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Frequently Asked Questions About the Vermont YankeeOn this page: Extended Power Uprate
Index to All Frequently Asked Questions Pages
NRC believes that the detailed engineering inspection described in the response satisfies the PSB request for system coverage and inspection size. The inspection will sample components across multiple systems to verify that design bases have been correctly implemented and to pro-actively identify latent design issues. The inspectors will use risk and engineering methods to select risk-significant components from multiple systems. The pilot inspection incorporates the “vertical slice” practice of existing and past design and engineering inspections to review the design, operation, maintenance, and testing aspects of the components selected. Although the specific sampling of components has not been finalized, they will come from at least two safety-related systems (emergency core cooling, containment, or auxiliary) and two maintenance rule, non-safety-related systems (power conversion or auxiliary). The PSB suggested that the NRC assessment should involve about 16 on-site inspector weeks. The planned inspection will involve about 18 on-site inspector weeks. We are also performing a number of other inspections that will look at issues specifically related to power uprate. NRC believes that our detailed technical review of the Vermont Yankee power uprate request and inspections, as well as the detailed engineering inspection, will provide the information necessary for us to make a decision on the safe operation of Vermont Yankee under uprated conditions and will assist the PSB in informing the Board’s decision on future reliability of Vermont Yankee. How will the team choose the systems and components to be inspected? The scope will be determined using a graded selection criteria designed specifically to identify risk significant areas where components have low margins and/or where human actions are critical. This is intended to focus inspection efforts on design and engineering areas which have a high degree of safety significance. Additionally, the scope identification process will be informed by industry operating experience. For Vermont Yankee, the components and human actions selected will also be informed by issues related to the power uprate review process. How will findings from this inspection be managed? The results of the pilot inspection will be evaluated using the NRC’s significance determination process. In addition, as part of the pilot process, the NRC will assess the need to make additional enhancements to its inspection program to address any programmatic concerns identified by the inspections. How will you ensure the independence of the engineering inspection team ? The inspection team will consist of seven members including: a team leader, three contractors with design experience, and three experienced NRC inspectors. An observer from the State of Vermont will also accompany the team. To ensure the independence of the team, the contractors to be selected must have: (1) never been directly employed by Entergy or Vermont Yankee; (2) not performed contract work for Vermont Yankee or Entergy within the last two years; and (3) not performed inspections for the NRC at Vermont Yankee within the last two years. The NRC inspectors will not be current or former resident inspectors at Vermont Yankee or have participated in an engineering inspection at Vermont Yankee within the last two years. Is this the same size/type of inspection as the independent safety assessment that was conducted at Maine Yankee? The review being conducted of the proposed power uprate at VY is of a different scope and size than the independent safety assessment (ISA) conducted at Maine Yankee because the purpose is different. The ISA at Maine Yankee was performed in response to an investigation of allegations regarding use of a computer code. Because of the nature of these allegations and the involvement of the Inspector General, the NRC determined that the assessment at Maine Yankee should be conducted by persons independent of the routine oversight of Maine Yankee. The NRC decided to conduct the assessment at Maine Yankee because we had significant concerns with conformance to license requirements; a situation that does not exist currently at VY.
What is the difference between the PSB’s role and the NRC’s role in ensuring safety of the plant? The type of analysis recommended by the PSB focuses on the reliability of the plant and not the plant’s safety. The PSB is precluded under federal law from reviewing radiological concerns regarding the power upgrade. Specifically, the PSB’s stated concern is the effect that a large power uprate will have on the reliability of the plant output in light of recent difficulties that have been experienced by other boiling water reactors following extended power uprate implementation. NRC regulations and oversight process focus on ensuring nuclear safety regardless of the mode of plant operation. The NRC’s statutory authority does not extend to regulating the reliability of generator output. Decisions on power generation are left to the discretion of the licensee. The PSB’s concern focuses on the reliability of VY. To the extent that proper operation and maintenance of the plant affect the reliability of plant systems and, by extension, the reliability of plant output, the staff notes that many licensees have established performance criteria related to minimizing system unavailability as part of their maintenance rule implementation. The NRC’s Reactor Oversight Process (ROP) measures the number of scrams at a plant as one of its performance indicators, reflecting the relationship between stable and consistent operation and safety. In addition, the ROP baseline inspection program includes modules that assess licensees’ maintenance rule programs and implementation. Additionally, the ROP includes engineering inspections such as the "Safety System Design and Performance Capability" module that conduct vertical slice assessments. Why was an ISA performed at Maine Yankee? In December 1995, anonymous allegations were sent to the State of Maine and to the NRC regarding inadequate licensee power uprate analyses. The NRC Inspector General initiated an event inquiry into the allegations. The NRC Chairman, in response to the OIG report, and the Governor of the State of Maine, directed the staff to perform an ISA at Maine Yankee to address OIG issues, relative plant performance, and evaluate root causes of identified findings. What was the size and scope of the ISA that was performed at Maine Yankee in 1996? The Maine Yankee ISA team consisted of 25 members. In addition to three State representatives, six NRC contractors and an administrative assistant participated on the team. The team spent 4 weeks conducting onsite inspections. The ISA was conducted over a 3-month period. The team was independent in the sense that it did not include any Region I inspectors and the team reported directly to the Director of the Office for Analysis and Evaluation of Operational Data, and not to the regional management.
