Bulletin 89-02: Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor Darling Model S350w Swing Check Valves or Valves of Similar Design

                                                       OMB No.:  3150-0011
                                                       NRCB 89-02

                                  UNITED STATES
                          NUCLEAR REGULATORY COMMISSION
                             WASHINGTON, D.C.  20555

                                  July 19, 1989

                         BOLTING IN ANCHOR DARLING MODEL S350W SWING 


All holders of operating licenses or construction permits for nuclear power 


The purpose of this bulletin is to request addressees to identify, disassemble 
and inspect certain types of swing check valves which may contain Type 410 
stainless steel (SS) bolting material.  If the Type 410 SS bolting material is 
of sufficiently high hardness that it is susceptible to stress corrosion 
cracking (SCC), or has failed, addressees are requested to take appropriate 

Description of Circumstances:

The occurrences discussed below have raised concerns about the use of Anchor 
Darling swing check valves, Model S350W, and valves of similar design with 
internal preloaded bolting material of ASTM Specification A193 Grade B6 Type 
410 SS.

Diablo Canyon, Unit 2 - In October 1988, the licensee performed a scheduled 
preventive maintenance on a swing check valve in the residual heat removal 
(RHR) system.  This valve had been successfully stroked by hand several times 
before the mechanic detected slight movement of the retaining block.  Further 
investigation showed that both retaining block studs shown in Figure 1 were 
broken.  The retaining block studs (bolts) retain the blocks that hold the 
valve disk assembly in place to the valve body as shown in Figure 1.  The 
valve was an 8 inch pressure isolation valve in piping attached to the reactor 
coolant system hot leg.  One bolt was broken at the block to valve body inter-
face and the other bolt was broken inside the retaining block.  There were 
signs of significant corrosion product buildup on the failed bolts.  The valve 
was manufactured by Anchor Darling.  Details of this failure are given in NRC 
Information Notice 88-85, "Broken Retaining Block Studs on Anchor Darling 
Check Valves," dated October 14, 1988.

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D.C. Cook, Units 1 & 2 - At D.C. Cook, Unit 2, during maintenance on an 8 inch
Anchor Darling swing check valve installed in the low-pressure emergency core 
cooling system (ECCS), the licensee performed an inspection of the valve 
internals.  One of the two internal preloaded bolts was found broken and the 
other cracked.  As a result of this finding, the licensee inspected the corre-
sponding check valve in the redundant low-pressure ECCS train.  Again, one of 
the two internal preloaded bolts was found broken and the other cracked.  This 
discovery prompted the licensee to expand the inspection to Anchor Darling 
check valves of the same design as those in which the degraded studs were 
found.  This included 12 valves, all classified as pressure isolation valves, 
in the ECCS and RHR systems at this plant.  The licensee identified one 
accumulator outlet check valve with a cracked bolt. 

Following the licensee's decision to initiate inspection of Unit 2 check 
valves, Unit 1 went from power operation to hot shutdown (Mode 4) because of 
an unrelated event.  The licensee decided to inspect the four Unit 1 check 
valves that were accessible in Mode 4 and found one broken bolt in each of the 
two check valves installed in the low-pressure ECCS.  The licensee provided 
details of these failures in a letter dated October 28, 1988.  

J.A. FitzPatrick Plant - Licensee Event Report 87-003 identifies broken bolts 
from the High Pressure Coolant Injection (HPCI) Terry turbine throttle valve 
lifting beam.  The bolts of Type 410 stainless steel with hardness in the 
upper Rc30 range failed from intergranular stress corrosion cracking.  


