United States Nuclear Regulatory Commission - Protecting People and the Environment

Baffle-Former Bolts

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Baffle-former bolts help hold together a structure inside the reactor vessel of Westinghouse pressurized water reactors (PWRs).  As part of the license renewal process, licensees have committed to periodically inspect the reactor vessel internals, including the baffle-former bolts, for indications of degradation.  Recent operating experience and inspections have identified more baffle-former bolts with indications of degradation than anticipated.  In spring 2016, two PWRs, Indian Point Unit 2 and Salem Unit 1, identified a large number of degraded baffle-former bolts during refueling outage inspections.  Indian Point Unit 2 and Salem Unit 1 found and reported these degraded bolt conditions in event notifications (EN): EN 51829 and EN 51902.  Subsequently, two additional PWRs, D.C. Cook, Unit 2, and Indian Point, Unit 3 have identified similar numbers of degraded baffle-former bolts.

Operating experience indicates that the baffle-former bolts are more susceptible to degradation in older Westinghouse four-loop reactors that have a "down-flow" configuration and have baffle-former bolts made of Type 347 stainless steel.  There are seven U.S. reactors that match these characteristics:  Indian Point Units 2 and 3, Salem Units 1 and 2, D.C. Cook Units 1 and 2, and Diablo Canyon Unit 1 (Diablo Canyon Unit 2 has a different configuration than Unit 1).

The NRC's risk-informed assessment of the issue determined that degraded baffle-former bolts do not warrant the immediate shutdown of any plant.  The issue does not present a significant safety concern.

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There are structures located within Westinghouse reactor vessels that support and orient the reactor fuel assemblies and direct coolant flow through the core. The core baffle, one of these internal structures, is a set of vertical plates surrounding the outer rim of the reactor's fuel assemblies. The baffle provides lateral restraint to the core and directs coolant flow through the core. The vertical baffle plates are bolted to the edges of horizontal former plates, which are bolted to the inside surface of the core barrel. There are typically eight levels of former plates located at various elevations within the core barrel. The baffle-former bolts secure the baffle plates to the former plates. To cool the baffle structure, some water flowing through the reactor vessel is directed between the core barrel and the baffle plates in either a downward direction ("down-flow"), or an upward direction ("up-flow"). "Down-flow" plants place more stress on baffle-former bolts, which contributes to susceptibility of the bolts to degradation. Plants with the modified "up-flow" direction have shown little baffle-former bolt cracking as compared to the "down-flow" designs. Newer PWRs use the "up-flow" configuration and several older units have converted to the "up-flow" configuration.

Figures below: Baffle-former assembly bolts (left). This shows the three styles of baffle plate bolts: edge bolts, baffle-former bolts, and corner edge bolts. Baffle-former bolts are the bolts that have experienced the recent issues at Indian Point-2 and Salem-1. A typical core barrel baffle arrangement is shown (right).

Baffle-Former Bolts Baffle-Former Bolts
Typical baffle-former bolts are made of stainless steel (Type 347, Type 316, or Type 304), are approximately 5/8ths of an inch in diameter, and are typically 1.5 to 2 inches long. Units that have replaced baffle-former bolts have done so in accordance with a Westinghouse-approved design that uses bolts with less susceptible material properties and an improved head geometry to reduce stress concentrations. Bolts made from Type 316 stainless steel are less susceptible to degradation than those made from Type 347 stainless steel. Most baffle-former bolt designs secure the bolt heads with a welded lock tab (see photo below, right). These lock tabs normally retain the bolt heads should they become detached.
Baffle Bolts Baffle Bolt Lock Tab
Baffle bolts (left); typical baffle bolt lock tab (right)

Baffle-former bolt degradation was first noted back in the late 1980's in foreign plants, and has been periodically reported at U.S. plants. The U.S. Nuclear Regulatory Commission (NRC) issued Information Notice No. 98-11, "Cracking of Reactor Vessel Internal Baffle-Former Bolts in Foreign Plants," describing the foreign events.

Baffle-former bolts are subjected to significant mechanical stress and high levels of neutrons coming from the core for many years. Over time these conditions lead to degradation of the bolts, in the form of irradiation-assisted stress corrosion cracking. Significant cracking can reduce the load that a baffle-former bolt is able to support and eventually result in the detachment of the bolt head. Irradiation-assisted stress corrosion cracking is a known phenomenon and the inspection and management of this degradation mechanism is the subject of an NRC-approved Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) topical report, MRP-227-A, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines."

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Safety/Risk Implications

A bolt head or bolt lock tab may detach if the degradation of the baffle-former bolt is significant. During normal operation, the separation of a bolt head or a bolt lock tab can introduce loose parts or foreign material to the reactor coolant system. The loose parts or foreign material could impact fuel assemblies and could potentially lead to fuel leaks. In this case, the condition will be detected by routine monitoring of radioactivity in the reactor coolant system.

