ML20196G104
| ML20196G104 | |
| Person / Time | |
|---|---|
| Issue date: | 02/13/1997 |
| From: | Tripathi R NRC |
| To: | Lainas G, Strosnider J NRC |
| Shared Package | |
| ML20196G081 | List: |
| References | |
| TAC-M95280, NUDOCS 9705130435 | |
| Download: ML20196G104 (25) | |
Text
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From:
Raji Tripathi To:
WND2.WNP6.GCL,'WND2.WNP6.JRS2 M
'a""
Date:
~/13/97 4:32)m R S"*m h.
Subject:
!GR ENDORSEiENT T0: Gus Lainas/ Jack Strosnider, NRR r
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SUBJECT:
CRGR END0RSEMENT OF THE GL DEGRADATION OF CONTROL R00 DRIVE MECHANISM l
AND OTHER VESSEL HEAD PENETRATIONS On January 28, 1997, and February 4,1997, during the 299th and 300th meetings, respectively, the Committee to Review Generic Recuirements (CRGR) reviewed the.
Subsequently, the cognizant NRR staff workec with the CRGR l
staff to revise the text of this proposed generic action to address the Committee's comments.
On the 7th of February, the CRGR staff met with the cognizant staff to review l
~the CRGR comments. As indicated in the Final Issue Sheet, dated January 27, i
1997, additional editorial comments were also provided to the staff. A couple of iterations later and following another meeting with the staff on the 12th of February to re-review the Committee's comments and the revised text, the staff developed another red-line/ strike-out version of the GL (starting with l
the 12/26 version of the proposed GL, as submitted for CRGR review),
incorporating the CRGR-comments. and using the agreed-upon " boiler-plate" material to make this GL consistent with similar other GLs. This version was received on the morning of February 13th. A couple of minor omissions were identified to the staff. Subsequently, a yet modified GL was received on the afternoon of February 13th, which appears to have addressed the Committee's L
comments.
The Acting Chairman, CRGR has accepted the CRGR staff's recommendation to endorse the attached version of the revised GL. This e-mail formally relays the CRGR endorsement of the attached version of the modified GL. Please note that the CRGR staff has not yet completed the review of the modifed CRGR Review Package to ensure that consistent and corresponding changes have also been made to this document, This item is not on the critical path; however, it is our understanding that should any changes be deemed necessary to this document, the cognizant staff would make the corrections in a timely manner.
Please also note that this version of the GL plus the January 23, 1997 letter from NEI to Dr. Ross on this GL, will be included (among other documents) as l
attachments to the minutes of the CRGR meeting No. 299 and 300. These items will be placed in the PDR after the issuance of the GL.
If you hase any questions, please contact me at 415-7584.
cc:CRGR Members Brian Sheron Charlie Haughney Jim Shapaker CC:
DFR, TWD2.TWP0. JAM 1, WND1.WNP7.CJH, TWD2.TWP8.MRK....
i 9705130435 970407 PDR REVGP NROCRGR MEETING 299 PDR; i
a o-J
! (b) to the Combined Minutes of CRGR Meetina No. 299 and 300 Red-line/ strike-out version of the Prooosed Generic Lettgr "Dearadation of Control Rod drive Mechanism and Other Vessel Head Penetration" as endorsed by the CRGR UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555-0001 q
M595$$21993 GENERIC LETTER 9Z6-##:
DEGRADATION OF CONTROL ROD DRIVE MECHANISM AND OTHER VESSEL CLOSURE HEAD PENETRATIONS (TAC NO. M95280)
Addressees All holders of operating 1.icenses for pr.essurized water; reactors (PWRs),
except those who hiveicertifiedi;tbliajpermanent?cessationtofsoperations licensc: that~ hW6"bb6n^Madd'dTd"' pod 5?sf 06~5nly^ttitEs^l~~
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Puroose The U.S. Nuclear Regulatory Commission (NRC) is issuing this generic letter to (1) request addressees to describe their program for ensuring the timely inspection of PWR control rod drive mechanism (CRDM) and other vessel closure i
head penetrations and (2) require that all addressees provide to the NRC a written res)onse to the requested information. Th6i1hf6rsstT6MMDEstidifi5 hesdsditifitid3RCBtiffst6Nshifii05iipTfiH5EWths103fR150155aishc $101W
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Primary Water Stress Corrosion Crackina of Vessel Closure Head Penetrations Most PWRs have Alloy 600 CRDM nozzle and other vessel head closure penetrations (VHPs) that extend above the reactor pressure vessel head.