What does the NRC’s review of power uprates encompass? Our review is independent, thorough and comprehensive. A team of engineers and scientists with specialties in 17 different technical areas have started to review the VY power uprate application. The Review Standard (RS-001) is the document that guides the staff in the review. It takes into account lessons learned from previous power uprate reviews, including lessons learned from the Maine Yankee experience. In addition to reviewing information in Entergy’s application, we may perform audits at the plant or vendor sites as well as do our own confirmatory analyses. The RS provides guidance for determining when and what type of audits and confirmatory analyses are appropriate to supplement the review. What type of inspection is usually performed for a power uprate? Our procedure (Inspection Procedure 71004, "Power Uprates") focuses inspection activities on issues specifically related to power uprates such as changes that impact the integrity of barriers (e.g., higher flow rates which increases flow-accelerated corrosion and vibration at specific support points.) We also make use of the information we gain through the routine inspection activities that the NRC resident inspectors conduct every day. When appropriate, we will focus the inspection on risk-significant power uprate activities (e.g., plant modifications, post-maintenance testing, integrated plant operations). We began inspections of power uprate activities in April 2004 when Entergy made plant modifications that may later be used to operate the plant at the proposed power level. What actions were taken at VY as a result of the ISA at Maine Yankee? After the Maine Yankee ISA, we evaluated the computer programs at all plants, including VY, to ensure that the issues found at Maine Yankee did not exist at other plants. Two additional recommendations related to engineering inspections resulted. They were to (1) conduct design inspections at selected facilities and (2) modify the inspection procedure (IP-37550) to add periodic "vertical slice" engineering inspections. The first recommendation was subsumed in the Reactor Oversight Process (ROP) as an engineering baseline inspection. The second recommendation spurred the creation (along with the outcomes from the Millstone experience) of the architect-engineer (A-E) inspection program. A-E inspections were eventually conducted at 21 plants, including VY. It’s important to note that the NRC also issued a request for information to licensees pursuant to 10 C.F.R. § 50.54(f) in October 1996, regarding the adequacy and availability of design-basis information. The purpose of this request was to provide the NRC with added confidence and assurance that nuclear plants are operated and maintained within the design bases and any deviations are reconciled in a timely manner. This request was necessary on the basis of NRC's findings during inspections and reviews that identified broad programmatic weaknesses that have resulted in design and configuration deficiencies at some plants, including Millstone. Based on our assessment of VY’s response to our request, we concluded that an A-E inspection at VY was warranted. Many other Region I plants (Pilgrim, Millstone, and Salem for example) have undergone rigorous system reviews over the past 20 years. Why hasn’t this been performed at VY? The reviews conducted at other plants were done because of known performance problems or significant operational events. These kind of problems have not been observed at VY. Since 2000, we have performed a total of over 860 hours of safety system design inspections at VY. In 1997, we performed an A-E inspection to evaluate the capability of selected systems to perform the safety functions required by the design bases, the adherence of the systems to their design and licensing bases, and the consistency of the as-built configuration and system operations with the Final Safety Analysis Report. We concluded that the systems we evaluated were capable of performing their intended safety functions. Further inspections reviewed the licensee’s design basis documentation effort. Our inspection concluded that overall, the licensee’s activities in this area have been rigorous. Our inspection program has clear criteria for escalating the level and depth of inspections and we have not hesitated to perform such detailed inspections when conditions call for them. We have not observed these conditions at VY. What opportunities does the public have to participate in the power uprate review? Our process for reviewing license amendments includes placing a notice in the Federal Register and providing the public an opportunity to comment or seek hearing. The Federal Register notice for Vermont Yankee’s power uprate request was published on July 1, 2004. For all EPUs, we perform an environmental review to assess potential environmental impacts. The environmental assessment is also published in the Federal Register for public comment. Also for EPU applications, the