ML20129D867
| ML20129D867 | |
| Person / Time | |
|---|---|
| Issue date: | 07/25/1996 |
| From: | Jordan E Committee To Review Generic Requirements |
| To: | Miraglia F NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9609300215 | |
| Download: ML20129D867 (27) | |
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UNITED STATES i (kt_.I j NUCLEAR REGULATORY COMMISSION pgM
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wassinoron. o.c. 2ossmi July 25. 1996 MEMORANDUM TO:
Frank J. Miraglia, Jr., Deputy Director Office of Nuclear Reactor Regul FROM:
Edward L. Jordan, Chairman h
Committee to Review Generic Req rements
SUBJECT:
YOUR MEMORANDUM OF JULY 19,1996, ON CRDM CRACKING The subject memo to the CRGR requested that a formal review of a proposed GL on the subject of CRDM cracking is not warranted, because the GL involves no new or revised regulatory requirements. The letter would be issued for comment. The CRGR has no objection to issuance of the letter for comment, without CRGR review. We would like to review the final version, however.
There is some inconsistency in tae rationale. At step (ix) of the CRGR information it is stated that NRC is not requesting any new actions. Rather, it is requesting the information already gathered. Yet, the GL action 1.2.a makes it clear that the results for subsequent inspections are wanted (and clearly this information must be gathered in the future). Also, the backfit discussion notes that the NRC needs to determine whether to impose augmented requirements in order to maintain public health and safety. Perhaps a more appropriato phrase is to determine compliance with regulatory requirements.
I have attached a copy of a report prepared by INEL under AEOD sponsorship,
" Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking," NUREG/CR-6245 (October 1994). I believe that it would be a useful annex to send along with the GL, for information.
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3 waswtworon, o.c. anses4ooi July 19, 1996 MEMORANDUM TO:
Edward L. Jordan, Chairman Committee to Review Generic Requirements FROM:
Frank J. Miraglia, Jr., Deputy Director Office of Nuclear Reactor Regulation g
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SUBJECT:
REQUEST FOR ENDORSEMENT, WITHOUT FORMAL REVIEW, 0F TH PROPOSED GENERIC LETTER ENTITLED " PRIMARY W CORROSION CRACKING OF CONTROL ROD DRIVE M VESSEL HEAD PENETRATIONS" (TAC NO. M95280) i Review Generic Requirements (CRGR) endorse e
proposed Generic Letter GL).
published in the federal (Register for public comment.Following endorsem 1
NRR believes that formal review of the proposed GL is not warranted bec 1
the GL involves no new or revised regulatory requirements.
GL is to request addressees to describe their program for ensuring t The purpose of the i
inspection of the PWR CROM and other vessel head penetrations.
1 availability of the information derived from these monitoring and inspection The programs will aid NRC staff in assessing whether addressees are in complianc with existing rules and regulations.
) is the GL as proposed by the staff.
i to (1) require addressees to describe their program for assuring timelyT i
inspection of pressurized water reactor (PWR 1
provide to the NRC a written response to this Ge control rod drive mechanism i
i requested information.
The staff considers this GL to be Category 2. is the response to the questions contained in Section IV B o CRGR Charter.
the required responses regulated by 10 CFR 50.54(f).The respo A notice of opportunity for public comment on the proposed GL will be published in the federal Register prior to issuing the proposed GL.
have informed the NRC staff that they are taking appro The PWR
, and preclude a safety issue from developing.
plants age. penetrations (VHPs) has occurred and is expected to continue does not pose a safety concern in the near ters, the NRC s cracking of VHPs to be a safety concern for the long term based on the possibility of:
(1) exceeding the ASME Code margins if the cracks are
Contact:
C. E. Carpenter, NRR 415-2169
Edward L. Jordan
, l i
sufficiently deep and continue to propagate during subsequent operating cycles; and, (2) eliminating a layer of defense-in-depth 4
required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, i
Appendix A, GDC 14) continues to be satisfied, and to ensure that the safety i
significance of VHP cracking remains low, the NRC staff requires licensees to submit information to assess compliance with the above stated requirements.
The NRC staff finds that the requested information is also needed to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety.
i The staff is not establishing a new position for such compliance in this GL.
The Office of the General Counsel reviewed this GL and has no legal 2
objections.
Furthermore, OGC has determined that the proposed GL is not a
" Rule" under the provisions of the Small Business Regulatory Enforcement Fairness (SBREF) Act (see 5 U.S.C., Chapter 8) enacted on March 29, 1996.
The GL is sponsored by Brian W. Sheron, Director, Division of Engineering.
Attachments:
1.
Proposed Generic Letter, titled " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations" 2.
Response to CRGR Charter Questions Distribution: see next page DOCUMENT NAME:
G:\\CARPENTR\\CRDMCRGR.MEM
- See Previous Concurrence To receive a copy of this document, indicate in the box:
"C" - Copy without attachment / enclosure "E" - Copy with attachment / enclosure "N" - No copy EMCB: LPM E
EMC8:SC E
EMCB:BC E
DE:D l
CECarpenter*
KRWichman*
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05/22/96 05/22/96 05/28/96 05/30/96 OGC l
DRPM:D(A) l ADT:AD gl NRR:00 M / M LClark*
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06/10/96 06/24/96 G/7 /96 4' 7/N/96 RECORD. COPY I~
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Edward L. Jordan sufficiently deep and continue to propagate during subsequent operating cycles; and, (2) eliminating a layer of defense-in-depth for plant safety.
