Information Notice 2012-15, Use of Enclosures to Mitigate Leakage from Joints That Use A-286 Bolts

Use of Enclosures to Mitigate Leakage from Joints That Use A-286 Bolts

ML121740012
Person / Time
Issue date:	08/09/2012
From:	Laura Dudes, Mark Lombard, Mcginty T Office of Nuclear Reactor Regulation, Office of Nuclear Material Safety and Safeguards, Office of New Reactors, Division of Policy and Rulemaking
To:
Purnell, B A, NRR/DPR, 415-1380
References
IN-12-015
Download: ML121740012 (7)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS

WASHINGTON, DC 20555-0001 August 9, 2012 NRC INFORMATION NOTICE 2012-15: USE OF SEAL CAP ENCLOSURES TO MITIGATE

LEAKAGE FROM JOINTS THAT USE A-286 BOLTS

ADDRESSEES

All holders of an operating license or construction permit for a nuclear power plant issued under

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of

Production and Utilization Facilities, except those who have permanently ceased operations

and have certified that fuel has been permanently removed from the reactor vessel.

All holders of combined licenses issued under 10 CFR Part 52, Licenses, Certifications, and

Approvals for Nuclear Power Plants.

All holders of and applicants for an independent spent fuel storage installation license under 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste.

PURPOSE

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform

addressees of potential issues associated with the installation of seal cap enclosures

(enclosures) to mitigate leakage from A-286 bolted connections in nuclear power plant piping.

A-286 is a precipitation-hardened, iron-based super alloy specified as American Society for

Testing and Materials (ASTM) A453, Grade 660 material. The NRC expects recipients to

review the information in this IN for applicability to their facilities and consider taking action, as

appropriate. Suggestions contained in this IN are not NRC requirements; therefore, no specific

action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

Callaway Plant, Unit 1

Between 1985 and 1987, Callaway Plant, Unit 1, installed enclosures on four swing check

valves located on the chemical and volume control system charging header to mitigate gasket

leakage from the bolted body-to-bonnet flange joint. The enclosures were not part of the

pressure boundary. Each valve was constructed from austenitic stainless steel bodies and

bonnets, which were connected using bolting composed of A-286. In the unified alloy

numbering system, it is designated as UNS S66286.

In 1992, all four enclosures were removed for valve and bolting examination and maintenance.

A redesigned enclosure was installed on each valve at that time. During the 2002 refueling

outage, one of the valves fitted with the redesigned enclosure exhibited evidence of boric acid

leakage from the upper enclosure weld. The enclosure was permanently removed and the

valve bonnet and bolting were replaced, but no bolting issues were identified. During the 2004 refueling outage, a second valve fitted with the redesigned enclosure exhibited boric acid

leakage. The licensee removed the enclosure to perform maintenance on the valve. During

disassembly, 3 of the 12 valve bonnet closure studs failed with the application of negligible

force. A metallurgical analysis concluded that the studs most likely failed because of

intergranular stress corrosion cracking (SCC). This valve was repaired and the enclosure was

not reinstalled. Although inspections found no evidence of leakage or bolting degradation on

the two remaining swing check valves, the licensee removed the enclosures and replaced the

bolting on these valves.

South Texas Project, Unit 2

In February 1997, South Texas Project Nuclear Operating Company (STPNOC) discovered

steam wisping from the body-to-bonnet gasket of a swing check valve located in the 8-inch

nominal diameter safety injection piping at South Texas Project, Unit 2 (STP-2). This valve is

constructed from an austenitic stainless steel body (SA182, Type 316) and bonnet (SA240,

Type 316) joined using type A-286 bolts. As a corrective action, the licensee welded a metal

enclosure over the bonnet and the studs of the check valve to mitigate the leakage (Figure 1).

This enclosure is designed to contain leakage only; it is welded to, but not part of, the reactor

coolant system pressure boundary.

During a January 1999 outage, the licensee identified boric acid deposits on the insulation for

the check valve. The licensee took no action at this time because it did not observe active

leakage. In March 2001, the licensee cleaned the check valve and performed a liquid penetrant

examination on the welds, joining the enclosure and valve body based on the findings in

January 1999. No liquid penetrant indications were noted. In April 2010, the licensee observed

a 6-inch steam plume emitting from the check valve during plant startup. The licensee

inspected the seal cap welds and did not observe weld defects or boric acid deposits. The

licensee concluded that condensation in the enclosure bowl on top of the bonnet caused the

steam plume, so it took no action at that time.

During the October 2011 refueling outage, the licensee identified water and boric acid crystals

on the outside surface of the enclosure and on the valve bonnet (Figure 2). The licensee

performed liquid penetrant testing and identified flaws on the weld joining the enclosure to the

bonnet. Subsequently, the licensee repaired the fillet weld.

