ML20011C609
| ML20011C609 | |
| Person / Time | |
|---|---|
| Issue date: | 06/13/1989 |
| From: | Sniezek J Office of Nuclear Reactor Regulation |
| To: | Jordan E Committee To Review Generic Requirements |
| Shared Package | |
| ML20011C480 | List: |
| References | |
| IEIN-88-085, NUDOCS 8906290271 | |
| Download: ML20011C609 (15) | |
Text
/,g** *884gk WITS D STATES o
NUCLE A3 CECULATORY COMMISSION
(
WAnt.6 tog 1OW.D.C.PM66 JUN 131989 MEMORANDUM FOR:
Edward L. Jordan, Chairman Committee to Review Generic Requirements FROM:
James H. Sniezek, Deputy Director Office of Nuclear Reactor Regulation
SUBJECT:
REQUEST FOR REVIEW 0F A DRAFT BULLETIN REGARDING STRESS CORROSION CRACKING OF HIGO HARDNESS TYPE 410 SS INTERNAL PRELOADED BOLTING IN AN', HOR DARLING MODEL $350W SWING CHECK VALVES OR VALVES Or SIMILAR DESIGN 1988, the NRC staff issued Information hotice No. 88 85, ' Broken On October 14,k Studs On Anchor Darling Model $350W Swing Check Valves."
Retaining Bloc The information notice alerted addressees to potential problems relating to failure of internal retainine block stud material (ASTM A193 Grade B6 Type 410 stain-less steel ($$)) on Anchor Darling mode) $350W swing check valves at Diablo Canyon and D.C. Cook.
In October 1988, at Diablo Canyon, Unit 2, a scheduled preventive maintenance activity parformed on a check valve in the Residual Heat Removal S At D.C. Cc,ok, Units 1 &ystem revealed two retaining block studs to be broken.2, a scheduled mainten one of the two retaining block studs was broken and the other cracked.
In addition, LER 87 003 for the FitzPatrick plant identifies failed high hardness Type 410 bolts from the HPCI Terry turbine throttle valve lifting beam. The preliminary root cause of Type 410 $$ material failure was deternined to be stress corrosion cracking (SCC). As a result, compliance with General Design Criteria (GDCs) 1, 14, 30, 32, 34, 35 of Appendix A 10 CFR 50, and 10 CFR 50.556 may be suspect because the integrity of the primary coolant boundary and the operability of accident mitigation systems may be adversely affected.
The proposed bulletin requests licensees or construction pemit holders to identify Anchor Darlirig model $350W swing check valves supplied with retaining block studs of ASTM A193 Grade 86 Tyr* 410 $$ and similar valves with internal preloaded bolting of this material.
e identification is to be followed by disassembly and inspection of the sta, et valves. The topic of this proposed bulletin was reviewed with industry re>resentatives during a January 12, 1989 meeting.
The schedules proposed for t'.e actions and reports would ensure that the concern would be resolved in approximately two years af ter receipt of this bulletin.
CCNTACT: Prasad Kadambi, KRR 492-1153 960?7l YA
c-l~
l, Edward L. Jordan 2
JUN 18 $$$
The enclosed draft bulletin would ensure compliance with 10 CFR Part $0, i
c Appendix A. GDC 1, 14, 30, 32, 34, 35, and possibly others.
Hence, no further staff evaluation pursuant to 10 CFR $0.54(f) is required.
r Enclosed are the proposed bulletin and background information, including infornetion obtained from the Diablo Canyon, FitzPatrick, and D.C. Cook plants.
We request that review of this package be scheduled at CRGR's earliest conve-t
[ngineering and Systems Technology. y Lawrence Shao, Director, Division of nience.
The bulletin is sponsored b j
.]g.,,,aL,Y.4 ALUjE-7 James H. Snierek, Deputy Director O(ficeofNuclearReactorRegulation
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Enclosures:
1.
NRC Bulletin No. 89 XX 2.
CRGR ltem IV.B.
