switch calibrations. This issue was characterized as a non-cited violation of Technical Specification 5.4.1 and was determined to be of very low safety significance due to the availability of other emergency core cooling system leak detection methods (Section 4OA3.2).
Inspection Report# : 2000003

Occupational Radiation Safety

Significance: G Sep 23, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Implement Radiation Control Procedures for Posting Extra High Radiation Areas as Required by TS 5.4.1.a
A single event, resulting in two non-cited violations, involved: (1) a failure to implement radiation control procedures for posting an extra high radiation area as required by TS 5.4.1.a.; and (2) failure to lock or control entrance to an extra high radiation area as required by Technical Specification 5.7.2 and Title 10 CFR Part 20.1601. This event was determined to be of very low safety significance because minimal radiation exposure was received by the workers and inadvertent entry into the area of concern (i.e., containment building in the area near the personnel air lock) would not immediately result in workers being in radiation fields greater than 1000 milliroentgen equivalent man per hour (Section 2OS1).
Inspection Report# : 2000004

Significance: G Sep 23, 2000
Identified By: NRC
Item Type: NCV NonCited Violation
Failure to Control Access to High Radiation Areas as Required by 10 CFR Part 20.1601 and TS 5.7.2
A single event, resulting in two non-cited violations, involved: (1) a failure to implement radiation control procedures for posting an extra high radiation area as required by TS 5.4.1.a.; and (2) failure to lock or control entrance to an extra high radiation area as required by Technical Specification 5.7.2 and Title 10 CFR Part 20.1601. This event was determined to be of very low safety significance because minimal radiation exposure was received by the workers and inadvertent entry into the area of concern (i.e., containment building in the area near the personnel air lock) would not immediately result in workers being in radiation fields greater than 1000 milliroentgen equivalent man per hour (Section 2OS1).
Inspection Report# : 2000004

Significance: G Jun 24, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Prevent the Release of Radioactive Byproduct Material from the Radiological Control Area and Plant Site
A non-cited violation was identified for the failure to comply with the requirements of 10 CFR 20.1802. Specifically, on April 7, 2000, the licensee failed to prevent the release of radioactive byproduct material (e.g., a radioactive particle on a contract employee's lanyard) from the radiological control area and plant site. Based on the activity of the particle and the resulting occupational dose assessment for the affected contract employee, this finding was determined to be of very low significance (Sections OS2, 2PS3).
Inspection Report# : 2000003

Physical Protection

Significance: G Jun 24, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Secure Two Vital Area Openings Exceeding 96 Square Inches in February 1999
A non-cited violation of the Physical Security Plan was identified for the licensee's failure to secure two vital area openings exceeding 96 square inches in February 1999. This issue was determined to have very little significance, given the non-predictable basis of the failures and the fact that there was no evidence that the vulnerabilities had been exploited (Section 3PP2).
Inspection Report# : 2000003

Miscellaneous

Significance: N/A Feb 16, 2001
Identified By: NRC
Item Type: FIN Finding
Identification and Resolution of Problems
Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. In general, the licensee properly prioritized items (by Action Category) in its corrective action program database, which ensured that timely resolution and appropriate causal factor analyses were employed commensurate with safety significance. Some exceptions were noted in the area of problem identification, where all relevant issues of problems were not identified and equipment performance was adversely affected. The inspection identified three exceptions in the area of prioritization and evaluation of issues, where more comprehensive root cause determinations would have provided more effective evaluations and corrective actions. In the area of effectiveness of corrective actions, it was noted that the corrective action program was not timely in resolving various documentation deficiencies with Technical Specification (TS) surveillances, Updated Final Safety Analysis Report changes and TS bases changes. Previous non-compliance issues documented as non-cited violations were properly tracked and resolved via the corrective action program. The results of the last comprehensive corrective action program audit conducted by the licensee (September 1999) were properly entered and dispositioned in the corrective action program. Based on discussions with plant personnel and the apparently low threshold for items entered in the corrective action program database, the inspectors concluded that workers at the site generally felt free to raise safety concerns to their management.
Inspection Report# : 2001003