The l
stainless steel housing of the CRDM is screwed and seal-welded onto the top of l
the nozzle penetration, as shown in Figure 1.
(Figure 1 is for illustrative purposes only and is not intended to be.ig.ndicative of every hiidisidsthisi i
supply 3fstemg NSSS) vendor's CRDMs des n.) The weld between the nozzle and the housing 1s a dissimilar metal weTd; which is also called a b1 metallic weld. The nozzles protrude below the vessel hcad 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).
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Generally, there are 36 to 78 nozzles distributed over the low-alloy steel l
head.
The vessel head is semi-spherical and the head penetrations are-l vertical so that the CRD nozzles and'other VHPs are not perpendicular to the l
vessel surface except at the center.
The uphill side-(toward the center of l
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 l
(3-to 4-mm) [0.12-to 0.16 in.] between the nozzle and the sleeve.
!instrum@entinozzleMat} both (domestic [agdifor61gn[5@6~?il?Allby!6004 Bsiid5f W dil986 $ 56k5ihsisibsshifsp6Fisd Wii ddopplSks"E5~5ccur~redTnce '198E~fM c(gssvefal~~ '
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diffsrerytsNSSS9'rilEijt^nocle: at both de c tic and foreign rexter:
ever5TT15y 500 p?c urucF~iH5t frc-several different nuclear stc= cup)1y vendcr-s-The NRC staff identified primary water stress corrosion cracting (PWSCC) as an emerging technical issue L
to the Commission in 1989, after cracking was noted in Alloy 600 pressurizer heater sleeve penetrations at a domestic PWR facility.
The NRC staff reviewed the safety significance of the cracking that occurred, as well as the repair L
and replacement activities at the affected facilities.
The NRC staff L
determined that the cracking was not of immediate safety significance because 1
the cracks were axial, had a low growth rate.-were in a material with an l
extremely high flaw tolerance (high fracture toughness) and, accordingly, were l
l unlikely to propagate very far. These factors also demonstrated that any L
cracking would result in detectable leakage and the opportunity to take i
corrective action before a penetration would fail.
Further, kithiths sR6ehtf6[ofstheH64k!f60ndfitiBug'eys3?dsff$'hydfd5titfBtE5tingthe NRC staff ~i'E n~6t~hsare'"of"5iif"fsiTEfb of"hW"A116y 6001siseT"Elbi6F5 head penetration during plant operation. The NRC staff issued Information Notice 1
(IN) 90-10. " Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600,"
dated February 23.-1990, to inform the nuclear ' industry of the issue.
In September 1991, cracks were found in an Alloy 600 VHP in the reactor head -
at Bugey 3 a French PWR.
Examinations in PWRs in France, Belgium. Sweden.
l Switzerland, Spain, and Japan were performed and additional VHPs with axial l
cracks were detected in several European plants. About 2 percent of the VHPs 1
l examined to date contain short, axial cracks, Close examination of the VHP that leaked at Bugey 3 revealed ver minor incipient secondary circumferential
~ anese? utilities' have taksn steps' to cracking of the VHP.
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detect'and'aitigate~the damage: - to detect the,leakene at an early ~
~- stage at most of <their plants.' European and Japanesetutilit es have inspecte'd c
l mos; of:the CRDM nozzles' and repaired the nozzles'or replaced the vessel heads l
as appropriate.