3.
References Submitted to CRGR
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June 13, 1939 OMB No.: 3150-0011.
l NRCB 89-XX i
i UNITED STATES I
NUCLEAR REGULATORY COMMIS$10N i
0FFICE OF NUCLEAR REACTOR REGULATION l
WASHINGTON, D.C.
20555 l
Mr.y xx, 1989 o
NRC BULLETIN NO. 89-XX:
STRESS CORROSION CRACKING 0F HIGH-HARDNESS TYPE 410 STAINLESS STEEL INTERNAL PRELOADED i
SOLTING IN ANCHOR DARLING MODEL $350W SWING CHECK VALVF.S OR VALVES OF SIMILAR DESIGN
\\
Addressees:
1 All holders of operating licenses or ct.o:truction permits for nuclear power reactors.
Puroose:
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 sufficientl
. cracking (SCC)y high hardness that it is susceptible to stress corrosfon or has failed, addressees are requested to take appropriate.
actions.
I hscription of. Circumstances:
f The occurrences discussed below have raised concerns about the use of Anchor Darling swing ek ' valves, Model S350W, and valves of similar design with i
L internal preloaaed 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 i
(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 I were broker.
The retaining block studs (bolts) retain the blocks that hold the L
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 i.
l' coolant system hot leg. One bolt was broken at the block to valve body inter-fato and the other bolt was broken inside the retaining block. There were signs of significant corrosion product buildup on the failed Mits.
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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'+f NRCB Sg-XX l
May xx, 1989 e
Page 2 of 6 I
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n D.C. Cook Units ' & 2 - At D.C. Cook, Unit 2, during maintenance on an 8 inch ij Anchor Darling sw ng) check valve installed in the low-pressure emergency core i
cooling system (ECCS, the licensee perfomed an inspection oi' the valve internals. One of the two internal preloeded bolts was found broken and the corresponding check valve in the redundant low e licensee inspected the other cracked.
As a result of this finding th pressure ECCS train. Again, one of the two internal preloaded bolts was found broken and the other cracked.
l 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, i'
Yhis included 12 valves, all classified at pressure isolation valves, in the ECCS and RHR systems at this plant. The licensee identified one accumuhter a
i
' outlet chcck valve with a cracked bolt.
Following the licensee's decision to initiate inspection of Unit 2 check i
valves, Unit I went from pcwer operation to hot shutdown (Mode 4) because of an l
' unrelated event. The licensee decided to inspect the four Unit I check valves that were accessible in Mode 4 and found one broken belt in each of the two check valves installed in the low-pressure ECCS. The licensee provided details l
'of these failures in a letter dated October 28, 1988, i
J.A. FitzPatrick Plant - Licensee Event Report 87-003 identiilet broken bolts fro'n the M1gh Pressure Coolant injection (MFCI) Terry turbine throttle Yalve lifting beam. The bolts of Type 410 stainless steel with hardness in the upper Rc30 range failed from intergranular stress corrosion cracking.
Discussion:
These occurrences raise questio'ns concerning the operaH11ty 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 l
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 detemine the susceptibility of Type 410 4
$$ to SCC: hcat treatment, environment, ard stress magnitude. Metal hardntes is related to the heat treatment perforr.ed on the material and Rockwall hard-1 ness values below Rc26 are indicative of heat treatments that are generally i
less susceptible to SCC.
Before Winter 1974, hardness control was exercised only through meeting the tempering temperature requirement in ASME SA193-86.
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 haroness values exceeding Rc26.
The 86 bolting meterial with limitation on maximum hardness is designated as 86X. One of the two
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NRCB Sg-XX i
May xx, 1989 Page 3 of 6 l
1W.
broken bolts examined for the preliminary results indicated that the hardness l
7 han Rc36. This value is significantly higher than the desired range and is mure 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 i
the valve with the broken bolt was found to be rusty, while the others were I
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 be,vn a result of these valves being in a borated water l
cnvironment.
l Although the internal preloaded bolts of these check valves do not experience I
loading from valve operation or from system pressure, loading sides SCC, it is stress does result from initial torquing of the bo'its. Because excessive pre
~important that the valve maintenance manuals contain the valve vendor's torqu-t 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 t
growth of approximately 50 percent). However, this consideration does not 5
preclude using replacement stud material of similar thnrmal expansion as that
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of Type 410 SS provided that the replacement material is less susceptible to SCC.