These occurrences raise questions concerning the operability and reliability 
of Anchor Darling Model S350W swing check valves with Type 410 SS retaining 
block studs and valves of similar design with internal preloaded bolting.  The 
internal bolts of these valves are of ASTM specification A193 B6 Type 410 
martensitic stainless steel with a tempering temperature of 1100 F and a 
specified minimum tensile strength, but no maximum specified tensile strength. 
The licensees determined the preliminary cause of the Type 410 SS material 
failure to be SCC.  Three parameters determine the susceptibility of Type 410 
SS to SCC:  heat treatment, environment, and stress magnitude.  Metal hardness 
is related to the heat treatment performed on the material and Rockwell hard-
ness values below Rc26 are indicative of heat treatments that are generally 
less susceptible to SCC.  Before Winter 1974, hardness control was exercised 
only through meeting the tempering temperature requirement in ASME SA193-B6.  
The maximum hardness requirement of the ASTM A193-B6 was incorporated into the 
ASME Code in Winter 1974.  The current hardness requirements would probably 
have been met had the material actually been tempered at the required 1100 F 
for the appropriate time.  The susceptibility of martensitic steel (B6 and B7) 
to SCC increases for hardness values exceeding Rc26.  The B6 bolting material 
with limitation on maximum hardness is designated as B6X.  One of the two 

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                                                                 July 19, 1989
                                                                 Page 3 of 6

broken bolts examined for the preliminary results indicated that the hardness 
was Rc36.  This value is significantly higher than the desired range and is 
more susceptible to SCC than a properly heat-treated bolt.

The Anchor Darling check valves at Diablo Canyon are in lines filled with 
borated water and are not used during normal plant operation.  The interior of 
the valve with the broken bolt was found to be rusty, while the others were 
clear.  This may indicate that the bolt failure could have been a result of 
events or conditions present before system operation and that failure of the 
bolt may not have been a result of these valves being in a borated water 

Although the internal preloaded bolts of these check valves do not experience 
loading from valve operation or from system pressure, stress does result from 
initial torquing of the bolts.  Because excessive preloading aides SCC, it is 
important that the valve maintenance manuals contain the valve vendor's torqu-
ing requirements.  In addition, stress on the bolts could be produced by 
differential thermal expansion of the dissimilar metals (retaining block 
materials of Type 304 SS and bolts of Type 410 SS will result in a 
differential growth of approximately 50 percent).  However, this consideration 
does not preclude using replacement stud material of similar thermal expansion 
as that of Type 410 SS provided that the replacement material is less 
susceptible to SCC. 

Care should be taken when testing check valves after maintenance.  Recently 
(May 20, 1989) Salem Unit 1 experienced a loss of residual heat removal capa-
bility while flow testing a check valve in an accumulator line.  Related 
guidance on check valve disassembly and post-maintenance testing where the 
ASME Code, Section XI requirements are impractical can be found in Generic 
Letter 89-04, "Guidance on Developing Acceptable Inservice Testing Programs." 

Actions Requested: 

I.   For all licensees of operating reactors:

     A.   All licensees of operating reactors are requested to disassemble and 
          inspect all safety-related Anchor Darling Model S350W swing check 
          valves supplied with internal retaining block studs of ASTM specifi-
          cation A193 Grade B6 Type 410 SS.  Licensees should review the 
          design of other safety-related check valves to determine if similar 
          designs and material selection to the Anchor Darling Model S350W are 
          used.  If so, such valves should be similarly inspected.  The 
          inspection by disassembly should be performed as follows:

          1.   If any of the internal bolting is to be reused, it should be 
               inspected for cracks using surface inspection techniques (pene-
               trant or magnetic particle).  Cracked bolting should be 
               replaced and a failure analysis performed including chemical 
               analysis to confirm material type.
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                                                                 Page 4 of 6

          2.   If all suspect bolting is to be replaced with bolting of 
               material and hardness specified in I.A.3, surface inspection 
               and failure analysis of the old bolting may not be needed 
               unless an unexpected failure mechanism is evident.

          3.   Reused and new bolting should be hardness tested for a maximum 
               Rockwell hardness value of Rc26.  Any internal preloaded 
               bolting that does not meet the hardness requirements should be 
               replaced by bolts of the same material with a maximum Rockwell 
               hardness of Rc26 or by an alternate material approved by the 
               valve manufacturer.

     B.   Inspections of Anchor Darling Model S350W swing check valves are 
          requested to be performed at the next refueling outage or scheduled 
          outage of sufficient duration (four weeks or longer) that begins 90 
          days after receipt of this bulletin.  Documentation review to 
          identify similar swing check valves with internal preloaded Type 410 
          SS bolting in the facility and the inspections are requested to be 
          performed at the next refueling outage that begins 180 days after 
          receipt of this bulletin.