In April 2012, the licensee again found boric acid crystals in the weld joining the enclosure to the

bonnet. Subsequently, the licensee removed the enclosure on the check valve and performed

ultrasonic and visual examinations of the bolts. It did not find any degradation on the bolts. The

enclosure was subsequently reinstalled. Permanent repair of the valve and removal of the

enclosure are currently scheduled for the next refueling outage planned in 2013.

BACKGROUND

Alloy A-286 is procured to meet the requirements of ASTM A453 (American Society of

Mechanical Engineers (ASME) SA453), Grade 660, or ASTM A638 (ASME SA638), Grade 660.

This alloy has been used in a variety of nuclear applications including reactor vessel internals bolting, control rod drive mechanism parts, reactor coolant pump shafts and bolting and other

applications. In many applications alloy A-286 has performed satisfactorily. The resistance of

this alloy to general corrosion is similar to that of 300 series stainless steels. It is not

susceptible to boric acid corrosion. There have been several instances, however, in which this

alloy has failed in service because of SCC.

NRC Information Notice (IN) 90-68, Stress Corrosion Cracking of Reactor Coolant Pump Bolts, dated October 30, 1990, discusses service failures of A-286 bolting that attached reactor

coolant pump-turning vanes to the pump shaft in an international nuclear plant and SCC of A-

286 bolting in reactor vessel internal components at four different Babcock and Wilcox designed

reactors. IN 90-68, Supplement 1, dated April 14, 1994, describes SCC failures of A-286 reactor coolant pump bolting at a Westinghouse designed reactor.

Additional service failures of A-286 bolting caused by SCC include cracking of top guide bolts

discovered at an international nuclear plant in 1982 and, later, at ABB-Atom boiling-water

reactors (BWRs). The plants used the bolts to attach guide bars to the top guide or core grid, which aligns the top end of the fuel assemblies. The bolts that failed were highly loaded;

however, lower stressed components (less than 30 percent of yield strength) made of alloy A-

286 did not experience cracking.

Laboratory studies have shown the susceptibility of A-286 to SCC in reactor coolant

environments. In general, susceptibility increases with applied loading and with dissolved

oxygen content in the environment. For high-purity, low-oxygen environments similar to

pressurized-water reactors (PWRs), A-286 may not be susceptible to SCC unless loaded above

the yield strength. For high-purity reactor-coolant environments that have higher oxygen

content typical of BWR coolant chemistries, susceptibility has been established at loading levels

of 60 percent of the yield strength.

NRC report, NUREG-6923, Expert Panel Report on Proactive Materials Degradation

Assessment, published in March 2007, notes that the role of impurities, including oxygen

introduced during plant shutdown and possibly consumed only slowly in confined crevices, in

helping crack initiation is clear from all the evidence available. Once initiated, cracks grow

relatively easily even in well-controlled pressurized-water reactor (PWR) primary water.

Service experience with A-286 bolting that is not wetted (i.e., external to the reactor coolant

system) has been good, with no reported failures caused by SCC.

DISCUSSION

The environment inside of an enclosure that is installed on a leaking flange is not necessarily

similar to the high-purity, low-oxygen environment inside a PWR reactor coolant system. When

the enclosure is installed, it is full of air, but if the joint is leaking, the enclosure can slowly fill

with leaking reactor coolant. The leaking reactor coolant initially will be a reducing environment, but the oxygen in the trapped air will dissolve and saturate the borated water. The environment

inside the enclosure will consist of hot, oxygen-saturated water, which will be much more

oxidizing than PWR normal coolant chemistry. Since there is no mechanism for exchanging the

water in the enclosure, the enclosure is similar to a dead leg connected to the reactor coolant

system through a tortuous leak path. The water in the enclosure will remain oxygen saturated

until all of the oxygen is consumed by electrochemical reactions with the metal surfaces in the

enclosure. Electrochemical reactions that cause SCC are likely to occur with the A-286 bolting

and enclosure attachment welds. As stated previously, laboratory studies have investigated the susceptibility of A-286 to SCC in

reactor-coolant environments. The data from these studies indicate that cracking increases with

increased oxygen content. Since the enclosure environments are likely to be higher in oxygen

than typical reactor-coolant environments, it is likely that cracking of A-286 materials in an

enclosure environment will be more severe than that identified in these studies. Because of the

increased oxygen content in the enclosures, it is unclear that mitigating factors identified in

laboratory testing, such as reducing tensile stresses, will preclude cracking in enclosure

environments. The best and potentially only information currently available concerning cracking

of A-286 material in enclosure environments is from Callaway, in which three studs failed, and

from STP-2, in which no cracking was observed.