In Japan.1theThree most' wastible vessel heads are being replacedi even though'no cracks were found 'in tie;nozzlestof these heads.
In
- Franceo Elecricite de France (EdF) is planninMon replacing all vessel heads as a preventative measure.
Inservice inspect on of the upper head.is now' ~
l required in Sweden, 'and replacement of the Ringhals'2 vessel head was planned?.
Removable insulation on the vessel head and: leakage monitoring systems are instaQed at; French and Swedish plants;for earlyLoetectiortof leakage.,
l 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 plan included a review of the safety assessments by the PWR Owners Grou)s. the development of VHP mock-ups by the Electric Power Research Institute (E3RI),
i the qualification of inspectors on the VHP mock-ups by EPRI the review of 3roposed generic acceptance criteria from the Nuclear Utility Management and Resource Council (NUMARC) [now the Nuclear Energy Institute (NEI)], and VHP l
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@ Pag'E^~3Tof"11"^
l D Bd997 inspections. As part of this action plan, the NRC staff met with the Westinghouse Owners Group (WOG) on January 7, 1992, the Combustion Engineering Owners Group (CE0G) on March 25, 1992, and the Babcock & Wilcox Owners Group (B&WOG) on May 12. 1992, to discuss their respective programs for investigating PWSCC of Alloy 600 and to assess the possibility of cracking of VHPs in their respective plants since all of the plants have Alloy 600 VHPs.
Subsequently, the NRC staff asked NUMARC to coordinate future industry actions because the issue was applicable to all PWRs.
Meetings were held with NUMARC/
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NEI and the PWR Owner's Groups on the issue on August 18 and November 20, 1992. March 3. 1993. December 1, 1994, and August 24, 1995.
Summaries 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 Dwners Groups submitted safety assessments, dated February 1993, through NUMARC to the NRC on this issue. After reviewing the industry's 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 aredominately i
axial in orientation (2) the cracks would result in detectale leakage before catastrophic failure, and (3) the leakage would be detected during visual examinations performed as part of surveillance walkdown inspections before significant damage to the reactor vessel closure head would occur.
In addition, the NRC staff had concerns related to unnecessary occupational radiation exposures associated with eddy current or other forms of non-destructive examinations (NDEs), if performed manually.
Field experience in foreign countries has shown that occupational radiation exposures can be significantly reduced by using remotely controlled or automatic equipment to conduct the inspections.
In 1093, the nuclear industry developed remotely operated inservice inspection equipment and repair tools that reduced radiation exposure. Techniques and procedures developed by two vendors were successfully demonstrated in a blind qualification protocol developed and administered by the EPRI NDE Center.
In 1
the demonstrations, examinations by rotating and saber eddy current and ultrasonics showed a high probability of detection of the flaws which were also sized within reasonable uncertainty bounds.
The qualification testing also demonstrated that personnel qualified through the EPRI program can reliably detect PWSCC in CRDM nozzles.
Intearated Attack of CRDM at Zorita In 1994, circumferential intergranular attack (IGA) associated with the J-groove weld in one of the CRDM penetrations was discovered at Zorita, a Spanish reactor.
This IGA is a different degradation mechanism than the PWSCC described above.
It is believed to have resulted from the combination of ion exchange resin tielid bed intrusions, which resulted in high concentrations of l
sulfates.
Zorita~has 37 CRDM penetrations, of which 20 are active penetrations and 17 are spare penetrations.
Sixteen of the 17 s)are penetrations showed stress corrosion cracking and IGA. The cracts were both axial and circumferential.
Four of the active CRDM penetrations had significant cracking with axial and circumferential cracks.
Two cation resin
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Nge incress events occurred at Zorita.
In August 1980, 40 liters [10.57 U.S.
gaIlons]ofcationresinenteredthereactorcoolantsystem(RCS).