Significance: N/A Dec 23, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Technical Specification 5.4.1 and Regulatory Guide 1.33, Section 7, for failing to have adequate procedures to control the release of radioactive material during a pressurizer gas space venting evolut
Technical Specification 5.4.1 and Regulatory Guide 1.33, Section 7, for failing to have adequate procedures to control the release of radioactive material during a pressurizer gas space venting evolution on October 14, 2000, as described in the licensee's corrective action program. Reference PIPs C-00-04914 and 05241.
Inspection Report# : 2000005

Last modified : May 03, 2001

Catawba 2

Initiating Events

Significance: G Sep 23, 2000
Identified By: Licensee
Item Type: FIN Finding
Reactor Trip Caused by Moisture Intrusion into Main Feedwater Pump 2B Speed Control Circuitry
Poor workmanship and inadequate oversight of turbine building roof repairs, coupled with inadequately constructed roof drainage systems, resulted in a June 5, 2000, Unit 2 reactor trip. Water from heavy rains that day could not be properly drained from the turbine building roof, partially due to debris and other roofing material that had collected in the drainage system. Water overflowed from the roof and into the turbine building, and leaked into the 2B main feedwater pump turbine speed control cabinet. A secondary plant transient resulted, which ultimately led to a turbine trip/reactor trip. This issue was determined to be of very low safety significance because it did not affect the ability of mitigating systems to perform their safety functions (Section 4OA3.1).
Inspection Report# : 2000004

Mitigating Systems

Significance: GMar 30, 2001
Identified By: NRC
Item Type: FIN Finding
Failed to Demonstrate Performance of the Station Drinking Water System as Backup Cooling Water to the Unit 1 and 2 A Train Charging Pumps
The licensee failed to demonstrate that the performance or condition of the station drinking water system, a risk-important system that provides backup cooling water to the Unit 1 and 2 A train charging pump motors and bearing oil coolers, was being effectively controlled through the performance of appropriate preventive maintenance (including surveillance activities). This resulted in a failure to recognize and correct a degraded system pressure condition, until it was identified by the inspectors. The degraded pressure condition was determined to be of very low safety significance because an analysis performed by the licensee demonstrated that the backup function to cool the charging pumps and motors would have been provided at the degraded pressure (Section 1R12.2).
Inspection Report# : 2000006

Significance: N/A Mar 30, 2001
Identified By: NRC
Item Type: FIN Finding
Failure to Identify Two Maintenance Preventable Functional Failures Affecting the Unit 2 Auxiliary Feedwater System
The inspectors identified a failure to identify two maintenance preventable functional failures (MPFFs) affecting the Unit 2 auxiliary feedwater system, one involving the turbine-driven auxiliary feedwater pump, the other involving the A motor-driven pump. Both of these occurred on October 5, 2000, following an inadvertent transfer of pump control to a local control panel. Although the finding did not involve a violation of the maintenance rule, it represented a recurring performance problem in this area as the latest of several missed maintenance preventable functional failure determinations involving different safety systems over the last year and a half. This finding was of very low safety significance because the failure to identify these MPFFs did not directly affect the ability of the auxiliary feedwater system to perform its safety function (Section 1R12.1).
Inspection Report# : 2000006

Significance: G Jun 24, 2000
Identified By: Licensee
Item Type: FIN Finding
Steam generator power operated relief valve 2SV-19 failed to open on April 15, 2000, due to mispostioned nitrogen pressure regulators
Steam generator power operated relief valve 2SV-19 failed to open on April 15, 2000, due to mispostioned nitrogen pressure regulators, which are required to function during a design basis event involving the loss of normally available instrument air. The licensee determined the mispositioned regulators to be a human performance issue, but were not able to pinpoint when the actual mispositioning took place. This issue was determined to have very low safety significance due to the availability of other steam generator power operated relief valves and diverse means of cooling the secondary plant (Section 1R22.2).
Inspection Report# : 2000003