The information collected will enable the staff to verify that the margins required by the ASME Code,.as specified in Section 50.55a of Title 10 of the Code of federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) continues to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the NRC staff requires licensees to submit information to assess compliance with the above stated requirements.
The NRC staff finds that the requested information is also needed to determine if the imposition of an augmented inspection program, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety.
The staff is not establishing a new position for such compliance in this GL.
The Office of the General Counsel reviewed this GL and has no legal objections.
Furthermore, 0GC has determined that the proposed GL is not a
" Rule" under the provisions of the Small Business Regulatory Enforcement Fairness (SBREF) Act (see 5 U.S.C., Chapter 8) enacted on March 29, 1996.
The GL is sponsored by Brian W. Sheron, Director, Division of Engineering.
Attachments:
1.
Proposed Generic Letter, titled " Primary Water Stress Corrosion Cracking of Control Rod Drive Mechanism and Other Vessel Head Penetrations" 2.
Response to CRGR Charter Questions Distribution:
)
PUBLIC PD I-1 EMCB Reading OGC DOCUMENT NAME:
G:\\CARPENTR\\CRDMCRGR.MEM
- See Previous Concurrence To receive a copy of this document, indicate in the box:
"C" - Copy without attachment / enclosure "E" - Copy with attachment / enclosure "N" - No copy EMCB: LPM E
EMCB:SC E
EMCB:BC E
DE:D E
CECarpenter*
KRWichman*
JRStrosnider*
BWSheron*
05/22/96 05/22/96 05/28/96 05/30/96 OGC lE DRPM:D(A)
E ADT:AD E
NRR:DD E
LClark*
BKGrimes*
ACThadani*
FJMiraglia*
06/10/96 06/24/96 07/09/96 07/19/96 RECORD COPY
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UNITED STATES NUCLEAR REGULATORY COMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 JuKl9AJ996 i
GENERIC LETTER 96-#d:
PRIMARY WATER STRESS CORROSION CRACKING OF CON i
DRIVE MECHANISM AND OTHER VESSEL HEAD PENETRATION (TAC NO. M95280)
Addressees
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All holders of operating licerses for pressurized water reactors (PWRs),
j except those licenses that have been amended to possession-only status.
1 i
Purnose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) request addressees to describe their program for ensuring the timely j
inspection of PWR control rod drive mechanism (CROM penetrations and (2) require that all addressees pro) vide to the NRC a written and other vessel head i
response to this generic letter relating to the requested information.
Backaround Most PWRs have Alloy 600 CROM nozzle and other vessel head penetrations (VHPs that extend above the reactor pressure vessel head.
The stainless steel housing of the CROM is screwed and seal-welded onto the top of the nozzle penetration, as shown in Figure 1.
The weld between the nozzle and the housing is a dissimilar metal weld, which is also called a bimetallic weld.
The nozzles protrude below the vessel head, thus exposing the inside surface of the nozzles to reactor coolant.
The control rod drive (CRD) nozzles and other VHPs are basically the same for all PWRs worldwide, which use a. U.S.
design (except in Germany and Russia).
Generally, there are 36 to 78 nozzles distributed over the low-alloy steel head.
The vessel head is semi-spherical and the head penetrations are vertical so that the CRD nozzles and other VHPs are not perpendicular to the vessel surface except at the center.
The uphill side (toward the center of the head) is called the 180-degree location and the downhill side (toward the outer periphery of the head) is called the 0-degree location.
Most nozzles have a thermal sleeve with a conical guide at the bottom end and a small gap (3-to 4-am) between the nozzle and the sleeve.
The NRC staff identified primary water stress corrosion cracking (PWSCC) as an emerging technical issue to the Commission in 1989, after cracking was noted in Alloy 600 pressurizer heater sleeve penetrations at a domestic PWR facility.
Other leaks have occurred since 1986 in several Alloy 600 pressurizer instrument nozzles at both domestic and foreign reactors from sevaral different nuclear steam supply system vendors.
The NRC staff reviewed the safety significance of the cracking that occurred, as well as the repair
GL 96-##
l July 19,'1996 Page 2.of 9 i
and replacement activities at the affected facilities.
The NRC staff detennined that the cracking was not of immediate safety significance because j
the cracks were axial, had a low growth rate, were in a material with an j
unlikely to propagate very far. extremely high flaw tolerance (high fracture i
cracking would result in detectable leakage and the opportunity to tak corrective action before a penetration would fail.
The NRC staff issued Inconel 600," dated FebruaryInformation Notice 90-10. " Primary Wat i
23, 1990, to inform the nuclear industry of the j
issue.
i Bugey 3, a French PWR.In December 1991, cracks were found in an Alloy 60 i
Sweden, Spain, and Japan have uncovered additional VHPs with i
About 2 percent of the VHPs examined to date contain short, axial cracks.
Close examination of the VHP that leaked at Bugey 3 revealed very minor t
j incipient secondary circumferential cracking of the VHP.
}
An action plan was implemented by the NRC staff in 1991 to address PWSCC of Alloy 600 VHPs at all U.S. PWRs.