If a significant number of baffle-former bolts in a concentrated area are damaged or missing, then the additional stresses associated with a loss of coolant accident (LOCA) or seismic event could displace baffle plates. LOCA stresses are more likely to displace the baffle plates than seismic stresses. In the worst-case scenario of a large break LOCA, the potential exists for a baffle plate to detach or deform. This would likely only affect the fuel assemblies next to the baffle plates and would not meet the definition of "core damage." A detached or deformed plate is unlikely to challenge the ability to cool the fuel assemblies. It is also unlikely to challenge the ability to safely shut down the reactor because most units do not have control rods located in fuel assemblies near the edge of the core. Many factors reduce the potential safety implications of the worst-case scenario, including the fact that most baffle-former bolts that fail inspection are only partially cracked and can still bear some load. In addition, baffle plate edge bolts remain intact and provide significant restraint against plate detachment but this factor is not normally credited in the minimum bolting analyses.

Reactor designs other than Westinghouse have not exhibited issues with baffle-former bolt degradation. All but two Combustion Engineering-designed PWRs employ a welded, rather than bolted, baffle assembly (or "core shroud"). The two Combustion Engineering units with bolted core shrouds use Type 316 stainless steel bolts and have observed no significant issues. Babcock & Wilcox-designed PWRs generally use Type 304 stainless steel baffle-former bolting and also have not observed any significant issues. General Electric-designed boiling water reactors are unaffected by this issue because they use a different design involving a welded core shroud assembly.

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Recent Issues

Indian Point Unit 2 – During its 2016 spring refueling outage, while performing inspections adhering to the standards of MRP-227-A, the licensee discovered baffle-former bolt degradation issues.  During this review it was determined that 227 of 832 baffle-former bolts inspected at the plant were potentially degraded (these bolts either had ultrasonic (UT) indication of cracking, visual defects such as a missing bolt head or cracked locking tab weld, or were inaccessible for UT).  This issue was reported under EN 51829, "BAFFLE BOLT INDICATIONS IDENTIFIED DURING INSERVICE INSPECTION" on March 29, 2016.  The licensee (Entergy) replaced the potentially degraded bolts along with an additional 51 bolts to provide further margin for baffle plate structural integrity and to allow for testing of bolts that are considered to be non-degraded. 

Indian Point Unit 2 returned to service in late June.  Prior to the unit's restart, the NRC conducted an independent evaluation of the analysis done for Entergy on how many new bolts had to be installed to maintain safety margins and ensure the structural integrity of the baffle-former plates.  The agency also had specialist inspectors at the plant for first-hand observations and information-gathering on bolt-removal and -replacement activities.  Based on those reviews, the NRC concluded the reactor was safe to operate until at least its next refueling outage.  Refueling outages take place once every 18 to 24 months and the bolts will be subject to further inspections.

An NRC Blog posting titled: "An Outage Twist: Degraded bolts at New York Nuclear Plant Warrant Attention", was issued on the NRC public blog gateway on April 27, 2016.

Salem Unit 1 – During its refueling outage in spring 2016, the licensee (Public Service Enterprise Group (PSEG)) performed reactor vessel visual inspections and discovered degraded baffle-former bolts.  Numerous bolts with cracking or indications were noted based on visual examinations and subsequent UT inspections.  This included 18 bolt heads that were detached or protruding.  All of the detached bolt heads were subsequently recovered.  PSEG reported this condition in EN 51902,"ANOMALIES IDENTIFIED DURING VISUAL INSPECTION OF REACTOR VESSEL INTERNALS," on May 3, 2016.  On July 5, 2016, PSEG submitted Licensee Event Report (LER) 16-001-00 for Salem Unit 1 titled, "Regarding Unanalyzed Condition due to Degraded Reactor Baffle to Former Bolts."  A total of 189 potentially degraded bolts were replaced.

Salem Unit 1 was returned to service on July 30, 2016. Prior to the unit's restart, the NRC conducted an independent evaluation of the analysis done for PSEG on how many new bolts had to be installed to maintain safety margins and ensure the structural integrity of the baffle-former plates.  The NRC also reviewed the licensee's analysis of potential safety implications for Salem Unit 2 and determined the reactor remains safe.  In addition, the agency had specialist inspectors at the plant for first-hand observations and information-gathering on bolt removal and replacement activities.  Based on those reviews, the NRC concluded the reactor was safe to operate until at least its next refueling outage.  Refueling outages take place once every 18 to 24 months and the bolts will be subject to further inspections.

D.C. Cook Unit 2 – During the October refueling outage, D.C. Cook Unit 2 performed visual and UT examinations of all baffle-former bolts.  The UT examinations were performed earlier than originally scheduled in accordance with MRP interim guidance. The inspection found 181 baffle-former bolts to be potentially degraded.  Of these, six were replacement bolts installed in 2010.  These six bolts were sent for laboratory testing to further investigate the extent of degradation.  Indiana Michigan Power Co., the operator of D.C. Cook, replaced all the potentially degraded bolts and some additional bolts for a total of 201 bolts replaced, and performed analyses to support Unit 2's return to service and the continued operation of Unit 1.