Significance: G Jun 24, 2000
Identified By: NRC
Item Type: FIN Finding
Failure to properly classify a maintenace rule functional failure of the Unit 2 A steam generator power operated relief valve (2SV-19)
The licensee failed to properly classify a maintenace rule functional failure of the Unit 2 A steam generator power operated relief valve (2SV-19) when it failed to open on April 15, 2000. The licensee incorrectly assumed that the valve's failure was not a functional failure because other redundant valves were available at the time. This issue was determined to have very low safety significance because the licensee's error did not result in additional equipment unavailability (Section 1R12.1).
Inspection Report# : 2000003

Significance: GJun 24, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Provide Adequate Procedures for Performing Maintenance on Safety-Related Sump Pump Level Switches
Residual heat removal and containment spray pump room sump level alarm function was lost for several months up to February 2000 due to inadequate maintenance procedures associated with sump level switch calibrations. This issue was characterized as a non-cited violation of Technical Specification 5.4.1 and was determined to be of very low safety significance due to the availability of other emergency core cooling system leak detection methods (Section 4OA3.2).
Inspection Report# : 2000003

Barrier Integrity

Significance: G Jun 24, 2000
Identified By: NRC
Item Type: FIN Finding
Failure to properly evaluate plant risk associated with emergent work for the Unit 2 hydrogen ignition system on April 27, 2000.
The licensee did not properly evaluate plant risk associated with emergent work for the Unit 2 hydrogen ignition system on April 27, 2000. As a result, the unit was in an unevaluated increased risk condition while planned work associated with the containment spray system was ongoing. This condition was allowed by Technical Specifications and plant procedures, but plant procedures required that a written contingency plan be developed prior to the work commencing, which was not done. This issue was of very low safety significance due to the availability of diverse and redundant systems designed to accomplish the hydrogen mitigation and containment pressure control functions (Section 1R13).
Inspection Report# : 2000003

Occupational Radiation Safety

Physical Protection

Miscellaneous

Last modified : May 03, 2001

McGuire 1

Initiating Events

Significance: G Mar 17, 2001
Identified By: Licensee
Item Type: NCV NonCited Violation
Inadequate Corrective Actions for Recurring Problems with Shutdown Operations Involving Loss of Letdown and/or Inadvertent Reactor Coolant System Cooldown Transients
Inadequate corrective actions (10CFR50, Appendix B, Criterion XVI) for recurring problems with shutdown operations involving loss of letdown and/or inadvertent reactor coolant (NC) system cooldown transients. During a Unit 1 shutdown from Mode 2 to Mode 3 on March 9, 2001, NC system temperature went below minimum temperature for criticality due to overfeed of steam generators. This event occurred because of ineffective corrective actions to address procedural deficiencies and/or equipment problems complicating plant cooldown. This is captured in the licensee's corrective action program under PIP M-01-0986. This finding was determined to have very low safety significance and is being treated as a Non Cited Violation (Section 4OA7).
Inspection Report# : 2000007

Mitigating Systems

Significance: G Dec 15, 2000
Identified By: NRC
Item Type: FIN Finding
Depth and effectiveness of the licensee's evaluation and corrective actions for failures of the standby shutdown facility (SSF) diesel generator.
A finding was identified associated with the depth and effectiveness of the licensee's evaluation and corrective actions for failures of the standby shutdown facility (SSF) diesel generator. The licensee's corrective actions for recent SSF-related problems have not been commensurate with the risk significance of the system. A recent Problem Investigation Process report, which documented a jacket water coolant leak and subsequent emptying of the engine's radiator, was not screened to include a root cause evaluation. The licensee did not perform comprehensive corrective actions to evaluate the need for performing additional preventive maintenance on the SSF diesel generator components. The inspectors identified vendor-recommended maintenance practices that were not being implemented and service bulletins authored by the vendor that were not included in the associated controlled vendor manual located on site. This issue was determined to have very low safety significance because it was not directly linked to any specific period of unavailability for the SSF diesel generator. This instance of ineffective corrective action was an isolated example and is not considered indicative of the licensee's overall corrective action program. (Section 4OA2b).
Inspection Report# : 2000010