As explained more fully below, this action development of VHP mock-ups by the Electric Power the qualification of inspectors on the VHP mock-ups by EPRI, the review of proposed generic acceptance criteria from the Nuclear Utility Management and Resource Council (NUMARC) (now the Nuclear Energy Institute (NEI)), and VHP inspections. As part of this action plan Westinghouse Owners Group (WOG) on January, the NRC staff met with the 7, 1992, the Combustion Engineering Owners Group (CEOG) on March 25, 1992, and the Babcock & Wilcox Owners Group (B&WOG) on May 12, 1992, to discuss their respective programs for VHPs in their respective plants since all of the plants hav Subsequently, the NRC staff asked NUMARC to coordinate future industry actions because the issue was applicable to all PWRs.
NEI and the PWR Owner's Groups on the issue on August 18 and Novemb 1992 March 3, 1993, December 1, 1994, and August 24, 1995.
Suunaries of these meetings are available in the Commission's Public Document Room, 2120 L Street, N.W., Washington, D.C. 20555.
Each of the PWR Owners Groups submitted safety assessments, dated February 1993, through NUMARC to the NRC on this issue.
safety assessments and examining the overseas inspection findings, the NRC staff concluded in a safety evaluation dated November 19, 1993, that VHP cracking was not an immediate safety concern. The bases for this conclusion were that if PWSCC occurred at VHPs (1) the cracks would be predominately axial in orientation, catastrophic failure, a(nd (32) the cracks would result in detectable leakage before examinations performed as par)t of surveillance walkdown inspections the leakage would be detected during visual significant damage to the reactor vessel head would occur.
In addition, the NRC staff had concerns related to unnecessary occupational radiation exposures associated with eddy current or other forms of nondestructive examinations (NDEs), if performed manually.
Field experience in foreign countries has
1 GL 96-H July 19, 1996 Page 3^of 9 shown that occupational radiation exposures can be significantly reduced using remotely controlled or automatic equipment to conduct the inspections.
equipment and repair tools that reduced radiation exposur qualification protocol developed and administered by the Techniques and i
the demonstrations, examinations by rotating and saber eddy current andIn also sized within reasonable uncertainty bounds. ultrasonics sh reliably detect PWSCC in CROM nozzles.also demonstrated that p a
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i In 1994 circumferential intergranular attack IGA) associated with the J-groove, weld in one of the CROM penetrations w(as discovered at Z i
Spanish reactor.
This IGA is a different degradation mechanism than the PWSCC described above.
exchange resin bed intrusions, which resulted in high concentrat sulfates.
Zorita has 37 CROM penetrations, of which 20 are active penetrations and 17 are spare penetrations.
penetrations showed stress corrosion cracking and IGA. Sixteen of the 17 spare 3
i axial and circumferential.
The cracks were both significant cracking with axial and circumferential cracks.Four of the active i
i ingress events occurred at Zorita.
In August 1980, 40 liters of cation resin Two cation resin entered the reactor coolant system (RCS).
In September 1981, a mixed bed i
domineralizer screen failed and between 200 to 320 liters of resin e RCS.
The coolant conductivity remained high for at least 4 months after the j
ingress.
The increase in conductivity was attributed to locally high concentrations of sulfates.
I Sulfates were found around the crack areas and on i
the fracture surfaces.
occur in regions that are not subject to significant applied or resid j
stresses.
1 i
Resins Increases Potential for Stress Corrosion Crackin
" Ingress of Demineralizer Mechanism Penetrations," dated February 14, 1996, to alert addressees to the increased likelihood of sulfate-driven stress corrosion cracking of PWR CRDMs and other VHPs if domineralizer resins contaminate the RCS.
1 plants of the Zorita incident by issuing NSAL-94-028.The Westin Westinghouse reported similar to that at the Zorita plant.that no other plant had been found wo j
that U.S. plants monitor RCS conductivity on a routine basis, follow the i
guidelines on primary water chemistry, and monitor for sulfate three times a week.
The Westinghouse staff concluded that no immediate safety issue is i
The Westinghouse staff suggested that U.S. PWR plants r 3
chemistry and other operating records pertaining to sulfur ingress events.
The results of this review have not been reported to the NRC staff, and the 4
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GL 96-M July 19, 1996 I
Page 4 of 9 NRC staff does not have sufficient information to a The first U.S. inspection of VHPs took place in the spring of 1994 at the i
Point Beach Nuclear Generating Station, and no indications were uncovere j
any of its 49 CRDM penetrations.
The eddy current inspection at the Oconee a
Nuclear Generating Station in the fall of 1994 revealed 20 indications in on l-penetration.
indications because they were shallow. Ultrasonic testing (UT) did not rev i
tre less than one mil deep (0.03 nui).
UT cannot accurately size defects that a
with the original fabrication and may not grow; however, they will beThe i
reexamined during the next refueling outage.
A limited examination of eight in-core instrumentation penetrations conducted at the Palisades plant foun cracking.
An examination of the CRDM penetrations at the D. C. Cook plant in i
the fal' of 1994 revealed three clustered indications in one penetration.
i indications were 46 ime,16 mm, and 6 to 8 mm in length, and the deepest fla The j
was 6.8 mm deep.
The tip of the 46-am flaw was just below the J-groove weld.
Virginia Electric and Power Company inspected North Anna Unit I during i spring 1996 refueling outage.
hillsides) were examined on each outer ring CRDM penetrations and j
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indications were observed using eddy current testing.
The NRC staff was informed during a meeting on August Westinghouse had developed a susceptibility model for VHPs based on a num 24, 1995, that of fabrication of the VHP, microstructure of the VHP, i
i i.
