Information Notice 2017-03, Anchor/Darling Double Disc Gate Valve Wedge Pin and Stem-Disc Separation Failures: Difference between revisions

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{{#Wiki_filter:UNITED STATES
{{#Wiki_filter:ML17153A053 UNITED STATES


NUCLEAR REGULATORY COMMISSION
NUCLEAR REGULATORY COMMISSION
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OFFICE OF NEW REACTORS
OFFICE OF NEW REACTORS


WASHINGTON, DC 20555 June 15, 2017 NRC INFORMATION NOTICE 2017-03:               ANCHOR/DARLING DOUBLE DISC GATE VALVE
WASHINGTON, DC 20555  
 
June 15, 2017  
 
NRC INFORMATION NOTICE 2017-03:  
ANCHOR/DARLING DOUBLE DISC GATE VALVE


WEDGE PIN AND STEM-DISC SEPARATION
WEDGE PIN AND STEM-DISC SEPARATION
Line 43: Line 48:
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform
The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice (IN) to inform


addressees of operating experience regarding Anchor/Darling (a subsidiary of Flowserve)
addressees of operating experience regarding Anchor/Darling (a subsidiary of Flowserve)  
double disc gate valve (DDGV) failures. This IN provides a discussion of the recent LaSalle
double disc gate valve (DDGV) failures. This IN provides a discussion of the recent LaSalle


County Station Unit 2 Anchor/Darling DDGV failure, events at Browns Ferry that led to Part 21 reporting, and other operating experience that resulted in stem-disc separations. This
County Station Unit 2 Anchor/Darling DDGV failure, events at Browns Ferry that led to Part 21 reporting, and other operating experience that resulted in stem-disc separations. This


document contains information available to NRC staff as of May 2017. The NRC expects
document contains information available to NRC staff as of May 2017. The NRC expects


recipients of this IN to review the information for applicability to their facilities and consider
recipients of this IN to review the information for applicability to their facilities and consider


actions, as appropriate, to avoid similar problems. However, suggestions contained in this IN
actions, as appropriate, to avoid similar problems. However, suggestions contained in this IN


are not NRC requirements; therefore, no specific action or written response is required.
are not NRC requirements; therefore, no specific action or written response is required.


==DESCRIPTION OF CIRCUMSTANCES==
==DESCRIPTION OF CIRCUMSTANCES==
LaSalle County Station, Unit 2 


===LaSalle County Station, Unit 2===
On February 11, 2017, during a refueling outage at LaSalle County Station, Unit 2, the licensee
On February 11, 2017, during a refueling outage at LaSalle County Station, Unit 2, the licensee


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Initial analysis identified that a stem-disc separation occurred as a result of excessive wear of
Initial analysis identified that a stem-disc separation occurred as a result of excessive wear of


the valve stem threads and shear failure of the wedge pin. The licensee has not completed
the valve stem threads and shear failure of the wedge pin. The licensee has not completed


their root cause determination. The licensee reported this event in Licensee Event Report
their root cause determination. The licensee reported this event in Licensee Event Report


(LER) 2017-003-00, dated April 12, 2017 (Agencywide Documents Access and Management
(LER) 2017-003-00, dated April 12, 2017 (Agencywide Documents Access and Management
Line 75: Line 80:
The licensee had been using industry guidance to perform visual evaluations and diagnostic
The licensee had been using industry guidance to perform visual evaluations and diagnostic


testing on the valve. The guidance was based on earlier operating experience from an event on
testing on the valve. The guidance was based on earlier operating experience from an event on
 
October 20, 2012, at Browns Ferry Nuclear Plant, Unit 1. This event resulted in two reports


ML17153A053 under 10 CFR Part 21, Reporting of Defects and Noncompliance. The first Part 21 report was
October 20, 2012, at Browns Ferry Nuclear Plant, Unit 1.  This event resulted in two reports under 10 CFR Part 21, Reporting of Defects and Noncompliance. The first Part 21 report was


issued by the Tennessee Valley Authority (TVA), Anti-Rotation Pin Failure in Anchor Darling
issued by the Tennessee Valley Authority (TVA), Anti-Rotation Pin Failure in Anchor Darling
Line 89: Line 92:
Double-Disc Gate Valve at Browns Ferry Nuclear Plant Unit 1, dated February 25, 2013 (ADAMS Accession No. ML13064A012).
Double-Disc Gate Valve at Browns Ferry Nuclear Plant Unit 1, dated February 25, 2013 (ADAMS Accession No. ML13064A012).