Significance: G Jun 17, 2000
Identified By: Self Disclosing
Item Type: NCV NonCited Violation
Failure to Follow Emergency Procedure Concerning Auxiliary Feedwater Suction Supplies
A non-cited violation of Technical Specification 5.4.1.a was identified for two examples of the licensee's failure to follow the emergency procedure generic enclosure used for maintaining auxiliary feedwater (CA) suction sources during reactor trip recovery. This resulted in the inadvertent isolation of the preferred CA suction supply and actuation of the service water system to provide CA to the steam generators. A lack of training and familiarity with the applicable emergency procedure generic enclosure was found to be a contributor to this finding. The safety significance of this violation was very low because the CA system was able to perform its function of steam generator decay heat removal (Section 04.03).
Inspection Report# : 2000008

Physical Protection

Significance: G Sep 16, 2000
Identified By: NRC
Item Type: NCV NonCited Violation
Failure of the Electronic Switching to Provide the Central Alarm Station Operator with the Capability to Properly Assess Potential Penetrations at the Perimeter Prior to Individuals Gaining Access
A non-cited violation of the Physical Security Plan was identified for the failure of the licensee's electronic switching on September 12, 2000, to provide the central alarm station operator with the capability to properly assess potential penetrations at the perimeter prior to individuals gaining access to the protected area (Section 3PP3.2)
Inspection Report# : 2000005

Miscellaneous

Significance: N/A Dec 15, 2000
Identified By: NRC
Item Type: FIN Finding
Identification and Resolution of Problems
Overall, the licensee's corrective action program was effective at identifying, evaluating, and correcting problems. The threshold for entering problems into the corrective action program was sufficiently low. Reviews of operating experience information were comprehensive. In general, the licensee properly prioritized items (by Action Category) in its corrective action program database, which ensured that timely resolution and appropriate causal factor analyses were employed commensurate with safety significance. One exception involved a recent condition adverse to quality in which the standby shutdown facility's (SSF) diesel generator was unavailable following the complete draining of radiator coolant because of heater shell pin-hole leaks. The licensee did not perform an in-depth root cause analysis and thorough corrective actions following its discovery of the degraded condition. Also, for potential safety equipment operability issues, the licensee did not always conduct or document thorough evaluations of present or past inoperability. Previous non-compliance issues documented as non-cited violations were properly tracked and resolved via the corrective action program. The results of the last comprehensive corrective action program audit conducted by the licensee (September 1999) were properly entered and dispositioned in the corrective action program. Based on discussions with plant personnel and the apparently low threshold for items entered in the corrective action program database, the inspectors concluded that workers at the site generally felt free to raise safety concerns to their management.
Inspection Report# : 2000010

Last modified : May 03, 2001

McGuire 2

Mitigating Systems

Significance: G Mar 17, 2001
Identified By: Licensee
Item Type: NCV NonCited Violation
Failure to Follow Procedure PT/2/A/4350/026C, Auxiliary Shutdown Panel Verification
Failure to follow procedure (Technical Specification 5.4.1) for PT/2/A/4350/026C, Auxiliary Shutdown Panel Verification. The procedure indicates that all manipulations of controls at the panel shall be performed by a licensed reactor operator. A non-licensed operator performed the auxiliary shutdown manipulations during the performance of the test, contrary to the requirements of the procedure. This is captured in the licensee's corrective action program under PIP M-00-4140. This finding was determined to have very low safety significance and is being treated as a Non Cited Violation (Section 4OA7).
Inspection Report# : 2000007