Advisory Committee on Reactor Safeguards (ACRS) reviews our draft safety evaluation. The ACRS review process involves a public meeting between the ACRS and the NRC staff to discuss the draft safety evaluation. All portions of this meeting except discussions of any proprietary information are open to the public. The public may also make oral statements during the meeting. Has a 20% uprate ever been approved by the NRC? Yes. We approved a 20-percent uprate at the Clinton Nuclear Power Station on April 5, 2002. Clinton has not experienced any technical problems associated with this power uprate. What are the effects on operating temperatures and pressures, emergency core cooling system pumps, etc.? In evaluating a power uprate request, the NRC reviews information submitted by the licensee to confirm that the plant can operate safely at the higher power level. The information reviewed includes changes in operating parameters (e.g., temperature, pressure) to confirm that the applicable structures, systems, and components (e.g., emergency core cooling pumps) will continue to perform their intended safety functions. Based on the fact that the NRC has never denied a power uprate request, how can we have assurance that they are sufficiently independent and objective enough to do the right thing? The NRC has often denied licensee requests. Since 1993, we have denied approximately 400 requests to use alternative methods for testing pumps and valves in nuclear power plants. More frequently, however, when it becomes clear that we are not likely to approve a request, the licensee will withdraw it. So far this year (FY04) licensees have withdrawn 20 licensing actions. Last year (FY03), licensees withdrew 36 of them. Even more frequently than withdrawing an amendment, the licensee will modify its request based on our review. Our licensing process involves a lot of interaction between our technical experts and the licensee. As we review the application and perform confirmatory calculations, we ask questions. They respond. Most times, they modify their request, sometimes they withdraw it, and sometimes, we reject it. Vermont Yankee is close to 30 years old. Isn’t a 20% uprate too much? We haven’t completed our review of Entergy’s application so we don’t know if 20-percent uprate is too much. We will only approve a license amendment request if there is reasonable assurance that the health and safety of the public can be protected by operation in the proposed manner. Vermont Yankee may be close to 30 years old, but the components that make up the plant undergo routine rigorous testing. We require plants to inspect and test pumps, valves, piping, and the functionality of safety systems on a routine basis. Some components are tested every 3 months. Some are tested every outage. This routine testing and inspection verifies that the components and systems can adequately perform their function if called upon. In addition, the plant has, over the years, performed maintenance and replaced equipment as necessary to keep systems in top condition. Why did the NRC issue a letter to Entergy Nuclear notifying them that the proposed amendment to allow a power uprate at Vermont Yankee Nuclear Power Station was incomplete? Several areas in Entergy’s initial application and subsequent submittals required additional information to allow us to complete the detailed review of the application. Some of the areas for which insufficient information was provided include: steam dryer integrity, flood protection equipment, internally generated missiles, and ultimate heat sink. Entergy submitted this additional information at the end of January 2004. We performed an acceptance review of this information and determined that Entergy has provided the necessary information such that the detailed review can be expected to be completed within a 12-month review schedule. Will the power uprate affect the amount of spent fuel which is currently stored on site? Will the power uprate generate more nuclear waste? Based on our understanding of the spent fuel pool capacity at VY, the ability to perform a full core offload into the spent fuel pool will be shortened by 1 year if the power uprate is implemented using Entergy’s current plan. VY will continue to store its spent fuel in accordance with existing requirements and guidelines. When nuclear plants run out of storage room for spent fuel in the spent fuel pool, an option is to store the fuel in dry cask storage. The NRC has authorized safe storage of spent fuel in dry casks at many nuclear plants with no adverse effects. A typical analysis for power uprate is to verify that an unplanned full core offload of fuel will not challenge the integrity of the spent fuel pool or the temperature guidelines established for the plant. What changes have to be made to the plant to accommodate such a large increase in power? Typically, equipment modifications are necessary to handle the additional stresses from increased flow and power generated. Most equipment modifications