VHP on the head.
results are incorporated into the model.Each time a plant's VHPs are inspecte All domestic Westin been modeled and the ranking has been given to each licensee.ghouse PWRs have Babcock & Wilcox (8&W ], also developed a susceptib In addition,
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penetration nozzles an)d other VHPs in B&W reactor vessel de j
domestic B&W PWRs have been modeled and the ranking has been given to All licensee.
had performed an initial susceThe NRC staff was further informed that Comb present, neither Westinghouse,ptibility assessment for the CE PWRs.
At FTI, nor CE has submitted its models and j
assessments to the NRC staff for review.
3 ahPene 1996, NEI submitted a white paper entitled " Alloy 600 V
a on P significance of PWSCC in PW ess Corrosion Cracking," which reviews the j
the issue.
The program outlined in he white paper is based on the
" ' # '8 **"*9I"9 assumption that the issue i n
i n
j describesaneconomicdecisfonto rather than a safety issue, and be used by PWR licensees to evaluate the probability of a VHP developing a crack or a through-wall leak during a plant's lifetime This inf j
evaluate the need to conduc n w uld then be used by a PWR licensee to informed NEI in the several P inspection at their plant.
The NRC staff NEI that the issue was on1 eatings listed above that it did not agree with hasinitiatedinsomeU.S.pfant
[ king ad e n t y ha no y
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i-GL 96-#
July 19, 1996 Page 5 of 9 other VHPs to PWSCC to justify an inspection plan bas l
considerations alone.
Discussion analyses by the PWR Owners Groups, the NRC st 1
dated November 19, 1993, and the PWSCC found in the CRDMs in European reactors.
On the basis of the results of the first five inspections of U.S.
PWRs, the PWR Owner's Groups' analyses, and the European experience, t j
staff has determined that there is a high probability that VHPs at other plants may contain similar axial cracks caused by PWSCC.
Zorita, residual stresses are sufficient to cause circumfe Further, if any j
4
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intergranular stress corrosion cracking (IGSCC).
i After considering this information, the NRC staff has concluded that VHP j
cracking does not pose an immediate or near term safety concern.
i NRC staff recognizes that the scope and timing of inspections may vary for Further, the different plants depending on their individual suceptibility to this form of degradation.
j In the long ters, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation.
i The vessel head provides the vital function of maintaining a reactor pressure boundary.
Cracking in the VHPs has occurred and is expected to continue to j
occur as plants age.
concern for the long ters based on the possibility of (1) exceed American Society of Mechanical Engineers (ASME) Code for margins if the cra cycles, and (2) eliminating a layer of defense in depth f j
Therefore, in order to verify that the margins required by the ASME Code, as i
specified in Section 50.55a of Title 10 of the Code of Fe i
Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) is continue remains low, the NRC staff believes that an integrated, lo i
li which includes periodic inspections and monitoring, is necessary. program, determine if the imposition of an augmented inspection In i
j 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and safety.
The NRC staff recognizes that individual PWR licensees may wish to determine (i.e., B&WOG, CEOG, WOGtheir inspection activities based on an integrated or some subset thereof, to take advantage of inspection results from,other plants that have s)imilar susceptibilities.
NRC staff does not wish to discourage such group actions but notes that such The
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an integrated industry inspection program must have a well-founded technical basis that justifies the relationship between the plants and the planned implementation schedule.
GL 96-#
July 19, 1996 Page 6 of 9 Reouired Information The information requested by items 1 and 2, below, is required by the NRC staff to determine if the imposition of an augmented inspection program is required, while the information requested by item 3 relates to the potential i
for domestic resin intrusions, such as occtrred at Zorita.
i Addressees are required to provide the following information:
[
1.
Regarding inspection activities:
4 1.1 A description of all inspections of CRDMs and other vessel head penetrations performed to the date of this generic letter, including the results of these inspections.
1.2 If you have developed a plan to periodically trispect the CRDM and other vessel head penetrations:
i Your schedule for first, and subsequent, inspections of the 3
a.
CRDM and other vessel head penetrations, including the technical basis for your schedule.
i l
i b.
Your scope for the CRDM and other vessel head penetration inspections, including whether you plan to inspect from the 4
top or bottom of the head, the total number of penetrations have thermal sleeves, which are) spares, and which instrument or other penetrations.
1.3 If you have agi developed a plan to periodically inspect the CRDM i
and other vessel head penetrations, provide your technical or i
safety basis for not periodically inspecting your VHPs; or, your i
schedule for developing such a plan and the basis for that t
- schedule, i
2.
i A description of the evaluation methods and results used to arsess the susceptibility of the CRDM and other VHPs in your plant to PWSCC, including the susceptibility ranking of your plant and the factors used to determine this ranking.
Other than or in addition to the boric acid
+
visual examination (see Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure Eoundary Components in PWR Plants," dated March 17, 1988),
include a description of all relevant data and/o-tests used to develop crack initiation and crack crowth models, and the 4
methods and data used to validate these models.
Include a statement explaining the applicability of these models to the VHP cracking issue.
Also, if you are relying on any integrated industry inspection program, provide a detailed description of this program.
3.
A description of any resin intrusions in your plant, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water
i GL 96-N
~
July.19, '1996 Page 7;ofL9 Chemistry Guidelines recommendations for primary water sulfate leve 4
including the following information:
l 3.1 i
Were the intrusions cation, anion, or mixed bed?