Significance: G Dec 16, 2000
Identified By: Licensee
Item Type: NCV NonCited Violation
Inadequate procedure for removal of 120VAC inverters from service
Inadequate procedure (TS 5.4.1) for removal of Unit 2 120VAC vital inverters from service. During plant solid RCS operation in Mode 5, de-energizing the vital inverters resulted in an inoperable Low Temperature Overpressure Protection (LTOP) system required by Technical Specification 3.4.12. The finding was determined to have very low safety significance (Section 4OA7).
Inspection Report# : 2000006

Physical Protection

Miscellaneous

Last modified : May 03, 2001

This item appeared in The Times & Free Press on Wednesday, December 29, 1999.

NRC Urges TVA Develop Better Plan For Ice Backup System at 2 N-Plants

By DAVE FLESSNER
Business Editor

Federal regulators have given a cold shoulder to TVA's plan to verify the amount of ice in an emergency backup system at the Sequoyah and Watts Bar nuclear power plants.

The Nuclear Regulatory Commission has asked TVA to come up with a better method of weighing the 3 million pounds of ice in the containment walls at each plant. The NRC's rejection this month of the initial plan prolongs a troubling issue for TVA and the other operators of the nine Westinghouse-designed reactors that use ice condensers as part of their safety systems.

Debris problems in the condenser system at the Donald C. Cook nuclear plant in Michigan helped force a 2-year repair outage at that plant. A TVA whistleblower made similar claims in 1995 at the Watts Bar Nuclear Plant. Last year, a federal judge ordered TVA to rehire Curtis Overall, who claimed he lost his job for reporting that he found 200 screws and fragments in the ice condensers at Watts Bar.

"Here we are nearly five years after this issue surfaced at Watts Bar and more than two years after the D.C. Cook plant was shut down because of condenser problems and TVA has still been unable to fix this problem," said Jim Riccio, an attorney for the Public Citizen's Critical Mass Energy Project in Washington. "TVA is not meeting plant regulations and even the NRC, which I think has largely been taken over by the nuclear industry, recognizes that."

TVA officials insist that the ice condenser system is still reliable in the unlikely event of an accident at one of its reactors.

"We're going to supply the NRC more information about what we propose to be a part of the technical specifications to monitor the performance of the ice condensers," TVA spokeswoman Barbara Martocci said.

Robert Martin, the NRC project manager overseeing the review of the ice condenser issue, said TVA and NRC officials should meet next month to discuss the issue.

"We certainly haven't found anything that calls us to shut down any of these plants (with the ice condensers)," he said.

The ice condenser systems are designed to relieve pressure, temperature and the possibility of a radioactive release if a steam pipe breaks in a reactor containment building. When the system works, the steam hits the condenser ice and is turned to water. The water is captured and never leaves the building.

The ice condensers allowed Westinghouse to design a smaller and less expensive reactor containment building at its pressurized water reactors.

TVA, Duke Power Co. and Michigan Power Co. are the utilities that own Westinghouse reactors that use ice condensers. The utilities have formed an Ice Condenser Mini Group to address concerns about the reliability of the condensers. TVA is responsible for developing technical specifications to weigh the ice in the plants.

In a recent letter to TVA, NRC project manager Ronald W. Hernan rejected TVA's initial plan, claiming that nearly half of the 1,994 ice baskets at Sequoyah can't be adequately weighed under the proposed surveillance plan.

The NRC continues to give TVA high marks for the operation of its nuclear plants overall, however. The Sequoyah plant is part of a new pilot reporting system the NRC launched this year and the plant is rated in the top "green" category of all areas of plant performance monitored by the NRC.

"TVA continues to maintain all the systems in top working order and we will work to address this ice condenser issue," Ms. Martocci said. "We are very comfortable in saying that we believe our plants will operate as designed."