Indian Point Unit 3 – During the spring of 2017, the licensee for Indian Point Unit 3 performed UT examination of 100 percent of the baffle-former bolts, finding 259 (31 percent) potentially degraded, including 256 that had UT indications and 3 untestable bolts.   The licensee observed no visual indications of bolt failure, such as loose or protruding bolt heads or missing or broken locking bars.  In addition, due to the lack of visually failed bolts, and the fact that a relatively small percentage of bolts broke during removal, the overall structural integrity of the baffle assembly at Indian Point Unit 3 prior to bolt replacement may have been comparable or better than that of the first three plants to find significant degradation.   The licensee replaced a total of 270 bolts, including all potentially degraded bolts and 11 additional non-degraded bolts.

Salem Unit 2 – The licensee (PSEG) will complete the initial UT inspection of 100 percent of the baffle-former bolts during their Spring 2017 outage.

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Response and Next Steps

The NRC evaluated the degraded baffle-former bolts issue for potential reduction in safety margin using NRC guidance LIC-504, "Integrated Risk-Informed Decision-Making Process for Emergent Issues."  The LIC-504 process guides the decision on whether immediate regulatory action, such as ordering a plant to shutdown, is required.  This initial assessment is based on an overall risk assessment of the issue.  In this case, the NRC determined that this situation does not warrant the immediate shutdown of any plant because it did not result in an unacceptable increase in the core damage frequency.  The LIC-504 evaluation was completed on October 20, 2016.

Westinghouse performed an acceptable bolting pattern analysis for each unit with potentially degraded baffle-former bolts to determine the minimum required number of baffle-former bolts and to determine which degraded bolts required replacement before restart of these units.  In general, all licensees have replaced all potentially degraded bolts and any bolts that were inaccessible for UT examination.  The NRC staff will assess and review the Westinghouse analysis and root cause information provided by each licensee, including the results of metallurgical analysis, in order to develop an appropriate longer-term regulatory response.  In addition to inspecting the repairs and analyses of the two units where degradation was observed, the NRC staff is inspecting the licensees' evaluation supporting continued operation of the other units at those sites.

Plants with similar designs are expected to enter this relevant operating experience into their corrective action programs.  Part of that process is to assess the issue for impact on the operability of the affected units.  The NRC's resident inspectors will ensure the licensees have taken these actions and will evaluate each licensee's response.  Although Diablo Canyon Unit 2 is a Westinghouse four-loop PWR, it was previously converted to the "up-flow" configuration and has Type 316 stainless steel bolting.  A 100 percent visual inspection of the Diablo Canyon Unit 2 baffle-former bolts was performed in May 2016, and no indications of degradation were identified.

Westinghouse issued a Nuclear Safety Advisory Letter (NSAL) 16-1 on July 5, 2016. This document was addressed to plant owners and discussed Westinghouse's basis for concluding that the issue does not create a substantial safety hazard.  NSAL 16-1 provides a breakdown of plants into tiers based on susceptibility to baffle-former bolt degradation and recommends follow-up inspection actions.  For plants with designs and configurations similar to Indian Point Unit 2 and Salem Unit 1 (Tier 1a), the NSAL recommends conducting UT examinations of the baffle-former bolts at the next scheduled refueling outage.  NRC staff met with EPRI/MRP staff on July 19, 2016, and Westinghouse presented an overview of the information in NSAL 16-1 at this meeting.  On November 16, 2016, the NRC and industry representatives briefed the Advisory Committee on Reactor Safeguards Metallurgy and Fuels Subcommitee on the recent baffle-former bolt degradation.  See the transcript of the meeting for more detail, including the presentations by the NRC and industry.

The MRP convened a special focus group on May 16, 2016, to support an integrated approach among industry organizations and address the recent operating experience with baffle-former bolts.  The baffle-former bolt focus group has issued two letters transmitting NEI 03-08 with "needed" interim guidance for all Westinghouse-design plants.  The first letter, dated July 27, 2016, covers Tier 1 plants, while the second letter dated March 15, 2017, covers all Westinghouse plants.  The interim guidance calls for accelerated initial inspection schedules for downflow plants (Tier 1 and Tier 2) and places limits on the maximum time to re-inspect the baffle-former bolts following the initial inspection, depending on the findings of the initial inspection.   For Tier 1a plants, the interim guidance is consistent with NSAL 16-1 in calling for initial UT inspection at the plant's next refueling outage.  The NRC staff is reviewing this guidance and will document its position on the guidance publically when complete.  The interim guidance letters modified the guidance for baffle-former bolt inspection in MRP-227-A and MRP-227, Rev. 1.  The NRC staff will document its position on the interim guidance in its safety evaluation of MRP-227, Rev. 1, which is currently under review. The focus group will continue working to establish an improved fundamental understanding of the degradation mechanisms and develop potential changes to the MRP-227 inspection guidance as needed.

The NRC anticipates issuing a generic communication on the issue once more root-cause information becomes available.  The NRC staff has determined that there is no immediate safety concern, and that this issue does not warrant the immediate shutdown of any plant.  The NRC will ensure the condition is suitably understood and addressed and that appropriate regulatory actions are taken.

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Page Last Reviewed/Updated Friday, August 11, 2017