3.2 What were the durations of these intrusions?
i 3.3 Do your RCS water chemistry Technical Specifications follow the s
EPRI guidelines?
1 2
3.4 i
Identify any RCS chemistr;* excursions that exceed your plant administrative limits for the followin i
chlorides or fluorides, oxygen, boron,g species:
- sulfates, and lithium.
3.5 Identify any conductivity excursions which may be indicative of J
resin intrusions, provide your technical assessment of each excursion and your followup actions.
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3.6 i
Provide your assessment of the potential for any of these for IGA of VHPs and any associated plan for inspec
?
i Reautred Resnonse 1
All addressees are required to submit a written response with the information requested above within 90 days from the date of this letter.
I Any inspection results that do ngi satisfy the acceptance criteria identified in the NRC staff's safety assessment dated November. 16, 1993, should be reported to the NRC staff prior to plant restart.
t i
Commission, ATTN: Address the required written reports to the U.S. Nuclear Re or affirmation under the provisions of Section 182a, Atomic E i
1954, as amended, and 10 CFR 50.54(f).
In addition, submit a copy to the appropriate regional administrator.
age of records, proprietary data, etc.) that licensees ma i
ascertaining whether they have all of the data 1
their CRDMs and other vessel head penetrations. pertinent to the evaluation of j
For this reason, the above time periods are allowed for the responses.
i Related Generic Communications a
i (1)
(PWSCC) of Inconel 600," dated FebruaryInformation Notice 4
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23, 1990, i
(2)
Information Notice 96-11. " Ingress of Domineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 1996.
-, +. - -
GL 96-##
July 19, 1996 i
Pqe 8 of 9 Backfit Discussion This generic letter only requires information from the addressees under the provisions of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f).
The information collected will enable the staff to verify that th required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 Appendix A, GDC 14) continues to be satisfied, and to e 10 CFR Part 50, submit information to assess compliance with the above stated requirements.
The NRC staff finds that the requested information is also needed to determine i
if the imposition of an augmented inspection program, pursuant to i
The staff is not establishing a new position for such compl j
generic letter.
and no documented evaluation or backfit analysis need be prepar i
Federal Reaister Notification i
A notice of opportunity for public comment was published in the Federal Register (XXiFRIXXXXX) W (e'g $ tenliceEs M thComments were received from on date.
the number orcommsetors 4 ty 2
s ses ree'indstry organizatlessitwoLpublicElsterestigrouisfeedftweindividuals)bthe staff eva save been made available 'n the public Copies of document room.
l Pacerwork Reduction Act Statement This generic letter contains information collections that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information i
collections were approved by the Office of Management and Budget, approval j
number 3150-0011, which expires July 31, 1997.
i The public reporting burden fo average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response,r this collection of information is estimated to including the time for reviewing instructions, and completing and reviewing the collection of information. search
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The U.S. Nuclear Regulatory Commission is seeking public connent on the potential impact of the collection of information contained in the generic letter and on the following issues:
1.
Is the proposed collection of information necessary for the proper performance of the fun:tions of the NRC, including whether the 4
information will have practical utility?
2.
Is the estimate of burden accurate?
3.
Is there a way to enhance the quality, utility, and clarity of the information to be collected?
GL 96-N July-19, 1996 i
Page,9;of 9 4.
How can the burden of the collection of information be m including the use of automated collection techniques?
suggestions for reducing this burden, to the Informa Management Branch, T-6 F33 U.S. Nuclear Regulatory Commission i
20555-0001, Affairs, NED8-10202 (3150-0011),and to the Desk Officer, Office of Info i
Office of Management and Budget i
Washington, DC 20503.
to, a collection of information unless it displays a c control number.
i If you have any questions about this matter please contact one of the Regulation (NRR) project manager. technical contacts listed below l
J 4
Brian K. Grimes, Acting Director i
Division of Reactor Program Management Office of Nuclear Reactor Regulation 4
Technical contacts:
Keith R. Wichman 4
(301) 415-2757 James Medoff (301) 415-2715 e-mail: krwenrc. gov e-mail: jxmenrc. gov f.
Lead Project Manager:
C. E. Carpenter, Jr.
i (301) 415-2169 i
e-mail: cec 9nrc. gov j
Attachments:
1.
References 2.