II.  For all applicants for Operating Licenses:

     A.   The Actions Requested are the same as I.A. above.

     B.   The implementation of the Actions Requested is requested to be 
          complete before fuel loading, or, if fuel loading occurs within 90 
          days of receipt of this bulletin, at the first refueling outage 
          after receipt of this bulletin.

Reporting Requirements: 

Activities performed in response to this bulletin shall be documented and 
maintained in accordance with plant procedures for safety-related equipment 
and reported as follows: 

1.   Addressees who do not have Anchor Darling Model S350W swing check valves 
     with Type 410 SS bolts subject to this bulletin and do not have valves of 
     similar design with preloaded Type 410 SS bolt material shall within 180 
     days of receipt of this bulletin provide a letter of confirmation to the 
     NRC of these facts.

2.   Addressees who do have swing check valves subject to this bulletin shall 
     provide a letter to the NRC within 60 days of completion of the inspec-
     tions stating the number of valves inspected and the number of valves 
     found to have service induced cracking of bolting. 
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                                                                 July 19, 1989
                                                                 Page 5 of 6

The documentation of the valve inspection to be maintained by the licensee 
shall summarize the inspection findings and include the items listed below:

     a.   The number and location of subject swing check valves inspected.

     b.   The number of subject swing check valves where broken and/or cracked 
          retaining block bolts are found.

     c.   The extent of any cracking found, the nondestructive examination 
          methods used and the acceptance criteria employed.

     d.   The number of subject bolts that were replaced and type of material 

     e.   The results of any failure analysis.

3.   Licensees unable to meet the above schedules shall submit a report to the 
     staff with technical justification and alternative schedules as appropri-
     ate within 30 days after the need for scheduler relief is realized.

Although not requested by this bulletin, addressees are encouraged to work 
collectively to address the technical concerns associated with this issue, as 
well as to share information regarding valves similar to Anchor Darling Model 

The letters required above shall be addressed to the U.S. Nuclear Regulatory 
Commission, ATTN:  Document Control Desk, Washington, D.C.  20555, under oath 
or affirmation under the provisions of Section 182a, Atomic Energy Act of 
1954, as amended and 10 CFR 50.54(f).  In addition, a copy shall be submitted 
to the appropriate Regional Administrator. 

This request is covered by Office of Management and Budget Clearance Number 
3150-0011 which expires December 31, 1989.  The estimated average burden hours 
is 60 person-hours per valve, including assessment of the new recommendations, 
searching data sources, gathering and analyzing the data, and preparing the 
required letters.  These estimated average burden hours pertain only to these 
identified response-related matters and do not include the time of actual 
implementation of physical changes consistent with the requested actions.  
Send comments regarding this burden estimate or any other aspect of this 
collection of information, including suggestions for reducing this burden, to 
the Records and Reports Management Branch, Division of Information Support 
Services, Office of Information Resources Management, U.S. Nuclear Regulatory 
Commission, Washington, D.C.  20555; and to the Paperwork Reduction Project 
(3150-0011), Office of Management and Budget, Washington, D.C.  20503. 

The radiation dose that would be incurred by the actions in this bulletin is 
strongly dependent on the location of the valves in question.  The limited 
experience to date indicates that the dose can range from less than 0.1 
person-rem per valve to about 2.5 person-rem per valve depending on the 
location of the valve within the plant system. 
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                                                                 July 19, 1989
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If you have any questions about this matter, please contact one of the techni-
cal contacts listed below or the Regional Administrator of the appropriate 
regional office.

                              Charles E. Rossi, Director
                              Division of Operational Events Assessment
                              Office of Nuclear Reactor Regulation

Technical Contacts:  Thomas McLellan, NRR
                     (301) 492-3218

                     C. David Sellers, NRR
                     (301) 492-0930

                     N. Prasad Kadambi, NRR
                     (301) 492-1153

1.  Figure 1.  D.C. Cook 1 and 2 and Diablo Canyon 2
2.  List of Recently Issued NRC Bulletins


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