===Browns Ferry Nuclear Plant, Units 1 and 2===
Browns Ferry Nuclear Plant, Units 1 and 2
 
On October 20, 2012, a Browns Ferry Nuclear Plant, Unit 1, high-pressure coolant injection
On October 20, 2012, a Browns Ferry Nuclear Plant, Unit 1, high-pressure coolant injection


(HPCI) steam isolation valve, which also serves as a containment inboard isolation valve, failed
(HPCI) steam isolation valve, which also serves as a containment inboard isolation valve, failed


its local leak rate test. Investigation revealed that, although the event was not a stem-disc
its local leak rate test. Investigation revealed that, although the event was not a stem-disc


separation, the wedge pin failed and one of the two disc retainers (see Figure 1) fell from the
separation, the wedge pin failed and one of the two disc retainers (see Figure 1) fell from the
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stem-to-upper wedge connection into a space between the valve discs, causing one of the two
stem-to-upper wedge connection into a space between the valve discs, causing one of the two


discs not to properly seat. The valve was a 10-inch Anchor/Darling DDGV. It was installed in
discs not to properly seat. The valve was a 10-inch Anchor/Darling DDGV. It was installed in


2007 and had not been disassembled since installation. The licensee, TVA, submitted a
2007 and had not been disassembled since installation. The licensee, TVA, submitted a


10 CFR Part 21 report, dated January 4, 2013 (ADAMS Accession No. ML13008A321), for this
10 CFR Part 21 report, dated January 4, 2013 (ADAMS Accession No. ML13008A321), for this


failure. In the report, TVA determined that the wedge pin failed because the vendor had not
failure. In the report, TVA determined that the wedge pin failed because the vendor had not


properly torqued the stem-to-upper wedge connection during manufacture.
properly torqued the stem-to-upper wedge connection during manufacture.
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Before the 2012 failure, Browns Ferry Nuclear Plant experienced two other wedge pin failures in
Before the 2012 failure, Browns Ferry Nuclear Plant experienced two other wedge pin failures in


10-inch Anchor/Darling DDGVs. The first wedge pin failure involved a Unit 2 HPCI outboard
10-inch Anchor/Darling DDGVs. The first wedge pin failure involved a Unit 2 HPCI outboard


steam isolation valve installed in 2001 that failed during testing that same year. The vendor
steam isolation valve installed in 2001 that failed during testing that same year. The vendor


determined that the stem-to-upper wedge connection was not properly torqued. The second
determined that the stem-to-upper wedge connection was not properly torqued. The second


wedge pin failure occurred in 2008 and involved a Unit 1 HPCI outboard steam isolation valve
wedge pin failure occurred in 2008 and involved a Unit 1 HPCI outboard steam isolation valve


installed in 2006 that failed during local leak rate testing. Internal inspection of the valve
installed in 2006 that failed during local leak rate testing. Internal inspection of the valve


revealed that the stem-to-upper wedge connection was not properly torqued. Figure 1 Typical Anchor/Darling DDGV (ADAMS Accession No. ML13064A012)
revealed that the stem-to-upper wedge connection was not properly torqued. Figure 1 Typical Anchor/Darling DDGV (ADAMS Accession No. ML13064A012)  
 
Surry Power Station, Unit 2


===Surry Power Station, Unit 2===
On February 2, 2011, Surry Power Station, Unit 2, tripped as a result of a low-flow condition in
On February 2, 2011, Surry Power Station, Unit 2, tripped as a result of a low-flow condition in


the reactor coolant system (RCS) C loop. The low-flow condition was the result of a stem-disc
the reactor coolant system (RCS) C loop. The low-flow condition was the result of a stem-disc


separation of the RCS C loop isolation valve. The valve was an Anchor/Darling 30-inch
separation of the RCS C loop isolation valve. The valve was an Anchor/Darling 30-inch


