- About NRC           - The Commission
      - Statutory Authority
      - Strategic Plan
      - Budget &  Performance
      - Organization & Functions
      - Locations
      - History
      - Values
      - Policymaking
      - Radiation Protection
      - Fire Protection
      - How We Regulate
      - Grants
      - Civil Rights
      -     
    - NuclearReactors           - Operating Reactor Locations
      - Reactor Quick Links
      - Power Reactors
      - Research and Test Reactors
      - Operating Reactors
      - Operator Licensing
      - New Reactors
      - Operator Licensing for New Reactors
      - Reactor Safety Research
      - Advisory Comm (ACRS)
      - Decommissioning of Nuclear Facilities
      -     
    - NuclearMaterials           - Fuel Cycle Facility Locations
      - Uranium Recovery Locations
      - Materials Quick Links
      - Special Nuclear Material
      - Source Material
      - Byproduct Material
      - Med, Academic & Ind Uses
      - Source Materials Facilities
      - Uranium Recovery
      - Fuel Cycle Facilities
      - Materials Transportation
      - Materials Research
      - Advisory Comm (ACMUI)
      - Decommissioning of Nuclear Facilities
      - Materials Environmental Reviews
      -     
    - RadioactiveWaste           - Waste Quick Links
      - Low-Level Waste
      - Waste Incidental to Reprocessing
      - Incidental Waste Criteria
      - High-Level Waste
      - Uranium Mill Tailings
      - LLW Disposal
      - HLW Disposal
      - Spent Fuel Storage
      - Spent Fuel Transport
      - Waste Research
      -     
    - NuclearSecurity           - Security Spotlight
      - NRC Responsibilities
      - Domestic Safeguards
      - Radioactive Material Security
      - Facility Protection
      - Nuclear Material Control
      - International Safeguards
      - Information Security
      - Emergency Preparedness
      - &;   
    - Public Meetings& Involvement           - Public Meeting Schedule
      - Commission	    Schedule
      - Conferences & Symposia
      - Documents for Comment
      - Brochures & Fact Sheets
      - Involving Stakeholders
      - Info Quality Guidelines
      - Rulemaking Process
      - Licensing
      - Enforcement
      - Hearing Process
      - Hearing Applications
      -     

**UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF NUCLEAR REACTOR REGULATIONWASHINGTON, D.C. 20555-0001**

March 25, 1998

**Addressees**

All holders of operating licenses for pressurized-water reactors (PWRs) except those who have permanently ceasedoperation and have certified that fuel has been permanently removed from the reactor vessel.

**Purpose**

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to the cracking ofreactor vessel internal baffle former bolts (see Figures 1 and 2) found at several foreign PWRs and to inform addressees ofactions taken and planned by domestic PWR owners groups in response to this experience. It is expected that the recipientswill review the information for applicability to their facilities and consider actions, as appropriate, to avoid similarproblems. However, suggestions contained in this information notice are not NRC requirements; IN 98-11

**Description of Circumstances**

Reactor vessel internals are structures located within the reactor vessel that support and orient the reactor fuel assembliesand direct coolant flow through the core. The core baffle is part of the internals structure, which consists of vertical platesthat surround the outer faces of the peripheral fuel assemblies. The baffle directs coolant flow through the core. The verticalplates are bolted to the edges of horizontal former plates that 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 bolts that secure the baffleplates to the former plates are referred to as "baffle former bolts."

European plants identified the cracking of baffle former bolts as early as 1988 and this problem continues to occur.Although this cracking is not fully understood, testing of cracked bolts suggests an age-related intergranularstress-corrosion cracking process influenced by bolt material, fluence, stress, and temperature. The reported crackingoccurred in 316 cold-worked stainless steel bolts. Most of the cracking reported has been in four French 900-MWe(megawatt electric) PWRs.

An investigation of the cracking was discussed in a paper contained in the Proceedings of the International SymposiumFontevraud III, dated September 12-16, 1994, held at the Royal Abbey of Fontevraud, France. The symposium paperreports that the cracking of baffle former bolts seems to be limited to the first six PWRs operated by Electricit� de France(EDF), which are all of the same design and are identified as the "CPO" series. Further, the paper reports that bolt crackinghas not been seen in the other French 900-MWe plants (the "CPY" series), or in the 1300-MWe plants. The paper notes thatthere are differences between the two series with regard to bolt design, bolt material, operating conditions, and reactorcoolant flow paths. Some plants in both groups have been in operation for approximately the same number of hours. Theplants which reported the greatest number of cracked bolts are Fessenheim Unit 2 and Bugey Unit 2, both of which areCPO series plants. The number of cracked bolts identified at these plants are 29 and 54, respectively. All of the baffleformer bolts (960) in each plant were tested ultrasonically.