In September 1981, a mixed bed demineralizer screen failed and between 200 to 320 1
liters -[52.~83 to 84.54 U.S. gallons] of resin entered the RCS. The coolant conductivity remained high for at least_4 months after the ingress.
The.
increase in conductivity was attributed to locally high concentrations of i
sul fates. Sulfates were found around the crack areas and on the fracture.
surfaces.
It is important to note that sulfate cracking can occur in regions that are not subject to significant applied or residual stresses.
The.NRC staff issued IN 96-11'. " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism i
Penetrations." dated February'14. 1996, to alert addressees to the increased l
likelihood of sulfate-driven stress corrosion cracking of PWR CRDMs and other
.VHPs if demineralizer resins contaminate the RCS.
l De Westinghouse st-a# notified the WOG plants, the B&WOG plants, and the CEOG l
plants of the Zorita incident by issuing NSAL-94-028.
Westinghouse re)orted l
l that no other plant had been found worldwide that had experienced craccing l
similar to that at the Zorita plant. De Westinghouse sta# further reported l
that U.S. plants monitor RCS conductivity on a routine basis, follow the EPRI guidelines on primary water chemistry, and monitor for sulfate three times a L
week. De Westinghouse st+# concluded that no immediate safety issue is l
involved and that the conclusions in its CRDM safety evaluation remain valid.
l-4 e Westinghouse st+ # suggested that U.S. PWR plants review their RCS 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 1
NRC staff does not have sufficient information to ascertain whether any l
significant primary system resin Feid bed intrusions have occurred at any U.S.
~~
PWR l
The first U.S. inspection of VHPs took place in the-spring of 1994 at the l
Point Beach Nuclear Generating Station, and no indications were detected in any of its 49 CRDM penetrations. The eddy current inspection at the Oconee Nuclear Generating Station in the fall of 1994 revealed 20 indications in one L
Ultrasonic testing (UT) did not reveal the depth of these
' indications because they were shallow.
UT cannot accurately size defects that i
are less than one mil deep (0.03 mm). These indications may be associated with the original fabrication and may not grow: however, they will be l
reexamined during the next refueling outage. A limited examination of eight in-core instrumentation penetrations conducted at the Palisades plant found no cracking. An examination of the CRDM penetrations at the D. C. Cook plant in I
the fall of 1994 revealed three clustered indications in one penetration.
The indications were 46 mm [1.81 in.],16 mm [0.63 in.]. and 6 to 8 mm [0.24 to 0.31 in.] in length, and the deepest flaw was 6.8 mm [0.27 in.] deep. The tip of the 46-mm [1.81 in.] flaw was just below the J-groove weld.
t Virginia Electric and Power Company ins)ected North Anna Unit 1 during its L
spring 1996 refueling outage.
Some hig1-stress areas (e.g., upper and lower i
hillsides) were examined on each outer ring CRDM penetrations and no i
. indications were observed using eddy current testing
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l The NRC staff was informed during a meeting on August 24, 1995, that-Westinghouse had developed a susceptibility model for VHPs based on a numoer of factors, including o)erating temperature, years of power operation, method of fabrication of the vip. 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 j
been modeled and the ranking has been given to each licensee.
In addition,
'a the NRC staff was informed that Framatome Technologies. Inc. [FTI, formerly Babcock & Wilcox (B&W)]. also developed a susceptibility model for CRDM 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, none of the PWR Owners Groups (i.e., WOG. B&WOG, or CEOG) has i
submitted its models and assessments to the NRC staff for review.
l By letter dated March 5. 1996. NEI submitted a white paper entitled " Alloy 600 RPV Head Penetration Primary Stress Corrosion Cracking," which reviews the i
significance of PWSCC in PWR VHPs and describes how the industry is managing the issue. The program outlined in the NEI white paper is based on the assumptionthattheissueisprimidlyaneconomiceseratherthanasafety issue, and describes an economiFdsc1sion tool to 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 information would then be used by a PWR licensee to evaluate the need to conduct a VHP inspection at their plant.
