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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 ultrasonic (UT) 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 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 baffle-former bolts of all seven of these reactors have been inspected and bolts have been replaced to restore structural integrity.
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. The industry through the Electric Power Research Institute (EPRI) has issued updated inspection guidance for baffle-former bolts which the NRC found acceptable, as documented in a staff assessment. Licensees of PWRs have been performing the inspections as recommended by the updated guidance, and performing corrective actions as needed, including replacement of degraded bolts.
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.
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."
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.
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. In addition to inspecting the repairs and analyses of the units where degradation was observed, the NRC staff inspected the licensees' evaluations supporting continued operation of the other units at those sites.
Baffle-former bolts removed from reactors, both degraded and non-degraded, have been sent for laboratory testing to confirm the cause of the cracking, verify accuracy of the inspections, and determine the remaining strength of partially cracked bolts. These results show that the bolts cracked due to a combination of irradiation-assisted stress corrosion cracking and fatigue, consistent with the NRC staff's previous understanding of the issue. The results also showed many bolts which the ultrasonic inspection identified as partially cracked retain a significant amount of strength.
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 inspections of the baffle-former bolts at the next scheduled refueling outage. Tier 1b plants are similar in design to Tier 1a, except that the baffle-formed bolt material is Type 316 stainless steel. For Tier 1b, the NSAL recommended visual inspection of the bolts at the next refueling outage followed by UT inspection the following refueling outage.
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 1a and Tier 1b plants, while the second letter dated March 15, 2017, covers all Westinghouse plants. The interim guidance calls for accelerated initial inspection schedules for "down-flow" 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 interim guidance does not change the initial inspection schedules for "up-flow" plants. The NRC staff reviewed this guidance and documented its position on the guidance in a staff assessment. The interim guidance letters modified the guidance for baffle-former bolt inspection in MRP-227-A and MRP-227, Rev. 1. EPRI also agreed to incorporate its updated guidance on baffle-former bolt inspections into the final version of MRP-227, Revision 1, which is currently under review by the NRC staff. Operating experience from implementing the updated guidance is detailed below.
At a public meeting with industry (EPRI, Westinghouse, the Pressurized Water Reactor Owner's Group, and licensees) on July 19, 2016, the NRC and industry discussed industry activities related to baffle-former bolts, including an overview of NSAL 16-1 On November 16, 2016, the NRC and industry representatives briefed the Advisory Committee on Reactor Safeguards Metallurgy and Fuels Subcommittee 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 NRC held another public meeting with the industry on baffle-former bolt issues on July 13, 2017.
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 staff will continue to monitor the examination, repair and evaluation of baffle-former bolts through the NRC inspection process.
Section 104 of the Nuclear Energy Innovation and Modernization Act, required the NRC to submit to the appropriate congressional committees a report explaining revisions made to guidance on the baseline examination schedule and subsequent examination frequency for baffle-former bolts in pressurized-water reactors (PWRs) with down-flow configurations or a report explaining why current guidance is sufficient. The NRC provided the report to Congress on April 9, 2019, explaining why its considers the current guidance is sufficient.
The report concluded that further revision to baffle-former bolt guidance is not necessary, because the NRC staff has reviewed EPRI's guidance related to this issue, has found it acceptable, and has verified that licensees of susceptible reactors are properly implementing this guidance. The report also noted that initial baffle-former bolt examinations recommended by EPRI's updated guidance have been completed for all reactors in the two groups most susceptible to baffle-former bolt cracking, and that these examinations indicate that corrective actions appear to be effective.
Details of baffle-former bolt inspections and repairs can be found in an EPRI presentation. The NRC also gave an update at a May 2019 public meeting on its recent activities related to baffle-former bolts. As of June 1, 2019, initial ultrasonic examinations have been completed for at least 19 reactors, with follow-up examinations also performed at four reactors in the most susceptible group.
Inspections of Most Susceptible Group of Reactors
Westinghouse four-loop reactors with Type 347 stainless steel baffle-former bolts operating in a downflow configuration are the most likely reactors to have baffle-former bolt cracking. The updated EPRI guidance directed that the initial ultrasonic inspection of all of the baffle-former bolts in these plants be completed by the next refueling outage after the guidance was issued in July 2016. These inspections have now been completed at all seven of these reactors, and degraded bolts plus additional non-degraded bolts have been replaced.
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 UT indication of cracking, visual defects such as a missing bolt head or cracked locking tab weld, or were inaccessible for UT inspection). 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, 2016. 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. The licensee performed a follow-up UT inspection of all the baffle-former bolts during the spring 2018 refueling outage, finding no damage to the replacement bolts and 13 additional degraded original bolts.
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.
During the spring 2019 refueling outage, the licensee observed 31 original baffle-former bolts with visual indications of degradation. The licensee therefore performed an ultrasonic examination of all the original baffle-former bolts and a sample of the bolts replaced in 2016. The licensee identified a total of 228 potentially degraded baffle-former bolts, including the 31 bolts with visual indications. One baffle-former bolt identified by UT as degraded was a replacement bolt. The licensee plans to replace 272 baffle-former bolts, including additional non-degraded bolts to provide additional margin. The bolt replacement is expected to be complete in mid-June, 2019.
Salem, Unit 1 is the only plant in the most susceptible group to find a large number of degraded baffle-former bolts during a follow-up examination. The NRC staff is currently monitoring the licensee’s corrective actions and root cause analysis and will assess whether additional guidance is needed with respect to timing of the follow-up examination of baffle-former bolts.
D.C. Cook Unit 2 – During the October 2016 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. The licensee performed a follow-up UT inspection of all baffle-former bolts during the spring 2018 refueling outage, finding no degraded replacement bolts, and four degraded original bolts. The licensee also replaced an addition 210 bolts and converted the reactor to upflow to reduce the stresses on the bolts.
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.
Other Reactors in Susceptible Group -
Baffle-former bolt UT inspections were performed at the remaining three reactors in the most susceptible group (Diablo Canyon, Unit 1, D.C. Cook, Unit 1, and Salem, Unit 2) during 2017, finding smaller numbers of degraded bolts. The licensees for these reactors replaced all the degraded bolts plus preventively replaced additional non-degraded original bolts.
Inspections at Other Downflow Plants
Sequoyah, Units 1 and 2 are Westinghouse four-loop reactors operating in "down-flow," but with Type 316 stainless steel baffle-former bolts. These reactors are considered the second most susceptible group to baffle-former bolt cracking; however, the bolt material makes them slightly less susceptible to bolt cracking. Both reactors have now completed ultrasonic examinations of the baffle-former bolts, finding a minimal number of degraded bolts. In addition, the updated EPRI guidance directed Westinghouse two-loop and three-loop reactors operating in a downflow configuration to complete the initial ultrasonic inspection of the baffle-former bolts on an accelerated schedule. These inspections have been completed at most of the reactors. These plants have found lower levels of degraded bolts, ranging from zero to ten percent bolts being potentially degraded. These plants generally performed an analysis to show they could safely operate with the few degraded bolts.
Inspections at Up-flow Plants
Several plants that were modified to convert from "down-flow" to "up-flow" have performed ultrasonic inspections, finding low levels of degraded bolts. These include Westinghouse two-loop reactors Point Beach, Units 1 and 2, and North Anna, Unit 1, a Westinghouse three-loop reactor. Diablo Canyon Unit 2, a Westinghouse four-loop PWR, 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.
Page Last Reviewed/Updated Monday, November 02, 2020