DDGV. Inspection of the valve internals revealed that the wedge pin failed and the upper
DDGV. Inspection of the valve internals revealed that the wedge pin failed and the upper


wedge threads exhibited excessive wear. The root cause was determined to be flow-induced
wedge threads exhibited excessive wear. The root cause was determined to be flow-induced


vibration coupled with inadequate torque of the stem-to-upper wedge connection, documented
vibration coupled with inadequate torque of the stem-to-upper wedge connection, documented


in LER 2011-001-00, dated April 1, 2011 (ADAMS Accession No. ML11105A032). A similar
in LER 2011-001-00, dated April 1, 2011 (ADAMS Accession No. ML11105A032). A similar


stem-disc separation occurred in 1999 on the RCS A loop isolation valve (also an
stem-disc separation occurred in 1999 on the RCS A loop isolation valve (also an


Anchor/Darling 30-inch DDGV). In the 1999 event, the wedge pin failed, allowing the stem to
Anchor/Darling 30-inch DDGV). In the 1999 event, the wedge pin failed, allowing the stem to


unthread from the upper wedge connection (see LER 1999-003-00, dated July 30, 1999 (ADAMS Legacy No. ML9908120152)).
unthread from the upper wedge connection (see LER 1999-003-00, dated July 30, 1999 (ADAMS Legacy No. ML9908120152)).


===River Bend Station, Unit 1===
River Bend Station, Unit 1  
 
On May 21, 2007, an unexplained drop in the reactor recirculation system loop A flow occurred
On May 21, 2007, an unexplained drop in the reactor recirculation system loop A flow occurred


at River Bend Station, Unit 1. Reactor power lowered to approximately 96.5-percent power with
at River Bend Station, Unit 1. Reactor power lowered to approximately 96.5-percent power with


no operator action. Operators determined that the most probable cause for the condition was
no operator action. Operators determined that the most probable cause for the condition was


that the loop A discharge isolation valve caused partial flow blockage. The valve was a
that the loop A discharge isolation valve caused partial flow blockage. The valve was a


20-inch Anchor/Darling DDGV. Further investigation during the plant shutdown revealed a
20-inch Anchor/Darling DDGV. Further investigation during the plant shutdown revealed a


stem-disc separation and severely worn stem and upper wedge threads. The wedge pin failed, with the two portions extending into the upper wedge still in place while the piece that
stem-disc separation and severely worn stem and upper wedge threads. The wedge pin failed, with the two portions extending into the upper wedge still in place while the piece that


transverses the shaft was missing. The licensee identified several contributing causes for the
transverses the shaft was missing. The licensee identified several contributing causes for the


valve failure, including inadequate torque of the stem-to-upper wedge connection during the previous valve assembly, and flow-induced vibration at the disc assembly caused by turbulent
valve failure, including inadequate torque of the stem-to-upper wedge connection during the previous valve assembly, and flow-induced vibration at the disc assembly caused by turbulent
Line 167: Line 173:
Wedge pin failures and stem-disc separation events associated with Anchor/Darling DDGVs
Wedge pin failures and stem-disc separation events associated with Anchor/Darling DDGVs


have occurred at both pressurized-water reactor and boiling-water reactor plants. As previously
have occurred at both pressurized-water reactor and boiling-water reactor plants. As previously


mentioned, the 2012 event at Browns Ferry Nuclear Plant, Unit 1 resulted in the issuance of
mentioned, the 2012 event at Browns Ferry Nuclear Plant, Unit 1 resulted in the issuance of
Line 175: Line 181:
(ADAMS Accession No. ML13064A012).
(ADAMS Accession No. ML13064A012).