The i
NRC staff informed NEI in the several meetings listed above that it did not i
agree with NEI that the issue was @i sbhi J.S. plants and the industry h JiiiiMT enly economic.
Inspection have shown that cracking h : initiated not prcvided sufficient technical justification regarding succc:tibility Of the CRD" and Other VHP: tc PWSCC to justify an inspection plan n:cd on econ 0-ic consideration: 310nc.
s Discussion The results of domestic VHP inspections are consistent with the February 1993 l
analyses by the PWR Owners Groups, the NRC staff safety evaluation 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 inspections of U.S.
i PWRs. the PWR Owner's Groups' analyses, and the Euro)ean experience, the NRC i
staff has determined that itfi Q^ ii'b r & acks caused by PWSCC.
Furtfier, if any Fobibid there i ligh probabilit" that VHPs at other plants contain siiiiiThr a 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).
l After considering this information, the NRC staff has concluded that VHP cracking does not pose an immediate or near term safety concern. Further, the i
NRC staff recognizes that the scope and timing of inspections may vary for i
i different plants depending on their individual susceptibility to this form of degradation.
In the long term, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation.
i The vessel closure head provides the vital function of maintaining reactor L
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g pressure boundary.
Cracking in the VHPs has occurred and is expected to l
continue to occur as plant.s 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 continues to believe that an integrated, long s,rm program, which includes periodic te inspections and monitoring of3 HP is necessary.
This was the conclusion of the staff's November 19, 1993',' safety evaluation, which stated, in part,
...the staff recommends that you consider enhanced leakage detection by visually examining the reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area nondestructive examinations should be performed to ensure there is no unexpected cracking in domestic PWRs.
These examinations do not have to be conducted immediately... As the surveillance walkdowns pro)osed by NUMARC are not intended for detecting small leaks, it is conceivable tlat some affected PWRs could potentially operate with small undetected leakage at CRDM/CEDM 3enetrations.
In this regard, the staff believes that it is prudent for 4UMARC to consider the imalementation of an enhanced leakage detection method for detecting small leacs during plant operation."
In addition, the NRC staff finds that the requested information is also needed to determine if the imposition of an augmented inspection 3rogram, pursuant to 10 CFR 50.55a(g)(6)(ii), is required to maintain public lealth and safety.
The NRC staff recognizes that individual PWR licensees may wish to determine their ins)ection activities based on an integrated industry inspection program 1
(i.e., B& DOG, 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 grou) actions but notes that such an integrated industry inspection 3rogram must lave a well-founded technical basis that justifies the relations 1ip between the plants and the planned implementation schedule.
Reauested Information The information requested in items 1 and 2. below. is needed by the NRC staff to verify compliance with 10 CFR 50.55a and 10 CFR Part 50. Appendix A.
GDC 14, and to determine Whsthef if the imposition of an augmented inspection program, pursuant to 10 CFR~50 55a(g)(6)(ii), is required, while the information requested in item 23 relates to the bccurfindeTof potential for domestic resin tiead intrusions in domestic PWRs, such as occurred at Zorita.
Within 120 days of the date of this generic letter, sicij addressees is' are l
requested to provide a3rittin9dpdrtithdtWncludes the following in'fdrmation
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1; Regarding inspection activities.
1.1 A descri) tion of all inspections of CRDMs and other VHP5 vc::cl closurc =d pen ^tration: gerformed to the date of thii' generic j
letter, including the resu ts of these inspections 8 1.2 If you hav^ developed a plan hisIbeenid6V615fied to periodically inspect the CRDM and other ve55^1^51550FF M5dd' penetration: VH.P.s':
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Rf6Vidsiths yew schedule for first, and subsequent inspections of~the'~CRDM and other VHPi vc :cl c10:ure head ponctrations, t
including the technical ~bssis for t$5 yew schedule.