required for a power uprate are in the turbine building because they affect power generation. Some components that help achieve safe shutdown, however, will be affected by the uprate. This is a primary focus of the NRC’s review. What are the environmental effects from an extended power uprate? Are fish going to be harmed due to warmer temperatures? Is there going to be an increase in the amount of radiation that is routinely released into the environment? The temperature of the water returned to the river from the plant is primarily regulated by agencies other than the NRC. However, the NRC recognizes the importance of thermal pollution and its contribution to environmental issues. If the water temperature approaches the established temperature limit, operators have options available such as placing the plant’s existing cooling towers in service or may reduce power in order to avoid exceeding any limits. The NRC will review the safety-related function of the cooling towers as part of the power uprate review. Based on past experiences with power uprates, the NRC is not aware of any significant effects from the small temperature increase. At a higher power level, more radioactive fission products will accumulate in the fuel. During regular operation, the radiological release rate of many radwaste systems is administratively controlled, and does not change with operating power. During a postulated accident, an increase in the radioactive fission products in the fuel could lead to an increase of radiation released into the atmosphere during a postulated accident. However, prior to approving such an uprate, the NRC will verify that all plausible releases into the environment are bounded within the limits set by Federal regulations. The licensee is required to provide the NRC with this analysis with their power uprate application. Have there been any problems that were not reviewed during the amendment review that have occurred post-power uprate? Recently there have been cases of steam dryer cracking and flow-induced vibration damage on components and supports for the main steam and feedwater lines at some of the power plants that are operating at an uprated power level. Our regional inspectors performed inspections to identify the causes of these issues and evaluated many of the repairs performed by the licensees. We continue to monitor the industry’s response to these issues and will consider the need for additional regulatory action, if needed. In addition, the NRC has requested specific information from Entergy regarding steam dryer cracking and flow-induced vibration as part of its review of the VY EPU application. Quad Cities had to temporarily shutdown due to a safety problem shortly after it increased its power output to 18%. Doesn’t this show that the NRC process is flawed and that it’s not safe to increase the power? No. The increased power level did not compromise the ability to safely shut down the reactor. As discussed in NRC Special Inspection Report 50-265/03-11 dated August 7, 2003, the licensee for Quad Cities shut down the reactor due to cracking of the reactor steam dryer from flow-induced vibrations as a result of the EPU. The root cause of this cracking was determined to be insufficient industry experience and knowledge of flow-induced vibration dryer failures. The dryer cracked as a result of fatigue caused by flow-induced vibrations created by higher steam flows. An inspection revealed that the steam dryer cover plate cracked, which allowed steam to bypass the dryer. Corrective actions were implemented. The lessons learned from the Quad Cities experience are being applied to our review of subsequent power uprates. In addition, the NRC issued Information Notice 2002-26 to notify the industry of the issue and is considering further regulatory action. Why doesn’t VY have to adhere to the same licensing requirements that a new plant being licensed would have to adhere to? The NRC frequently updates its regulations as a result of improvements to technology and based on operating experience. When requirements are changed, the NRC applies a rigorous evaluation standard to determine if the safety benefit of the new requirements justifies imposing the changes on existing licensees. For example, VY was licensed before the final General Design Criteria (GDC) were adopted. At the time it was licensed, VY was evaluated by the NRC on a plant-specific basis, and was determined to be safe. When the final GDC were approved, the Commission stressed that the final GDC were not new requirements but were promulgated to more clearly articulate the licensing requirements and practice in effect at that time. The Commission determined that imposing the final GDC on plants licensed prior to issuance of the final GDC would provide little or no safety benefit. However, when a new regulation provides a substantial increase in safety other cases, the NRC has imposed new regulations on nuclear facilities (e.g., environmental qualification of electrical equipment). 