In its Part 21 report, Flowserve concluded that at Browns Ferry Nuclear Plant, Unit 1:
In its Part 21 report, Flowserve concluded that at Browns Ferry Nuclear Plant, Unit 1:  
        failure was due to the shearing of the wedge pin which serves a joint locking


function at the threaded interface between the valve stem and upper wedge. The
failure was due to the shearing of the wedge pin which serves a joint locking
 
function at the threaded interface between the valve stem and upper wedge. The


pin is designed to ensure that the joint does not loosen due to vibration and other
pin is designed to ensure that the joint does not loosen due to vibration and other


secondary loads. On some valve designs, the pin also is used to attach the disc
secondary loads. On some valve designs, the pin also is used to attach the disc


retainers to the upper wedge. The pin shearing allowed rotation of the stem
retainers to the upper wedge. The pin shearing allowed rotation of the stem


during the closing stroke when the valve was seating and ultimately resulted in
during the closing stroke when the valve was seating and ultimately resulted in
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Flowserve has completed an evaluation of the failure and concluded the root
Flowserve has completed an evaluation of the failure and concluded the root


cause of the wedge pin failure was excessive load on the pin. The stem
cause of the wedge pin failure was excessive load on the pin. The stem


operating torque exceeded the torque to tighten the stem into the upper wedge
operating torque exceeded the torque to tighten the stem into the upper wedge


before installation of the wedge pin. The additional stem torque produced a load
before installation of the wedge pin. The additional stem torque produced a load


on the wedge pin creating a stress which exceeded the pin shear strength
on the wedge pin creating a stress which exceeded the pin shear strength


causing the failure. The recommended assembly stem torque did not envelope
causing the failure. The recommended assembly stem torque did not envelope


the operating torque for the TVA application providing the potential for an over
the operating torque for the TVA application providing the potential for an over


load situation and ultimate failure. The operating torque for the TVA valve was
load situation and ultimate failure. The operating torque for the TVA valve was


unusually high due to the fast closing time of the actuator and very conservative
unusually high due to the fast closing time of the actuator and very conservative
Line 216: Line 223:
larger, operated by an actuator that applies torque on the stem to produce the
larger, operated by an actuator that applies torque on the stem to produce the


required valve operating thrust. An operating stem torque greater than the
required valve operating thrust. An operating stem torque greater than the


assembly stem torque can provide the opportunity for excessive pin load and
assembly stem torque can provide the opportunity for excessive pin load and
Line 224: Line 231:
The stems on most double-disc (DD) gate valves larger than size 2" are attached
The stems on most double-disc (DD) gate valves larger than size 2" are attached


to the upper wedge using UN [unified constant pitch] threads. A pin is installed
to the upper wedge using UN [unified constant pitch] threads. A pin is installed


through the hub of the upper wedge and stem threaded section to prevent the
through the hub of the upper wedge and stem threaded section to prevent the
Line 230: Line 237:
stem from loosening and eventually unscrewing from the wedge. In addition, the
stem from loosening and eventually unscrewing from the wedge. In addition, the


disc retainers on some DD gate valves are attached using the wedge pin. See
disc retainers on some DD gate valves are attached using the wedge pin. See


Figure 1. The output torque of the actuator is transmitted to the stem/wedge joint
Figure 1. The output torque of the actuator is transmitted to the stem/wedge joint


through the stem and is resisted by the disc wedge pack, therefore the stem to
through the stem and is resisted by the disc wedge pack, therefore the stem to


wedge connection is loaded by the stem torque and thrust. The wedge pin is not
wedge connection is loaded by the stem torque and thrust. The wedge pin is not


designed to withstand the full actuator output torque. The actuator torque
designed to withstand the full actuator output torque. The actuator torque


direction tends to tighten the stem into the wedge during closing and tends to
direction tends to tighten the stem into the wedge during closing and tends to


loosen the stem during opening. Flowserve's Part 21 also states:
loosen the stem during opening. Flowserve's Part 21 also states:  
          Flowserve recommends that all critical Anchor/Darling Double-Disc Gate valves with
 
Flowserve recommends that all critical Anchor/Darling Double-Disc Gate valves with


threaded stem to upper wedge connections and actuators that produce a torque on
threaded stem to upper wedge connections and actuators that produce a torque on
Line 253: Line 261:
failure, events at Browns Ferry that led to Part 21 reporting, and other operating experience that
failure, events at Browns Ferry that led to Part 21 reporting, and other operating experience that


resulted in stem-disc separations. This document contains information available to NRC staff as
resulted in stem-disc separations. This document contains information available to NRC staff as


of May 2017. The licensee for LaSalle County Station, Unit 2 is in the process of completing
of May 2017. The licensee for LaSalle County Station, Unit 2 is in the process of completing


their root cause determination. Licensees can use this information in addition to the technical
their root cause determination. Licensees can use this information in addition to the technical


information provided in the 2013 Flowserve Part 21 report to consider actions, as appropriate, to
information provided in the 2013 Flowserve Part 21 report to consider actions, as appropriate, to
Line 264: Line 272:


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


matter to the technical contacts listed below.
matter to the technical contacts listed below.