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Pf6Vfdsiths Vew scope for the CRDM and other VHR vc :cl cle:urc inYdl556thtion inspections, including the total number of penetrations (and how many will be inspected), aM which l
penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
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pg3dW relying on my integrated industry in:pection progr=. provide detailed dcccription of this progr=
- 23. PMidi a description of any resin b5R intrusions in your :l=t, as LdisEFibed.in IN 96-11, that have exds~eded the current EPRI
)WR Primary
~
Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
f 23.1 Were the intrusions cation, an 6. or mixed bed?
i 23.2 What were-the durations of these intrusions?
23.3 Dod5?thsiplantis 00 your RCS water chemistry Technical Specifications fo1T5Fthe"EPRI'^ guidelines?
26.4 Identify any RCS chemistry excursions that exceed ths yew plant administrative limits for the following species:
sulfates, chlorides l_
or fluorides, oxygen, boron. and lithium.
23.5 Identify any conductivity excursions which may be indicative of resin
~'
intrusions M Pf6 vide?5.
rovide your technical assessment of each excursionhnd~an[yssefollowupactions.
23.6 Provide hii yew assessment of the potential for any of these
~
intrusiori5 to result in a significant increase in the probability for
~
IGA of VHPs and any associated plan for inspections.
Recuired Resoonse WithlW730TMyi?5fithiidatiT6EthisDyneffB16ttiFMMEREsddFiiinisWiffE4si fsd td%dibitsaiwittenWispdnseMndica';iiig$id l
infonnetlorGrillsbsisubmittedi&ndg2nwhethed)1whethef166notitheP^"ted"^~o is1111beisubmittediwithinithekequest6ditiakiperOdKAddresseebh65choosemot l
9 tofsubmitsth6? regnestedsinfoPastionRopstelunableitoisatisfyltheiriquested' iaction[thstMs!)naustidiscritielirntheir$responselan' y!altersethetcourse complettonidate RfAthejprpposedfahefativescoursegtl action i
i IsritiffrW111rF6VfWth6It6spohibittRthWniEasttssiddNffd6H6aE6Eli~fi
. _identi_fh_di._saffe_cted_isd_dres_ sees _uvill._ibe_inot._t edr~~~~^~~~~ "~~~~ ""
~m l-
,.11 addre::cc: chall submit in writing th^ inf-semation identified above within
^
l 120 day; fr= th^ date of this letter l
^ y in :cction result that d0 r,d Ostisfy the accept =cc,n_n,iteri; identified cr u,,, + wm
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reported to the FRC staff prior to pl=t restart.
t i
Address the-required written reports to the U.S. Nuclear Regulatory Commission. ATTN:
Document Control Desk. Washington. D.C. 20555. under oath or affirmation under the provisions of Section 182a. Atomic Energy Act of l
l
}
o GL 976-#
A iffl @ @"f l
l997 Pagi'3F 6f ^11~*
1 i
1954, as amended, and 10 CFR 50.54(f).
In addition, submit a copy to the appropriate regional administrator.
e The NRC recognizes the potential difficulties (number and types of sources, j
age of records, proprietary data, etc.) that. licensees may encounter while l
ascertaining whether they have all of the data pertinent to the evaluation of their CRDMs and other VHP3 vc :cl closure he:d penetrations..For this reason, l
the above time periodsTs allowed for.the responses.
Related Generic Communications I
i (1)
Information Notice 90-10. " Primary Water Stress Corrosion Cracking (PWSCC) of Inconel 600," dated February 23, 1990.
1 (2)
NUREG/CR-6245, " Assessment of Pressurized Water Reactor Control Rod l
Drive Mechanism Nozzle Cracking," dated October 1994.
i (3).
Information Notice 96-11. " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations," dated February 14, 1996.