3 Care should be taken when testing check valves after meintenance. Recently (May 20, 1989) Salem Unit 1 experienced a loss of resicual heat removal capability while flow testing a check valve in an accumult, tor line. An information notice on this mot is currently being prepared. Related l
guidance on check nho disassembly and post-maintenance testing where the ASME Code,Section XI requirenwnts are impractical can be found in Generic Letter 89-04.
l Actions Requested:
1.
Foi all licensees of operating reactors:
A.
All licensees of operating reactors are requested to disassemble and inspect al 56fety-related Anchor Darling Model 5350W swing check valves supplied with internal retaining block studs of ASTM specifi-catirm A193 Grade B6 Type 410 SS. Licensees should review the design of other safety-related check velves to determine if similar designs i
and meterial selection to the Anchor Darling Model 5350W are used.
If so, such valves should be similarly inspected.
Tha inspection by disassembly should be perfurrst as follows:
1.
Valve internals should be inspected for corroded or worn parts.
2.
If any of the interna) bolting is to be reused, it should be inspected for cracks using surface inspection techniques (pene-trant or r;agnetic particle).
Cracked bolting should be replaced end a failure analysis perforte including chemical analysis to l
confirm material type.
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i NRCB 89-XX H
May xx, 1989 1
Page 4 of 6 j
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3.
If all suspect bolting is to be replaced with bolting of material.
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and hardness specified in I.A.4, surface inspection and failure i
analysis of the old bolting may not be needed unless an unexpected
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failure mechanism is evident.
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l 4.
Reused and new bolting should be hardness tested for a maximum i
Rockwell hardness value of Rc26. Any internal preloaded bolting that does not meet the heroness requirements should be replaced t
by bolts of the same material with a maximum Rockwell hardnesc
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of Re?6 or by an alternate material approved by the valve q
manufacturer.
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-i 5.
When reassembling valves, the internal preloaded bolting should be preloaded according to the valve vendor's torquing i
specifications.
6.
To ensure proper bolt preloading for future valve maintenance i
activities, licensees should verify that their plant's mainte-nance manuals include the valve vendor's torquing specifications.
B.
Inspections of Anchor Darling Model 5350W 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 identi-fy similar swing check valves with internal preloaded Type 410 SS bolting in the facility and the inspections are requested to be perfonned at the next refueling outage that begins 180 days after receipt of this bulletin.
l
. II. For all applicants for Operating Licenses:
j 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.
Reportina 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.
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1 NRCB 89-XX
'l May xx, 1989 Page 5 of 6
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2.
Addressees who do have swing check valves subject to this bulletin shall
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provide a letter to the.NRC within 60 days of completion of the inspec-
.j tions stating the number of valves inspected and the number of valves found to have service induced cracking of bolting.
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The documentation of the valve inspection to be maintained by the licensee shall. susmerize the inspection findings and include the items listed l
below:
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The number and location of subject swing check valves inspected.
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a.
b.
The number of subject swing check valves' with broken and/or cracked J
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retaining block bolts.
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c.
Yhe extent of cracking found, the nondestructive. examination methods
^
L used and the acceptance criteria employed.
l d.
The rumber of subject bolts that were replaced and type of material used.
e.
The results of any failure analysis.
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3.
1.icensees unable to meet the above schedules shall submit a report to the ttaff with technical justification and alternative schedules as appropri-ate within 30 days after the need for schedular relief is realized.
5 Altho 0gh not requested by this. bulletin, addressees are encouraged to wn't collectively to address the technical concerns associated with this issue, as well as tc share information regarding valves similar to Anchor Darling Model
{
t S350W.