List of Recently Issued NRC Generic Letters i
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j CRGR REVIEW PACKAGE i 4 j PROPOSED ACTION: Issue a generic letter on the primary water stress corrosion cracking of control rod drive mechanism and other vessel head penetrations. ] CATEGORY: 2 RESPONSE TO RE0VIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW i (1) The proposed generic requirement or staff position as it is proposed to be sent out to licensees. Where the objective or intended result of a j proposed generic requirement or staff position can be achieved by setting a readily quantifiable standard that has an unambiguous i relationship to a readily measurable quantity and is enforceable, the i proposed requirement should merely specify the objective or result to be attained, rather than prescribing to the licensee how the objective or result is to be attained. j The information requested by items 1 and 2, below, is required by the NRC staff to determine if the imposition of an augmented inspection program is required, while the information requested by item 3 relates to the potential for domestic resin intrusions, such as occurred at 4 i Zorita. ) Addressees are required to provide the following information: 3 1. Regarding inspection activities: 4
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A description of all inspections of CRDMs and other vessel head 1.1 penetrations performed to the date of this generic letter, including the results of these inspections. I 1.2 If you have developed a plan to periodically inspect the CRDM and other vessel head penetrations: a. Your schedule for first, and subsequent, inspections of the CRDM and other vessel head penetrations, including the technical basis for your schedule. ? b. Your scope for the CRDM and other vessel head penetration inspections, including whether you plan to inspect from the top or bottom of the head, the total number of penetrations (and how many will be inspected), and which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations. 1.3 If you have agi developed a plan to periodically inspect the CRDM and other vessel head penetrations, provide your technical or safety basis for not periodically inspecting your VHPs; or, your schedule for developing such a plan and the basis for that schedule. J d I
CRGR REVIEW PACKAGE - i i A description of the evaluation methods and results used to assess 2. the susceptibility of the CRDM and other VHPs in your plant to PWSCC, including the susceptibility ranking of your plant and the factors used to determine this ranking. Other than or in addition to the boric acid visual examination (see Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," dated March 17,1988), include a description of all relevant data and/or tests used to develop crack initiation and crack growth models, and the methods and data used to validate these models. Include a statement explaining the i applicability of these models to the VHP cracking issue. j Also you are relying on any integrated industry inspection program,, if provide a detailed description of this program. 3. A description of any resin intrusions in your plant, as described in 1 IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate i levels, including the following information: 3.1 Were the intrusions cation, anion, or mixed bed? l 3.2 What were the durations of these intrusions? 3.3 Do your RCS water chemistry Technical Specifications follow the EPRI guidelines? 4 3.4 Identify any RCS chemistry excursions that exceed your plant administrative limits for the followin chlorides or fluorides, oxygen, boron,g species:
- sulfates, and lithium.
3.5 Identify any conductivity excursions which may be indicative of resin intrusions, provide your technical assessment of each excursion and your followup actions. i 3.6 Provide your assessment of the potential for any of these intrusions to result in a significant increase in the probability for IGA of VHPs and any associated plan for inspections. (ii) Draft staff papers or other underlying staff documents supporting the requirements or staff positions. (A copy of all materials referenced in the document shall be made available upon request to the CRGR staff. Any Commiittee member may request CAGR staff to obtain a copy of any reference material for his or her use.) (1) Information Notice 90-10, " Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600," dated February 23, 1990. (2) NRC staff safety evaluation, " Potential Reactor Vessel Head Adaptor Tube Cracking," dated November 19, 1993 (3) Informattor. Notice 96-11, " Ingress of Demineralizer Resins increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 1996.
i i CRGR' REVIEW PACKAGE t 1 (iii) Each proposed requirement or staff position shall contain the spon 1 office's position as to whether the proposal would increase requirements or staff positions or would relax or r, educe existing requirements or staff pos The results of domestic VHP inspections are consistent with the Febru j 1993 analyses by the PWR Owners Groups, the NRC staff safety evaluation i report dated November 19, 1993, and the PWSCC found in the CRDMs in European reactors. On the basis of the results of the first five i inspections of U.S. PWRs, the PWR Owner's Groups' analyses, and the i i European experience, the NRC staff has determined that there is a high i-probability that VHPs at other plants may contain similar axial cracks j caused by PWSCC. Further, if any significant resin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses are i sufficient to cause circumferential intergranular stress corrosion i cracking (IGSCC). 4 After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern. j Further, the NRC staff recognizes that the scope and timing of inspections may vary for different plants depend!ng on their individual i suceptibility to this form of degradation. degradation of the CROM and other VHPs is an important safetyIn the consideration that warrants further evaluation. The vessel head Cracking in the VHPs has occurred and is expected to c j l as plants age. concern for the long term based on the possibility of (1) e American Society of Mechanical Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent i operating cycles, and (2) eliminating a layer of defense in depth for j plant safety. Therefore in order to verify that the margins required by the ASME Code, as spec,ified in Section 50.55a of Title 10 of the Code i j of federal Regulations (10 CFR 50.55a) are met, that the guidance of j General Design Criterion 14 of Appendix A to 10 CFR Part 50 j (10 CFR Part 50, Appendix A GDC 14) is continued to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the } NRC staff believes that an integrated, long-term program, which includes periodic inspections and monitoring, is necessary. In addition, the NRC staff finds that the requested infomation is also needed to determine i if the imposition of an augmented inspection program, pursuant to } 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and j safety. The NRC staff recognizes that individual PWR licensees may wish to 3 i determine their inspection activities based on an integrated industry inspection program (i.e., B&WOG, CEOG, WOG, or some subset thereof), to take advantage of inspection results from other plants that have similar susceptibilities. The NRC staff does not wish to discourage such group actions but notes that such an integrated industry inspection program j must have a well-founded technical basis that justifies the relationship l between the plants and the planned implementation schedule. I