/ra/                                                 /ra/
/ra/  
Louise Lund, Director                                 Timothy J. McGinty, Director
 
/ra/  
 
Louise Lund, Director
 
Timothy J. McGinty, Director
 
Division of Policy and Rulemaking
 
Division of Construction Inspection


Division of Policy and Rulemaking                    Division of Construction Inspection
Office of Nuclear Reactor Regulation


Office of Nuclear Reactor Regulation                  and Operational Programs
and Operational Programs


Office of New Reactors
Office of New Reactors


Technical Contacts: John W. Thompson, NRR/DIRS               Michael Farnan, NRR/DE
Technical Contacts: John W. Thompson, NRR/DIRS
 
Michael Farnan, NRR/DE
 
301-415-1011 
 
301-415-1486
 
E-mail:  John.Thompson@nrc.gov
 
E-mail:  Michael.Farnan@nrc.gov
 
Note:  NRC generic communications may be found on the NRC public Web site, http://www.nrc.gov, under NRC Library, Document Collections.
 
ML17153A053; *concurred via email                      TAC No. MF9698 OFFICE
 
NRR/DPR/PGCB/LA
 
TECH EDITOR
 
NRR/DIRS/IOEB/TL
 
NRR/DIRS/IOEB/TL
 
NRR/DIRS/IOEB/BC
 
NAME
 
ELee
 
JDougherty*
JThompson*
EThomas*
HChernoff*
DATE
 
06/02/17
06/05/17
06/12/17
06/12/17
06/12/17 OFFICE
 
NRR/DPR/PGCB/LA
 
RI/DRS/DD
 
RII/DRS/DD
 
RII/DRP/D
 
RIII/DRS/D
 
NAME
 
ELee
 
JYerokun*
MMiller*
JMunday*
KOBrien*
DATE
 
06/12/17
06/13/17
06/13/17
06/13/17
06/07/17 OFFICE
 
RIV/DRP/D
 
RIV/DRS/DD
 
NRR/DSS/D
 
NRR/DRA/DD
 
NRR/DE/DD
 
NAME
 
TPruett*
JClark*
MGavrilas*
SWeerakkody* (A)
KCoyne* (A)
DATE
 
06/12/17
06/12/17
06/12/17
6/13/17
06/13/17 OFFICE
 
NRR/DLR/D
 
NRR/DIRS/D
 
NRR/DPR/PGCB/PM
 
NRR/DPR/PGCB/BC NRO/DCIP/D
 
NAME


301-415-1011                          301-415-1486 E-mail: John.Thompson@nrc.gov          E-mail: Michael.Farnan@nrc.gov
GWilson*
CMiller*
NMartinez


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


ML17153A053; *concurred via email      TAC No. MF9698 OFFICE  NRR/DPR/PGCB/LA    TECH EDITOR          NRR/DIRS/IOEB/TL    NRR/DIRS/IOEB/TL  NRR/DIRS/IOEB/BC
DATE


NAME    ELee              JDougherty*          JThompson*          EThomas*          HChernoff*
06/13/17  
DATE    06/02/17           06/05/17             06/12/17           06/12/17           06/12/17 OFFICE NRR/DPR/PGCB/LA    RI/DRS/DD            RII/DRS/DD          RII/DRP/D          RIII/DRS/D
06/13/17  
06/14/17  
06/14/17  
06/14/17 OFFICE


NAME    ELee              JYerokun*            MMiller*            JMunday*            KOBrien*
NRR/DPR/D
DATE    06/12/17          06/13/17            06/13/17            06/13/17          06/07/17 OFFICE  RIV/DRP/D          RIV/DRS/DD          NRR/DSS/D           NRR/DRA/DD        NRR/DE/DD