The'1etters required above shall be addressed to the U.S. Nuclear Regulatory i
Comission, ATTN: Document Control Desk, Washington, D.C.
20555, under oath or affirmation under the provisions of Section 182a Atomic Energy Act of 1954, i
as amended and 10'CFR 50.54(f).
In addition, a copy shall be submitted to the i
l appropriate Regional Administrator.
This request is covered by Office of Management and Budget Clearance Number i
3150-0011 which expires December 31, 1989. The estimated average burden hours is 60 person-hours per valve, including assessment of the new recomendations, t
p 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 L
f implementation of physical changes consistent with the requested actions. Send L
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 Comission.
Washington, D.C.
20555;andtothePaperworkReductionProject(3150-0011),
-Office of Management and Budget Washington, D.C.
20503.
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NRCB sg-xx.
P May xx. 1989-c,
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Page 6 of 6
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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-j rem per valve to about 2.5 person-rem per valve depending on the location of the valve within the plant system.
l If you have any questions about this matter, please contact one of the techni-l
. cal contacts listed below or the Regional Administrator of the appropriate j
u regional office.
j i
Charles E. Rossi, Dirtetor I
Division of Operational Events Assessment Office of Nuclear Reactor Regulation
' t Technical Contacts:. Thomas McLellan, NRR (301)492-3218 C. David' Sellers NRR l
-(301)492-0930 N. Prasad Kadambi.'NRR (301)492-1153
Attachment:
L1:;t of Recently Issued NRC Bulletins i
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l-June 13, 1939 i
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l CRGR Item IV.B. Contents of Packages Submitted to CRGR l
(Rev. 4 Stello to List 042387, des 41860342ff)
- o.
The following re or (regulatory) quirements apply for proposals to reduce' existing requirementsJ l
positions as well as proposals to increase requirements or (regulatory) positions.
Each package submitted to the CRGR for review shall i
include twenty (20) copies of the following infomation:
SUBJECT:
BULLETIN REGARDING STRESS CORROSION CRACKING OF HIGH. HARDNESS TYPE 410 STAINLESS STEEL INTERNAL PRELOADED BOLTING IN ANCHOR DARLING r
MODEL S350W SWING CHECK VALVES OR VALVES OF SIMILAR DESIGN Question:
I.
The proposed generic requirement or staff position as it is proposed to be i
sent out to licensees.
r i
Response
I The proposed staff position is set forth in the bulletin inclosure 1).
s Question:
II.. Draft staff papers or other underly(A copy of all materials referenced in ing staff documents supporting the requirements or staff positions.
the document shall be made available upon request to the CRGR staff. Any committee member may request CRGR staff to obtain a copy of any referenced material for his or her use.)
Response
1.
NRC Information Notice 88-85 " Broken Retaining Block Studs in Anchor Darling Check Valves " October 14, 1988.
2.
NRC " Evaluation of Justification for Continued Operation of Diablo Canyon, Unit 1," December 20, 1988.
3.
PGAE Licensee Event Report for Diablo Canyon Units 1&2, regarding Anchor Derling Check Valve retaining block stud breakage due to intergranular stress corrosion cracking, November 18, 1988.
4 Indiana Michigan Power report for D.C. Cook Units 1&2, " Degradation i
of Retaining Block Studs in Darling Valve and Manufacturing Company Clear Waterway Check Valves," October 28, 1988.
'5.
FitzPat '-k Licensee Event Report 87-003 regarding Type 410 cracked bolts..
qh hardness on the HPCI Terry Turbine Throttle Valve Lifting 1
i *
.c Question:
!!!. Each proposed requirement or staff position shall contain the sponsoring.
office s position as to whether the proposal would increase staff require-l ments or staff's positions, or would implement existing requirements or staff positions.
l
Response
A.