4 i CRGR REVIEW PACKAGE i ! } (iv) The proposed method of implementation with the concur renca (and i comments) of 0GC on the method proposed. i The concurrerece program offices or an explanation of any nonconcurrences. of affected { See attached concurrence page. I (v) NUREG/8R-0058 and NURESRegulatory analyses conforming to the j ensure compliance or ens /CR-3564.ure, define (This does not apply for backfits j in these cases a documented evaluation is required as discussed in, or IV.8.(1x).) i l Not applicable 1 (vi) Identification of the category of reactor plants to which the generic i requirement or staff position is to apply (that is, whether it is to i before a certain date, all Ols, all plants under constru j plants, all water reactors, all pWRs only, some vendor types, see vintage types such as 8WR 6 and 4, jet pump and nonjet pump plants, ) etc.). ) i All holders of operating licenses for pressurized water reactors (PWRs) 1 except those licenses that have been amended to possession-only status. i (vii) For backfits other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.10g. The backfit analysis shall include, for each category of N. actor plants, an evaluation Aich 3 i of other ongoing regulatory activities. demonstrates how the actio { The backfit analysis shall document for consideration information available concerning any of the i following factors as may be appropriate and any other information relevant and material to the proposed action: (a) Statement of the specific objectives that the proposed action is designed to achieve; Not applicable. (b) General description of the activity that would be required by the licensee or applicant in order to complete the action; Not applicable. \\ (c) potential change in the risk to the public from the accidental release of rad'oactive material; + Not applicable. (d) potential impact on radiological exposure of facility employees i and other onsite workers; Not applicable. a
CRGR REVIEW PACKAGE. (e) including the cost of facility downtime or the cost o construction delay; Not applicable. (f) The potential safety impact of changes in plant or operational complexity, including the relationship of pro regulatory requirements and staff positions; posed and existing Not applicable. (g) The estimated resource burden on the NRC associated with th proposed action and the availability of resources; Not applicable. (h) The potential impact of differences in facility type, design, or age on the relevancy and practicality of the proposed action; Not applicable. (1) Whether the proposed action is interin or final, and if interim, the justification for imposing the proposed action on an interim basis; Not applicable. (j) How the action should be prioritized and scheduled in light of other ongoing regulatory activities. may be appropriate in this regard: The following information 1. i The proposed priority or schedule, 2. i A stannary of the current backlog of existing requirements awaiting implementation. { 3. An assessment of whether implementation of existing i requirements should be deferred as a result, and 4. i Any other information that may be considered appropriate with regard to priority, schedule, or cumulative impact. For example, could implementation be delayed pending public comment? \\ Not applicable. (viii) For each backfit analyzed pursuant to 10 CFR 50.10g a)(2) (i.e., 4 not adequate protection backfits and not compliance (backfits), the i proposing Office Director's determination, together with the rational for the determination based on the consideration of paragraph (1) and (vii) above, that: (a) There is a substantial increase in the overall protection of i public health and safety or the common defense and security to be derived from the proposal; and i r i
). CRGR REVIEW PACKAGE - (b) The direct and indirect costs of implementation, for the j facilities affected, are justified in view of this increased protection. l } Not applicable. i (ix) For adequate protection or compliance backfits evaluated pursuant to j 10 CFR 50.10g(a)(4) (a) a documented evaluation consisting of: (1) the objectives of the modification i i (2) the reasons for the modification (3) the basis for invoking the compliance or adequate protection l exemption. ~ (b) in addition, for actions that were immediately effective (and the evaluation shall document the safety significance I j appropriateness of the action taken and (if applicable) j j consideration of how costs contributed to selecting the solution among various acceptable alternatives. Not applicable. The proposed generic letter is a request for information only. The NRC staff is not requesting any new actions from i PWR licensees to provide to the NRC information that } already told the NRC staff it has gathered, but has not shared with the l NRC to date. (x) 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 based 1 on the considerations or paragraphs (1) through (vii) above, that: T 3 (a) public health and safety and the common defense and security would be adequately protected if the proposed reduction in requirements or positions were implemented, and i (b) The cost savings attributed to the action would be substantial enough to justify taking the action. ) Not applicable. 1 (xi) 4 For each request for information under 10 CFR 50.54(f) (which is not subject to exception as discussed in III.A) an evaluation that includes j at least the following elements: (a) A problem statement that describes the need for the information in j terms of potential safety benefit. i
CRGR REVIEW PACKAGE. The NRC staff was informed during a meeting on August 24, 1995, that Westinghouse had developed a susceptibility model for VHPs based on a number of factors,. including operating temperature, years of power operation, method of fabrication of the VHP, microstructure of the VHP, and the location of the VHP on the head. Each time a plant's VHPs are inspected, the inspection results are incorporated into the model. All domestic Westinghouse PWRs have been modeled and the ranking has been I 1 given to each licensee. In addition, the NRC staff was informed that Framatome Technologies, Inc. [FTI, formerly Babcock & Wilcox (8&W)], also developed a susceptibility model for CROM penetration nozzles and other VHPs in B&W reactor vessel designs. All domestic B&W PWRs have been modeled and the ranking has been given to each B&W licensee. The NRC staff was further informed that Combustion Engineering (CE) had performed an initial susceptibility assessment for the CE PWRs. At present, neither Westinghouse, FTI, nor CE has submitted its models and assessments to the NRC staff for review. The results of domestic VHP inspections are consistent with the February 1993 analyses by the PWR Owners Groups, the NRC staff safety evaluation report dated November 19, 1993, and the PWSCC found in the CROMs in European reactors. On the basis of the results of the first five inspections of U.S. PWRs, the PWR Owner's Groups' analyses, and the European