NAME   TPruett*          JClark*              MGavrilas*          SWeerakkody* (A)    KCoyne* (A)
NAME
DATE    06/12/17          06/12/17            06/12/17            6/13/17            06/13/17 OFFICE  NRR/DLR/D          NRR/DIRS/D          NRR/DPR/PGCB/PM    NRR/DPR/PGCB/BC NRO/DCIP/D


NAME    GWilson*          CMiller*            NMartinez          AGarmoe (A)        TMcGinty
LLund


DATE   06/13/17          06/13/17            06/14/17            06/14/17          06/14/17 OFFICE  NRR/DPR/D
DATE


===NAME    LLund===
06/15/17}}
DATE    06/15/17}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Latest revision as of 20:23, 8 January 2025

Anchor/Darling Double Disc Gate Valve Wedge Pin and Stem-Disc Separation Failures
ML17153A053
Person / Time
Issue date: 06/15/2017
From: Louise Lund, Mcginty T
Division of Construction Inspection and Operational Programs, Division of Policy and Rulemaking
To:
Nancy Martinez
References
TAC No. MF9698 IN-17-003
Download: ML17153A053 (6)
v • d • e


ML17153A053 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

OFFICE OF NEW REACTORS

WASHINGTON, DC 20555

June 15, 2017

NRC INFORMATION NOTICE 2017-03:

ANCHOR/DARLING DOUBLE DISC GATE VALVE

WEDGE PIN AND STEM-DISC SEPARATION

FAILURES

ADDRESSEES

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

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

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

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

All holders of and applicants for a combined license under 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

PURPOSE

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

addressees of operating experience regarding Anchor/Darling (a subsidiary of Flowserve)

double disc gate valve (DDGV) failures. This IN provides a discussion of the recent LaSalle

County Station Unit 2 Anchor/Darling DDGV failure, events at Browns Ferry that led to Part 21 reporting, and other operating experience that resulted in stem-disc separations. This

document contains information available to NRC staff as of May 2017. The NRC expects

recipients of this IN to review the information for applicability to their facilities and consider

actions, as appropriate, to avoid similar problems. However, suggestions contained in this IN

are not NRC requirements; therefore, no specific action or written response is required.

DESCRIPTION OF CIRCUMSTANCES

LaSalle County Station, Unit 2

On February 11, 2017, during a refueling outage at LaSalle County Station, Unit 2, the licensee

was attempting to fill and vent the high-pressure core spray (HPCS) system when the Unit 2 HPCS injection isolation valve (an Anchor/Darling 12-inch DDGV) would not open on demand.

Initial analysis identified that a stem-disc separation occurred as a result of excessive wear of

the valve stem threads and shear failure of the wedge pin. The licensee has not completed

their root cause determination. The licensee reported this event in Licensee Event Report

(LER) 2017-003-00, dated April 12, 2017 (Agencywide Documents Access and Management

System (ADAMS) Accession No. ML17102B424).

The licensee had been using industry guidance to perform visual evaluations and diagnostic

testing on the valve. The guidance was based on earlier operating experience from an event on

October 20, 2012, at Browns Ferry Nuclear Plant, Unit 1. This event resulted in two reports under 10 CFR Part 21, Reporting of Defects and Noncompliance. The first Part 21 report was

issued by the Tennessee Valley Authority (TVA), Anti-Rotation Pin Failure in Anchor Darling

(Flowserve) Double Disc Gate Valve, dated January 4, 2013 (ADAMS Accession

No. ML13008A321) and the second by Flowserve, Wedge Pin Failure in Anchor/Darling

Double-Disc Gate Valve at Browns Ferry Nuclear Plant Unit 1, dated February 25, 2013 (ADAMS Accession No. ML13064A012).

Browns Ferry Nuclear Plant, Units 1 and 2

On October 20, 2012, a Browns Ferry Nuclear Plant, Unit 1, high-pressure coolant injection

(HPCI) steam isolation valve, which also serves as a containment inboard isolation valve, failed

its local leak rate test. Investigation revealed that, although the event was not a stem-disc

separation, the wedge pin failed and one of the two disc retainers (see Figure 1) fell from the

stem-to-upper wedge connection into a space between the valve discs, causing one of the two

discs not to properly seat. The valve was a 10-inch Anchor/Darling DDGV. It was installed in

2007 and had not been disassembled since installation. The licensee, TVA, submitted a

10 CFR Part 21 report, dated January 4, 2013 (ADAMS Accession No. ML13008A321), for this

failure. In the report, TVA determined that the wedge pin failed because the vendor had not

properly torqued the stem-to-upper wedge connection during manufacture.