What is the concern related to credit for containment overpressure? Will the NRC approve the request by VY to credit containment overpressure? In the ongoing review of the Vermont Yankee proposed license amendment to increase licensed power, the NRC is reviewing the licensee's calculation of the net positive suction head, or NPSH, available for the emergency core cooling pumps should they be required for emergency operation. The staff requires numerous conservatism in those calculations to ensure that under the worst case the available NPSH will not be underestimated. One of the conservatisms that the staff has previously relaxed in the direction of realism is credit for pressure in containment that results from the postulated accident itself. About 20 boiling water reactors have been allowed to formally credit accident overpressure in NPSH calculations, following staff review to ensure that adequate safety margin is maintained by the remaining conservatisms. Vermont Yankee's amendment request to credit accident overpressure is receiving the same scrutiny, and no decision has yet been made on the acceptability of their proposal. The licensee has been asked to assess the additional risk of crediting accident overpressure. A number of issues have been raised as a result of the Vermont Yankee amendment request concerning NRC's past practice of crediting containment overpressure. NRC is in the process of reviewing this practice, and when complete will update its regulatory guidance to further clarify its technical position. How can the NRC approve the VY request to credit containment overpressure since it is not in conformance with NRC regulations? The NRC staff will allow credit for containment accident pressure to be taken only when the licensee's analyses and justification for the proposed licensing basis change demonstrate that there is reasonable assurance that the credited pressure will exist for the events and time period for which the credit is required. Ensuring containment integrity and avoiding overcooling of the containment due to excessive use of containment sprays are key considerations in determining whether the credited pressure will be available during the required time period. The NRC's guidance regarding whether it is acceptable to credit containment accident pressure has evolved over the years. The current guidance is contained in Regulatory Guide (RG) 1.82, Revision 3, "Water Sources for Long-Term Recirculation Cooling Following a Loss-Of-Coolant Accident" dated November 2003. This RG describes methods acceptable to the NRC staff for implementing requirements with respect to the sumps and suppression pools performing the functions of water sources for emergency core cooling, containment heat removal, or containment atmosphere clean up. The NRC staff has discussed the need for more clarity in the guidance on credit for containment accident pressure. We realize the fact that we did not formally withdraw RG 1.1, which did not allow credit for containment accident pressure, has led to confusion about our technical position. We will both formally withdraw RG 1.1 since it has been superceded, and revise RG 1.82 to more clearly explain how credit for containment accident pressure can be found acceptable. Additional information on this issue can be found in our letter to the Vermont Department of Public Service dated June 29, 2004. Based on the flaws in the Reactor Oversight Process (ROP), as evidenced by the failure of the process to identify problems at Davis Besse, how can the public be assured that it will provide accurate assessments of safety at VY? The corrosion of the reactor vessel head at Davis Besse existed under both the former inspection process and the current ROP. We believe that no inspection process will always preclude new or unexpected failure mechanisms such as the corrosion of Davis-Besse reactor vessel head. This assessment is based on the sampling nature of any inspection program. In light of the event at Davis Besse, the Davis Besse Lessons Learned Task Force (LLTF) performed a comprehensive review of NRC's processes. One of the focal points of that review was the implementation and effectiveness of the inspection program at Davis Besse. Although the reactor oversight program is fundamentally sound, it was clear from Davis Besse that we needed to learn from our experience. The LLTF recommended several changes to a broad scope of NRC activities including enhancements to the inspection program. NRC senior management reviewed the recommendations and accepted 49 of the 51 for NRC action and ranked them as high, medium, or low priority. NRC is currently implementing an action plan to address the recommendations and has currently completed about a third of the actions. What is not reflected in that statistic is that significant progress has been made on many other actions that are not yet complete. One of the key changes was a renewed awareness with NRC inspectors and managers that NRC needs to be vigilant, to ask questions, to not become complacent in reviewing potential safety issues. There has been increased emphasis on inspector review of out of specification conditions and on aggressive followup to issues. NRC has also increased its inspections by regional inspectors and implemented measures to assure resident inspector continuity at the sites. We will continue to look for ways to learn from our experiences to provide continued assurance of public health and safety. Location of Spent Fuel Rod Segments
Who discovered them missing, Entergy or the NRC Resident Inspector? Entergy discovered the missing fuel rod segments after Entergy conducted a detailed inventory check for these two fuel rod segments. Entergy conducted the inventory check after the resident inspector staff identified that Entergy’s and its predecessor’s annual inventory check associated with these two spent fuel rod segments was not a “piece count” of the actual rod segments.
In 1980, during a fuel assembly reconstitution effort, two fuel rods broke into several pieces. Two of these pieces were short, one about 7 inches and the other about 17 inches. A fuel assembly consists of many fuel rods. Although the remaining segments of the fuel rods were retained in spent fuel assemblies stored in the spent fuel pool, these two short fuel rod segments could not be stored within a spent fuel assembly. Instead, these two segments were placed into a separate container located on the bottom of spent fuel pool, below approximately 40 feet of water.
In light of the current terrorist threat, what licensee/NRC plans were in place to prevent this from happening? NRC regulations require accurate accounting and complete control of all nuclear material. The regulations that apply to accounting and control of nuclear material at power reactors are 10 CFR 74.13, 74.15 and 74.19. Because of the current threat and because of the lessons learned from the Millstone 1 event, NRC put in place a temporary instruction directing resident inspectors to review the material accounting and control programs at power reactors. This review is the reason that the problem with the rod segments at Vermont Yankee was identified. The temporary instruction was developed to clarify the extent of possible historical material control and accountability problems at reactors, and to provide guidance for putting in place programs to prevent such problems in the future.