The >otential for the degradation of the Anchor Darling Model S350W l
chec( valves with internal preloaded.410 SS studs increases with service tine; therefore, the staff has determined that it is neces-i sary to request the licensees and construction permit holders for nuclear power reactors to disassemble and inspect these valves.
This augmented inspection is necessary because the presence of failed bolts may not be identified during nomal ASME Section XI valve testing and the degraded bolt condition could later hinder or prevent functioning of the valve as designed.
To ensure the continued function of the safety systems where the Anchor Darling Model S350W swing check valves with internal preloaded 410 SS bolts and similarly designed swing check valves are employed, the. staff proposes that these valves be disassembled and inspected for broken and/or cracked bolts. The staff proposed augmented inservice inspection is based on the provision of 10 CFR 50.55a(g)(6)(11).
B.
The requirenants of the proposed bulletin would ensure compliance with 10 CFR Part 50, Appendix A General Design Criteria (GDC) 1,14, 30, 32, 34, 35, and possibly others related to maintaining the integrity of safety-systems with Anchor Darling Model S350W swing chec c valves or similarly derigned valves.
In addition, the proposed bulletin is justified on the basis of 10 CFR Part 50, Appendix B, paragraphs VIII and XVI addressing identification of defective parts and taking corrective action, GOC 1 requires in part that the recognized codes and standards used in required testing be supplemented or modified as necessary to ensure a quality product in keeping with the required safety func-tion. The other GOC's relate to integrity of the primary coolant boundary and accioent mitigation systems.
Question:
1 IV. The proposed method of implementation with the concurrence (and any comments) of OGC on the method proposed, i
Response
The method of implementation will be the proposed bulletin (Enclosure 1).
A copy of this bulletin hhs been reviewed by OGC.
Comments received were incorporated.
OGC has no legal objection.
2
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-Question:
L V.
Regulatory analysir.
.- ?.unfoming to the directives and guidance of MUREG/8R-0058 and Nbat, 18.
_ Response:
A regulatory analysis is not required because the purpose of the bulletin -
is to bring addressees into compliance with the terms of their operating licente.
Question:
VI.
Identification of the category of reactor plants to which.the generic requirement or staff posit' on is to apply (that is, whether it is to apply to new plants only, new OLs [ operating licensees] only, OLs after a certain date, all CLs, all plants under construction, all plants, all water reactors, all PWRs [ pressurized water reactors) only, some vendor types, some vintage types such as BWR 6 and 4. jet pump and nonjet pumps plants,etc).
Resoonse:
The proposed bulletin would apply to all holders of operating licenses or construction permits for boiling water reactors and pressurized water 3
reactors.
Question:
VII. For each such category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other.
ongoing' regulatory activities. The evaluation shall document for consid-L eration information available concerning any of the following factors as may be appropriate and any other information relevant and material to the proposed action:
l l
A.
Statement of the specific objectives that the proposed action is 1:
designed to achieve.....
Rjsponse:
l The objective of this bulletin is to ensure that addressees inspect L
Anchor Darling Model S350W swing check valves and other valves of 8 i similar construction, which are used in safety related applications, and may contain high hardness Type 410 SS bolting.
The stress corrosion cracking of such bolting may have degraded the operability s
of the valves to unacceptable levels.
Question:
h 1
L B.
Genersi description of the activity that would be required by the licensee or applicant in order to complete the action.....
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Resoonse:
t The activity involves, in addition to records reviews, inspection of the internal preloaded bolting with surface insaction methods which would require valve disassembly.
If preloaded >olting of Type 410 SS exists within the valve, hardness testing and/or replacement with other bolting acceptable to the valve vendor would be required. The preload should be consistent with vendor recomnendations.
If the maintenance procedures do not already incorporate it, torquing specifications consistent with vendor reconnendations need to be placed there.
Question:
C.
Potential change in the risk to the public, from the accidental
'offsite release of radioactive material....