Before the 2012 failure, Browns Ferry Nuclear Plant experienced two other wedge pin failures in

10-inch Anchor/Darling DDGVs. The first wedge pin failure involved a Unit 2 HPCI outboard

steam isolation valve installed in 2001 that failed during testing that same year. The vendor

determined that the stem-to-upper wedge connection was not properly torqued. The second

wedge pin failure occurred in 2008 and involved a Unit 1 HPCI outboard steam isolation valve

installed in 2006 that failed during local leak rate testing. Internal inspection of the valve

revealed that the stem-to-upper wedge connection was not properly torqued. Figure 1 Typical Anchor/Darling DDGV (ADAMS Accession No. ML13064A012)

Surry Power Station, Unit 2

On February 2, 2011, Surry Power Station, Unit 2, tripped as a result of a low-flow condition in

the reactor coolant system (RCS) C loop. The low-flow condition was the result of a stem-disc

separation of the RCS C loop isolation valve. The valve was an Anchor/Darling 30-inch

DDGV. Inspection of the valve internals revealed that the wedge pin failed and the upper

wedge threads exhibited excessive wear. The root cause was determined to be flow-induced

vibration coupled with inadequate torque of the stem-to-upper wedge connection, documented

in LER 2011-001-00, dated April 1, 2011 (ADAMS Accession No. ML11105A032). A similar

stem-disc separation occurred in 1999 on the RCS A loop isolation valve (also an

Anchor/Darling 30-inch DDGV). In the 1999 event, the wedge pin failed, allowing the stem to

unthread from the upper wedge connection (see LER 1999-003-00, dated July 30, 1999 (ADAMS Legacy No. ML9908120152)).

River Bend Station, Unit 1

On May 21, 2007, an unexplained drop in the reactor recirculation system loop A flow occurred

at River Bend Station, Unit 1. Reactor power lowered to approximately 96.5-percent power with

no operator action. Operators determined that the most probable cause for the condition was

that the loop A discharge isolation valve caused partial flow blockage. The valve was a

20-inch Anchor/Darling DDGV. Further investigation during the plant shutdown revealed a

stem-disc separation and severely worn stem and upper wedge threads. The wedge pin failed, with the two portions extending into the upper wedge still in place while the piece that

transverses the shaft was missing. The licensee identified several contributing causes for the

valve failure, including inadequate torque of the stem-to-upper wedge connection during the previous valve assembly, and flow-induced vibration at the disc assembly caused by turbulent

flow through the valve coupled with partial extension of the disc assembly into the flow stream.

DISCUSSION

Wedge pin failures and stem-disc separation events associated with Anchor/Darling DDGVs

have occurred at both pressurized-water reactor and boiling-water reactor plants. As previously

mentioned, the 2012 event at Browns Ferry Nuclear Plant, Unit 1 resulted in the issuance of

10 CFR Part 21 reports by both TVA (ADAMS Accession No. ML13008A321) and Flowserve

(ADAMS Accession No. ML13064A012).

In its Part 21 report, Flowserve concluded that at Browns Ferry Nuclear Plant, Unit 1:

failure was due to the shearing of the wedge pin which serves a joint locking

function at the threaded interface between the valve stem and upper wedge. The

pin is designed to ensure that the joint does not loosen due to vibration and other

secondary loads. On some valve designs, the pin also is used to attach the disc

retainers to the upper wedge. The pin shearing allowed rotation of the stem

during the closing stroke when the valve was seating and ultimately resulted in

loss of the stem to upper wedge joint integrity.