experience, the NRC staff has determined that there is a high probability that VHPs at other plants may contain similar axial cracks caused by PWSCC. Further, if any significant resin intrusions have occurred at U.S. PWRs such as occurred at Zorita, residual stresses are sufficient to cause circumferential intergranular stress corrosion cracking (IGSCC). After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern. Further, the NRC staff recognizes that the scope and timing of inspections may vary for different plants dependir") on their individual suceptibility to this form of degradation. In the long ters, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation. The vessel head provides the vital function of maintaining a reactor pressure boundary. Cracking in the VHPs has occurred and is expected to continue to occur as plants age. The NRC staff considers cracking of VHPs to be a safety concern for the long term based on the possibility of (1) exceeding the American Society of Mechanical Engineers (ASME) Code for margins if the cracks are sufficiently deep and continue to propagate during subsequent operating cycles, and (2) eliminating a layer of defense in depth for plant safety. Therefore, in order to verify that the margins required by the ASME Code, as specified in Section 50.55a of Title 10 of the Code of federal Regulations (10 CFR 50.55a) are met, that the guidance of General Design Criterion 14 of Appendix A to 10 CFR Part 50 (10 CFR Part 50, Appendix A, GDC 14) is continued to be satisfied, and to ensure that the safety significance of VHP cracking remains low, the NRC staff believes that an integrated, long-term program, which includes periodic inspections and monitoring, is necessary. In addition, the NRC 2 staff finds that the requested information is also needed to determine if the imposition of an augmented inspection program, pursuant to l
CRGR REVIEW PACKAGE 10 CFR 50.55a(g)(6)(ii), is required to maintain public health and sa fety. t The NRC staff recognizes that individual PWR licensees may wish to 1 determine their inspection activities based on an int take advantage of inspection results from other plants that have similar susceptibilities. actions but notes that such an integrated industry inspectio must have a well-founded technical basis that justifies the relationship between the plants and the planned implementation schedule. (b) The licensee actions required and the cost to develop a response to the information request. The information requested by items 1 and 2, below, is required by the NRC staff to determine if the imposition of an augmented inspection program is required, while the information requested by item 3 relates to the potential for domestic resin intrusions, such as occurred at Zorita. Addressees are required to provide the following information: 1. Regarding inspection activities: 1.1 A description of all inspections of CRDMs and other vessel head penetrations performed to the date of this generic letter, including the results of these inspections. 1.2 If you have developed a plan to periodically inspect the CRDM and other vessel head penetrations: Your schedule for first, and subsequent, inspections a. of the CRDM and other vessel head penetrations, including the technical basis for your schedule. b. Your scope for the CRDM and other vessel head penetration inspections, including whether you plan to inspect from the top or bottom of the head, the total number of penetrations (and how many will be s' sleeves, w)hich are sinspected, and which penetrations ha other penetrations. pares, and which are instrument or 3 1.3 If you have agi developed a plan to periodically inspect the CRDM and other vessel head penetrations, previde your technical or safety basis for not periodically inspecting your VHPs; or, your schedule for developing such a plan and the basis for that schedule. 2. A description of the evaluation methods and results used to assess the susceptibility of the CRDM and other VHPs in your plant to 4 PWSCC, including the susceptibility ranking of your plant and the 1 i
CRGR REVIEW PACKAGE, factors used to determine this ranking. Other than or in addition to the boric acid visual examination (see Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants," dated March 17, 1988), include a description of all relevant data and/or tests used to develop crack initiation and crack growth models, and the methods and data used to validate these models. Include a statement explaining the applicability of these raodels to the VHP cracking issue. Also you are relying on any integrated industry inspection program,, if provide a detailed description of this program. 3. A description of any resin intrusions in your plant, as described in IN 96-11, that have exceeded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information: 3.1 Were the intrusions cation, anion, or mixed bed? 3.2 What were the durations of these intrusions? 3.3 Do your RCS water chemistry Technical Specifications follow the EPRI guidelines? 3.4 Identify any RCS chemistry excursions that exceed your plant administrative limits for the followin chlorides or fluorides, oxygen, boron,g species: sul fates, and lithium. 3.5 Identify any conductivity excursions which may be indicative ) of resin intrusions, provide your technical assessment of i each excursion and your followup actions. ) 3.6 Provide your assessment of the potential for any of these intrusions to result in a significant increase in the 1 probability for IGA of VHPs and any associated plan for j inspections. All addressees are required to submit a written response with the information requested above within 90 days from the date of this letter. l Any inspection results that do agl satisfy the acceptance criteria-identified in the NRC staff's safety assessment dated November 16, 1993, should be reported to the NRC staff prior to plant restart. l The public reporting burden for this collection of information is estimated to average 80 hours per response, including the time for reviewing instructions, searching existing data sources, gathering and }i nin he data needed, and completing and reviewing the collection The cost estimated for the collection of information is estimated to i j average $8000.00 ($100/ hour expended).
CRGR REVIEW PACKAGE. (c) An anticipated schedule for NRC use of the information. The NRC staff plans to make immediate use of the requested information and will develop an action plan for future generic actions based on the information collected by this proposed generic letter. (d) A statement affirming that the request does not impose new requirements on the licensee, other than for the requested information. Because the proposed generic letter only. requests information from the PWR licensees, and the requested information has already been collected the meeting on Augustby the licensees (as stated by the PWR Own 24,1995), the proposed generic letter does not impose new requirements on the licensees, other than submission of the requested information. (xii) An assessment of how the proposed action relates to the Commission's safety Goal Policy Statement. The NRC staff feels that the proposed Generic Letter has no impact on the Commission's Safety Goal Policy Statement since it is only requesting information. i -}}