**Discussion**

At the foreign plants, ultrasonic testing was performed to assess the integrity of the baffle former bolts. Five bolts wereremoved from the Bugey Unit 2 baffle assembly for a detailed investigation of the degradation process. One of the boltswas found to be broken, three were found to be cracked, and one was found to be sound. The conclusions reached in thesymposium paper are that (1) baffle bolt cracking has been limited to plants of the same design (CPO series), (2) boltcracking has occurred predominantly in zones of fluence and maximum temperature, (3) in the zones of maximum fluence,bolt cracking is found predominantly in the bolts under the highest mechanical stress, and (4) bolt cracking has occurred insome plants although not in other plants of the same design, a phenomenon that may be a consequence of varying boltmetallurgical properties and plant operating conditions.

The NRC is not aware of cracking of baffle former bolts in domestic PWRs. Domestic reactor baffle former bolts aresubject to the visual inservice inspection requirements of Section XI of the American Society of Mechanical EngineersBoiler and Pressure Vessel Code. However, the baffle bolt cracking reported in foreign PWRs has occurred at the junctureof the bolt head and the shank, which is not accessible for visual inspection.

Domestic PWR owners groups have met with the NRC staff to report on their current and planned activities regarding thepotential for baffle bolt cracking in domestic PWRs. The details of those meetings are discussed below.

The Westinghouse Owners Group (WOG) provided an assessment of the cracking of the baffle former bolts identified inforeign PWRs, including the potential impact of cracking on domestic Westinghouse plants, and provided information onits current and planned activities. The WOG stated that because of the large number of baffle former bolts in the baffleassembly, the failure of a few bolts should not have a significant safety impact. The WOG activities include (1)development of analytical methods and acceptance criteria for bolt analysis, (2) performance of risk-informed evaluations,(3) performance of analysis for three plant groupings (2-loop, 3-loop, and 4-loop) of what constitutes acceptable bolting,(4) continued participation in domestic and foreign related activities, (5) preparation of bid specifications for boltinspection equipment, and (6) preparations for bolt inspection and replacement. The WOG identified lead plant candidatesfor the 2-loop and 3-loop groups and a proposed inspection schedule for each group. The WOG indicated that the boltmaterial used in the 2-loop group is 347 stainless steel and the bolt material used in the 3-loop group is 316 cold-workedstainless steel.

The Babcock and Wilcox (B&W) Owners Group (B&WOG) provided information on its current and planned activities toaddress the potential for cracking of baffle former bolts in domestic B&W plants, including a presentation of its "PlantLicensing Reactor Vessel Internals Aging Management Program." The B&WOG provided a preliminary determination thatbolt cracking is not considered a significant safety issue for B&W plants. This determination is based upon knowledge ofthe baffle and bolting designs involved and is supported by conservative analyses that assume both normal-operating andmaximum accident-loading conditions. The B&WOG activities include (1) collection and evaluation of availableinspection and material data, (2) development and qualification of replacement bolt materials, and (3) preparation of apossible baffle bolt inspection on a lead plant during the next 10-year inservice inspection interval. The B&WOG indicatedthat the bolt material used in B&W plants is 304 stainless steel.

The Combustion Engineering (CE) Owners Group (CEOG) provided an assessment of the cracking of the baffle formerbolts reported in foreign PWRs, including the potential impact of the cracking on domestic CE plants. The CEOG believesthat the most likely mechanism for the cracking of cold-worked 316 stainless steel baffle former bolts in foreign plants isirradiation- assisted stress-corrosion cracking (IASCC). The CEOG indicated that only two of its plants use bolts to attachthe core shroud panels (i.e., the baffle plates) to the former plates. The CEOG believes that these two plants are lesssusceptible to IASCC because of several design differences: (1) the material used in these bolts is annealed 316 stainlesssteel, which is not cold worked; (2) the bolt stress from preload, as a percentage of yield strength, is much less than theEDF plants; (3) the differential pressure across the core shroud panels does not result in tensile loads on the panel (i.e., thebaffle) bolts during normal operation; and (4) the core shroud panel design allows for some flexing of the former platerelative to the core barrel, thus effectively reducing the load on the panel bolts.

This information notice requires no specific action or written response. If you have any questions about the information inthis notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear ReactorRegulation (NRR) project manager.

----