Flowserve has completed an evaluation of the failure and concluded the root

cause of the wedge pin failure was excessive load on the pin. The stem

operating torque exceeded the torque to tighten the stem into the upper wedge

before installation of the wedge pin. The additional stem torque produced a load

on the wedge pin creating a stress which exceeded the pin shear strength

causing the failure. The recommended assembly stem torque did not envelope

the operating torque for the TVA application providing the potential for an over

load situation and ultimate failure. The operating torque for the TVA valve was

unusually high due to the fast closing time of the actuator and very conservative

closing thrust margin.

This situation can potentially occur on any Anchor/Darling type double-disc gate

valve with a threaded stem to upper wedge connection, typically size 2.5" and

larger, operated by an actuator that applies torque on the stem to produce the

required valve operating thrust. An operating stem torque greater than the

assembly stem torque can provide the opportunity for excessive pin load and

potentially failure.

The stems on most double-disc (DD) gate valves larger than size 2" are attached

to the upper wedge using UN [unified constant pitch] threads. A pin is installed

through the hub of the upper wedge and stem threaded section to prevent the

stem from loosening and eventually unscrewing from the wedge. In addition, the

disc retainers on some DD gate valves are attached using the wedge pin. See

Figure 1. The output torque of the actuator is transmitted to the stem/wedge joint

through the stem and is resisted by the disc wedge pack, therefore the stem to

wedge connection is loaded by the stem torque and thrust. The wedge pin is not

designed to withstand the full actuator output torque. The actuator torque

direction tends to tighten the stem into the wedge during closing and tends to

loosen the stem during opening. Flowserve's Part 21 also states:

Flowserve recommends that all critical Anchor/Darling Double-Disc Gate valves with

threaded stem to upper wedge connections and actuators that produce a torque on

the stem be evaluated for potential wedge pin failure.

This IN provides a discussion of the recent LaSalle County Station, Unit 2 Anchor/Darling DDGV

failure, events at Browns Ferry that led to Part 21 reporting, and other operating experience that

resulted in stem-disc separations. This document contains information available to NRC staff as

of May 2017. The licensee for LaSalle County Station, Unit 2 is in the process of completing

their root cause determination. Licensees can use this information in addition to the technical

information provided in the 2013 Flowserve Part 21 report to consider actions, as appropriate, to

avoid similar problems.

CONTACT

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

matter to the technical contacts listed below.

/ra/

/ra/

Louise Lund, Director

Timothy J. McGinty, Director

Division of Policy and Rulemaking

Division of Construction Inspection

Office of Nuclear Reactor Regulation

and Operational Programs

Office of New Reactors

Technical Contacts: John W. Thompson, NRR/DIRS

Michael Farnan, NRR/DE

301-415-1011

301-415-1486

E-mail: John.Thompson@nrc.gov

E-mail: Michael.Farnan@nrc.gov

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

ML17153A053; *concurred via email TAC No. MF9698 OFFICE

NRR/DPR/PGCB/LA

TECH EDITOR

NRR/DIRS/IOEB/TL

NRR/DIRS/IOEB/TL

NRR/DIRS/IOEB/BC

NAME

ELee

JDougherty*

JThompson*

EThomas*

HChernoff*

DATE

06/02/17

06/05/17

06/12/17

06/12/17

06/12/17 OFFICE

NRR/DPR/PGCB/LA

RI/DRS/DD

RII/DRS/DD

RII/DRP/D

RIII/DRS/D

NAME

ELee

JYerokun*

MMiller*

JMunday*

KOBrien*

DATE

06/12/17

06/13/17

06/13/17

06/13/17

06/07/17 OFFICE

RIV/DRP/D

RIV/DRS/DD

NRR/DSS/D

NRR/DRA/DD

NRR/DE/DD

NAME

TPruett*

JClark*

MGavrilas*

SWeerakkody* (A)

KCoyne* (A)

DATE

06/12/17

06/12/17

06/12/17

6/13/17

06/13/17 OFFICE

NRR/DLR/D

NRR/DIRS/D

NRR/DPR/PGCB/PM

NRR/DPR/PGCB/BC NRO/DCIP/D

NAME

GWilson*

CMiller*

NMartinez

AGarmoe (A)

TMcGinty

DATE

06/13/17

06/13/17

06/14/17

06/14/17

06/14/17 OFFICE

NRR/DPR/D

NAME

LLund

DATE

06/15/17


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