Is there material in the missing fuel pins that terrorists could use, say, to make a dirty bomb or other dangerous weapon? The material is radioactive. Because it is radioactive, any attempt to remove it from the spent fuel pool and to move it off-site without the proper shielding would cause the radiation monitors to alarm. It would be unrealistic to speculate a scenario in which this material could have been taken offsite without the licensee’s knowledge.
Doesn't Entergy have procedures/safeguards to ensure this cannot occur? Entergy has procedures to control and document the insertion, movement, and removal of material in the spent fuel pool. In addition, Entergy has procedures to perform an annual physical inventory of material in the spent fuel pool. In this case Entergy may not have accurately documented or adequately controlled the movement or removal of the fuel rod fragments in the spent fuel pool.
Is this a generic problem? Can we assume this is a problem at other nuclear power plants? All nuclear plants will be inspected using the temporary instruction that was used at Vermont Yankee. Inspections were completed at the end of May 2004. The results of these inspections will be used to determine which plants should receive a more detailed review.
Does the licensee have responsibility for everything sent off-site? The licensee is responsible to assure that a shipment of radioactive material off-site complies with the appropriate NRC and DOT requirements. The actual material shipped off site must agree with the documented list describing the contents of the shipment. In addition special shipping containers must be used depending on the amount and type of radiation within the shipment.
How does this incident differ from a similar event a few years back at the Millstone nuclear plant? The incident at Millstone involved two full fuel rods each about 12 feet long for a total of 24 feet of fuel rod. The two short fuel rod segments involved about 2 feet of fuel rod.
Development of the inspection instruction that was used by the inspectors to question VY's accounting practices was initiated shortly after the Millstone event. However, it was postponed following 9/11 due to competing priorities and resources in the security area.
Since this the second nuclear plant that is missing fuel rods what is the agency doing about the other nuclear plants? The temporary instruction is being performed to give the NRC information about the material accountability programs at all power reactors. As a result of the events at Millstone and Vermont Yankee and after examination of the results of the TI at other reactors, the NRC is will reexamine its routine inspection priorities.
Is there an NRC person overseeing actions at the site? A special inspection has been initiated looking into the potentially missing spent fuel rod segments. We also have our two normally assigned resident inspectors monitoring Entergy’s actions.
How long do we require plants to hold onto their shipping records? The requirement says only 3 years. 10 CFR 74.19 says that records must be retained for as long as the licensee possesses the material plus three years.
The resident inspector’s statement on March 31, 2004, was based on information available to the resident inspector at that time. In March 2004, the resident inspectors conducted an inspection that looked at control and accounting for spent fuel in accordance with Temporary Instruction (TI) 2515/154 “Spent Fuel Material Control and Accounting at Nuclear Power Plants.” The inspectors reviewed Vermont Yankee’s processes for tracking the location of spent fuel, principally irradiated fuel rods that had been removed from fuel assemblies. They obtained information through interviews and review of spent fuel records to answer 12 specific questions contained in the TI. The questions required, in part, that the inspectors obtain a list of all irradiated fuel rods that have been removed from their parent assembly, determine the location of these fuel rods in the spent fuel pool, and to the extent possible by observing from the edge of the spent fuel pool, determine whether the fuel rods were in the recorded locations. Other questions required the inspectors to determine whether the Vermont Yankee staff conducted an annual physical inventory of special nuclear material. During their review, the resident inspectors identified that the method that Vermont Yankee staff had been using to perform the physical inventory for two fuel rod segments did not comply with Vermont Yankee’s procedure. These segments were stored in a container located on the bottom of the spent fuel pool. Rather than visually “counting” these two fuel rod segments, Vermont Yankee staff ensured the container where these two fuel rod segments were stored remained upright and in place on the bottom of the spent fuel pool. In response to the inspectors’ questions, on March 26, 2004, Vermont Yankee staff looked from the refueling floor to determine if they could confirm that the two fuel rod segments were located in the storage container on the bottom of the spent fuel pool. The staff believed they saw some indication that the fuel rod segments were in the container, but the indication was not of sufficient detail and clarity to definitively make that conclusion. The Vermont Yankee staff planned to perform an inspection during the April 2004 refueling outage that would provide the detail and clarity to confirm that the two fuel rod segments were in the container.
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