In addition to the A-286 bolting, the enclosure attachment welds exhibited multiple cracks at

both Callaway and STP-2. The enclosures at Callaway and STP-2 exhibited leakage of reactor

coolant through the 300-series austenitic stainless steel enclosure attachment welds. STP-2 performed a penetrant test and identified indications in the enclosure attachment weld and

concluded that the leakage was caused by fabrication defects such as a slag inclusion or

porosity. However, penetrant testing is not a sufficient technique, by itself, to identify the

metallurgical nature of indications. The enclosure attachment welds at STP-2 were penetrant

tested at least twice before the recent findings in 2011, once during original installation and

once in March 2001. Previous examinations identified no weld defects or indications. STPNOC

did not perform a metallurgical evaluation of the defects in the attachment welds.

Austenitic stainless steels, in general, are not susceptible to SCC in a PWR coolant

environment, but are susceptible to SCC in hot oxygenated water. The enclosures at Callaway

and STP-2 contained hot reactor coolant that was in contact with the oxygen-containing

atmosphere trapped within the enclosure. It is possible that the leakage through the enclosure

attachment welds at Callaway and STP-2 resulted from SCC that was caused by the pressure

stresses and exposure of the attachment welds to the aggressive, hot, oxygenated environment

inside the enclosure.

Detection of failures of the welds joining the valve to enclosure may be possible by detection of

leaks. However, in both instances above, failures were detected by insulation removal and

identification of boric acid deposits. No inspection techniques are currently identified to permit

detection of bolt failure without first removing the enclosure. Failure to identify bolting failures

could result in a loss-of-coolant accident.

In summary, this IN alerts licensees that failures of A-286 bolting and the enclosure-to-valve

welds may occur because of the unique environment which exists within valve enclosures.

Bolting failures may challenge the structural integrity of the primary system pressure boundary

and may result in a loss-of-coolant accident. Because of the differences between the

environment within enclosures and the environments in which laboratory testing was conducted, it is unclear if mitigating techniques, such as reducing bolt tensile stresses by reducing the

torque on bolts, will prevent crack formation. Additionally, at the present time, inspection of

bolting requires removal of the enclosure.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contact listed below or the appropriate Office of Nuclear Reactor

Regulation (NRR) and Office of New Reactors (NRO) project managers.

/RA/ /RA by JLuehman for/

Timothy J. McGinty, Director Laura A. Dudes, Director

Division of Policy and Rulemaking Division of Construction Inspection and

Office of Nuclear Reactor Regulation Operational Programs

Office of New Reactors

/RA/

Mark D. Lombard, Director

Division of Spent Fuel Storage

and Transportation

Technical Contact:

John C. Tsao, NRR James F. Drake, RIV

301-415-2702 817-200-1558 E-mail: John.Tsao@nrc.gov E-mail: James.Drake@nrc.gov

Robert O. Hardies, NRR

301-415-5802 E-mail: Robert.Hardies@nrc.gov

Enclosure: Figures

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library, Document Collections.

CONTACT

This IN requires no specific action or written response. Please direct any questions about this

matter to the technical contact listed below or the appropriate Office of Nuclear Reactor

Regulation (NRR) and Office of New Reactors (NRO) project managers.

/RA/ /RA by JLuehman for/

Timothy J. McGinty, Director Laura A. Dudes, Director

Division of Policy and Rulemaking Division of Construction Inspection and

Office of Nuclear Reactor Regulation Operational Programs

Office of New Reactors

/RA/

Mark D. Lombard, Director

Division of Spent Fuel Storage

and Transportation

Technical Contact:

John C. Tsao, NRR James F. Drake, RIV

301-415-2702 817-200-1558 E-mail: John.Tsao@nrc.gov E-mail: James.Drake@nrc.gov

Robert O. Hardies, NRR

301-415-5802 E-mail: Robert.Hardies@nrc.gov

Note: NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library, Document Collections.

ADAMS Accession Number: ML121740012 *via e-mail TAC No. ME8876 OFFICE DE:EPNB DE:SLS RIV:DRS:PSB2* Tech Editor* BC:DE:EPNB

NAME JTsao RHardies JDrake CHsu TLupold

DATE 7/11/2012 7/11/2012 7/12/2012 6/26/2012 7/16/2012 OFFICE D:NRR:DE BC:NMSS:SFST:SB* BC:DLR:RAPB BC:IP:ICAB PM:NRR:PGCB

NAME PHiland DPstrak RAuluck CAbrams BPurnell

DATE 7/18/2012 7/25/2012 7/20/2012 7/23/2012 07/31/2012 OFFICE BC:NRR:PGCB LA:NRR:PGCB D:NMSS:DSFST D:NRO:DCIP D:NRR:DPR

NAME DPelton CHawes MLombard LDudes TMcGinty

DATE 08/01/2012 08/01/2012 8/8/2012 8/2/2012 8/9/2012 OFFICIAL RECORD COPY

IN 2012-15 Enclosure

Figure 1: Diagram of the enclosure

Figure 2: Photo of the enclosure