Response _:
l The potential presence of broken bolts in safety-related check valves has an adverse affect on the likelihood of offsite release of radioactive material. The actions taken pursuant to this bulletin would decrease the risk to the public from such accidental releases.
f Question:
r D.
Potential impact on radiologict1 axposure of facility employees and other onsite workers
Response
The potential for radiological exposure strongly depends on valve locations. At D.C. Cook the staf f was informed that 19 valves were inspected resulting in total exposure of approximately 11 man-rems.
t However, four RHR valves in the population of 19 valves accounted for an exposure of'10 man-rem.
Question:
t E.
Installation and continuing costs associated with the action, includ-ing the cost of facility downtime or the cost of construction delay.....
Response
Installation costs will be incurred only if replacement of bolting is necessary.
These costs should be minimal.
No continuing costs are expected once assurance of bolt integrity is established.
With proper outage planning, no unusual fcility downtime is expected.
No construction delays are anticipated, Based on information from D.C.
Cook, the inspection related activities are expected to take approxi-mately 50 man-hours per valve. Record searches and documentations 4
u
- r, i
i are estimated at approximately 10 man-hours per valve. Based on information from Diablo Canyon and D.C. Cook, there may be 10 to 12 Anchor Darling check valves at a unit.
l Question:
l
,1 F.
.The potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing s
regulatory requirements and staff positions.....
Response
l No change in operational complexity is expected, j
- Question:
?
G.
The estimated resource burden on the NRC associated with the proposed action and the availability of such resources.....
Response
The staff does not expect to review individual plant submittals. Yne project managers are expected to verify that individual licensees have responded as required. The information received will be com-m l
piled and sumarized when the submittals are complete.
Question:
H.
The potential impact of differences in the facility type, design, or age on the relevancy.and practicality of the proposed action....
L
Response
(
L Based on the currently available information, no differences are 1
1
- expected, t
OvesL1on:
I.
Whether the propcsed action is interim or final, and if 1.ceri F
f ustification for. imposing the proposed action on an interir.1 Les :.
Response
l l.
The proposed action is final.
p lz Question:
VIII.
For each evaluation conducted pursuant to 10 CFR 50.109, the proposing i
L Office Director's determination together with the rational for the determination based on the consideration of paragraph I. through VII.
above that:
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A.
There is a substantial increase in the overall protection of public health and safety or the common defense and security to be derived from the proposal.
B.
The direct and indirect costs of iniplementation, for the facilities affected, are justified in view of this increased protection.
Response
The actions requested in this bulletin would ensure that licensees comply with those consnitments pursuant to their license which require that safety related equipnent perform their intended function when J
required to do so. Hence, under 10 CFR 50.109(a)(4)(1) no further cost analysis is required.
Ouestion:
IX.
For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing Office Director's determination, together with the rationale for the determination bastd on l
the ennsiderations of paragraphs I, through VII, above that:
A.
The public health and safety and the common defense and security would be adequately protected if the proposed reduction in require-I ments or positions were implemented.
B.
The cost savings attributed to the action would be substantial enough to justify taking action.
Response
No relaxation of requirements will occur from this bulletin.
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9 6
Y
l, 7 UNITED STATES l
NUCLEAR REGULATORY COP 9t!$$10N l
0FFICE OF NUCLEAR REACTOR REGULATION I
E WASHINGTON, D.C. 20655 October 14, 1988 l-NRC 1NFORMAT10N NOTICE NO. 88-85: BROKEN RETAINING BLOCK STUDS ON ANCHOR DARLING CHECK VALVES i
L
~ Addressees:
All holders of operating licenses or construction permits for nuclear power reactors.
l Puroose:-
l This infomation notice is being provided to alert addressees to potential 1'
problems relating to the failure of retaining block studs on Anchor Darling L
check valves and the possible generic implications. Diablo Canyon, Unit 2 and D.C. Cook Units 1 and 2 have recently reported problems with this type of failure.
It is expected that recipients will review the infonnation fer applicability to their facilities and consider actions, as appropriate, D l
svoid similar problems. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or l
written response is required.
Descriotion of Circumstances:
In October 1988 at Diablo Canyon, Unit 2, a scheduled preventative maintenance
- performed on a check valve in the Residual Heat Removal (RHR) System revealed
.that two retaining bicek studs (see drawing) were completely broken. The valve is an 8-inch pressure isolation valve in piping attached to the Reactor Coolant System hot leg. One stud was sheared at the block to valve body interface and the other stud was br') hen off inside the retaining block. There were signs of t
significant' corrosion product build-up on the failed studs. The valve was manufactured by Anchor Darling.
Discussion:
The licensee has taken# actions to repair the RHR valve, and a metallurgical evaluation of the failed studs.is underway. The companion RHR valve at Diablo Canyon, Unit 2, has been disassembled and found to be acceptable. The licensee has also disassembled eight 10-inch Anchor Darling swing check valves in the Safety Injection (SI) Systtem. These valves are also pressure isolation valves.
. These valves have also been found to be acceptable. The stud material in all ten of these valves is ASTM A193 Grade B6 Type 410 srainless steel. A discussion 4-of problems noted with 410 stainless steel parts in other valve applications is contained in Information Notice 85-59, " Valve Stem Corrosion Failures."
. m **
DOAUleull2 1
.i
IN 88-85 l
1 L
October 14, 1988 Page 2 of 2 Several weeks prior to the failure at Diablo Canyon, Unit 2. D.C. Cook, Unit 2
- discovered similar stud failures in Anchor Darling swing check valves. One broken stud and one cracked stud were discovered in 8-inch RHR low. head in-jection check valves. These valves are the second check val es back from the reactor coolant loop hot legs, and they therefora act as prmure isolation valves. At D.C. Cook, Unit 2, there are two adMtional 8-inch Anchor Darling
]
swing check valves.. There are also eight 10-inch Anchor Darling swing check l
valves in the $1 system; the.two RHR valves and six of the eight 51 valves l
have been inspected. One of the $1 accumulator injection check valves was i
also found to have a cracked stud. The remainder of the valves inspected were found to be acceptable.
At D.C. Cook, Unit 1, the two RHR injection check valves were inspected and each found to have one cracked stud. Two $1 accumulator injection check valves i
were inspected and found to be acceptable. Metallurgical evaluation of the' failed studs is ongoing with the preliminary analysis indicating that the failures are due to intergranular stress corrosion cracking.
Based upon discussions with the valve vendor, the NRC has learned that the af-facted valves at the plants discussed above are Anchor Darling Model. Number
$350WSC, Drawing Number g4-12892. The vendor also indicated that based pri-
. marily on experience with pressure boundary bolting, they have been using Type 17-4PH stainless steel for studs in borated water service. They no longer manufacture the $350WSC valve, but they recomend replacement studs be made of Type 17-4PH stainless steel rather than Type 410.
l The NRC staff believes that with seriously degraded studs, the retaining blocks could be dislodged if these valves are called upon to open rapidly during accident conditions. This could lead to blockage of the flow path, and the valves would be incapable of resenting.
Licensees may wish' to consider potential actions that would be appropriate if one of these check valves should fail in service prior to inspection of the bolts. Such actions might include appropriate procedures and operator training.
No specific action or written response is required by this information notice.
f If you have any questions about this matter, please contact.the technical
- contact listed below or the Regional Administrator of the appropriate regional office.
J Charles E. Rossi Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical
Contact:
Ted Sullivan, NRR (301)492-0901 Attachments.
- 1. Figure of Valve
- 2. List of Recently Issued NRC Information Notices uu
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IN 88-85 October 14, 1988-l Page 1 of 1 j
LIST OF RECENTLY !$$UED 1
NRC INFORMATION NOTICES Infomation Date of Notice No.
_.%bject Issuance issued to i
r fective Motor Shaft 10/20/88 All holders of CLs 88-84 e
Keys in Limitorque Motor or cps for nuclear Actuators power reactors.
88-83 Inadequate Testing of Relay 10/19/88 All holders of OLs
'l Contacts in Safety-Related or cps for nuclear Logic Systems power reactors.
88-82 Torus Shells with Corrosion 10/14/88 All holders of OLs and Degraded Coatings in or cps for SWRs.
8WR Containments 88-81 Failure of Amp Window 10/7/88 All holders of OLs Indent Kynar Splices or cps for nuclear and Thomas and Betts power, test, and Nylon Wire Caps During research reactors.
Environmental Quali-fication Testing 88-80 Unexpected Piping Movement 10/7/88 All holders of OLs Attributed to Thermal or cps for PWRs.
Stratification 88-79 Misuse of Flashing Lights 10/7/88 All holders of OLs for High Radiation Area or cps for nuclear Controls power reactors.
88-69, Supp 1 Movable Contact Finger 9/29/88 All holders of OLs Binding in HFA Relays or cps for nuclear Manufactured by General power reactors.
Electric (GE)
Implementation of Revised 9/22/8B All holders of OLs 88-78 NRC-Administered Requali-or cps for nuclear fication Examinations power reactors.
88-77 Inadvertent Reactor 9/22/88 All holders of OLs Vessel Overfill or cps for 8WRs.
t OL = Operating License CP = Construction Permit 4
18
. : =:...:::.=..:=....
. - - - r 1N 88-85 l
October 14, 1988 t
1 Page 2 of 2 j
i Several weeks prior to the failure at Diablo Canyon, Unit 2. D.C. Cook, Unit 2, discovered similar stud failums in Anchor Darling swing check valves. One broken stud and one cracked stud were discovered 'n 8-inch RHR low heiad in-s
,jection check valves. These valves are the second check valves back from the reactor coolant loop hot legs, and they therefore act as pressure isolation j
valves. At D.C. Cook, Unit 2, there are two additional 8-inch Anchor Darling swing check valves. There are also eight 10-inch Anchor Darlin swin check valves in the $1 systemt the two RHA valves and six of the ei h $1 y 1ves 1
have been inspected. One of the $1 accumulator injection che k valves was also found to have a cracked stud. The remainder of the valves inspected were found to be acceptable.
At D.C. Cook, Unit 1, the two RHR injection check valves were inspected and each found to have one cracked stud. Two $! accumulator injection check valves were inspected and found to be acceptable. Metallurgical evaluation of the l
failed studs is ongoing with the preliminary analysis indicating that the l
failures are due to intergranular stress corrosion cracking.
t I
Based upon discussions with the valve vendor, the NRC has learned that the af-fected valves at the plants discussed above are Anchor Darling Model Number i
l 5350WSC, Drawing Number 94-12892. The vendor also indicated that based pri-marily on experience with pressure boundary bolting, they have been using Type 17-4PH stainless steel for studs in borated water service. They no longer l
l manufacture the 5350WSC valve, but they recommend replacement studs be made of Type 17-4PH stainless steel rather than Type 410.
The DC staff believes that with seriously degraded studs, the mtainin b1rscks could be dislodged if these valves are called upon to open rapid y l
during accident conditions. This could lead to blockage of the flow path, and the valves would be incapable of reseating.
f Licensees.nay wish to consider potenti.a1 actions that would be appropriate if one of these check valves should fail in service prior to inspection of the l
bolts. Such actions might include appropriate procedures and operator training.
No specific action or written response is required by this infomation notice.
[
If you have any questions about this matter, please contact. the technical L'
contact listed below or the Regional Administrator of the appropriate regional office.
p Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation p
Technical
Contact:
'h, (301)492-0901 Attachments:
- 1. Figure of Valve L
- 2. List of Recently Issued NRC Infomation Notices L
- SEE PREVIOUS CONCURRENCE l
EAB:NRR ACT:C:EAB:NRR NRK:EMEB ACT:C:RI WRR: DEST C:0GCB:NRR D:DOEA:NRR L
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