Backfit Discussion under the" provisions"of Secti6n 182a~of ~the' AtdmicEnehgyWt' 6f~1954."'is amended, and 10 CFR 50.54(f) this generic letter transaits'an infomation request:for, the purpose of verifying compliance with applicable' existing ~ bli regulatory requirements : Specifica ly, the requested'inforestion would ena the NRC r;aff to^ determine tdiether or not the Ticensees" margins l
the ASME* Code.' as,specified in Section 50'55a of Title'10 of1the; required by" cose of' Federal Aegulatfons,(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.r^ ~
Appendix A, GDC 14) continues to be satisfied. 'and to ensurithat ths sfet9 significance of VHP cracking remains 10w.' 1he requested;taformation is also needed to,detemine whether an augmented insaction program, ' mrsuant' to 10CFRiS0.55a(g)(6)(iijuis required to main;ain pubEc healt1 and safety":
AdditibKillyRh67tkiickfitMiThithiEfitterided76H7hppc6Hd3hith6Tcontaktibf eneri ci6w~.-.e. wh h.a.sx.. wwu d,y:n *e..*l,.stamaxvc is'suancel.o. f#t. h. isi'gmw.a,4;v~.sv lett rNT ereforeMth fflha.5%,.<w,t4p.wmm.< %.s.- -ns ehform.dls bactDten.:a,1ys1Q mw c
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T"i generic letter only reg':1re: infer = tion fro-' the addresscc: under the arovision: cf Section 182: cf the.^tomic Energy.^.ct of 195'. :: : mended, and V.n c.r. o. e.n.. c a. m.,. r, u m - m. <. m,,-m..
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The infor= tion collected will enable the staff to verify that the =rgins f^^uired by the ^SME Code, :: Specified in Section 50.55: Of Title 10 of the Code c,f Fedcrs? 9cgulatfc=,a m(10 CFP, 50.553) ~c =t that the, guidance of cm-m, n m - 4,m m,, c-4.+m 4.m..
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significance Of '!HP cracking rc=in: 10w Siee^ the " C staff require l
licen:ce; to sub-'it infer = tion to ::cc: compliance wit" the above :tsted require ents, a justification for this rc^uested infor= tion need not bc
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e GL 976-##
April ~ ? 1997.
Page'39 of 11' requested infor= tion B sho needed to determine if the imposit40n of an augmented inspection ^rogram. pursuant to 10 CFR 50.55 (g)(5)(11), h required te =intain public hcElth and sfety.
The staff B not 0;tablishing a new
^00ition for such co"'plunce in thB generic let-ter Therefore. th u generic
. Tetter doc: not constitute a backfit and no documented evaluation or backfit analy:B need be prepared.
Federal Reaister Notification A notice of opportunity'for public comment was published in the Federal Register (61 FR 40253) on August 1, 1996, and extended on August 22, 1996 (61 FR 43393).
Comments were received from seven licensees, two industry organizations, and one Code Committee.
Copies of the staff evaluation of these comments have been made available in the public document room.
Paoerwork 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 collections were ap3 roved by the Office of Managenient and Budget, approval number 3150-0011, w1ich expires July 31, 1997.
The public reporting burden for this collection of information is estimated to average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
The U.S. Nuclear Regulatory Commission is seeking public comment 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 3 roper performance of the functions of the NRC, including whether tie information will have practical utility?
2.
Is the estimate of burden accurate?
l 3.
Is there a way to enhance the quality, utility, and clarity of the information to be collected?
l 4.
How can the burden of the collection of information be minimized, including the use of automated collection techniques?
Send comments on any aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch, T-6 F33, U.S. Nuclear Regulatory Commission. Washington, DC 20555-0001, and to the Desk Officer, Office of Information and Regulatory Affairs NE0B-10202 (3150-0011), Office of Management and Budget. Washington, DC 20503.
The NRC may not conduct or sponsor, and a person is not required to respond l.
to, a collection of information unless it displays a currently valid OMB control number.
I
April"Q'$f 11' GL 976-i
- 1997 Pags'4U~
1 If you'have any questions 'about this matter, please contact one of the technical contacts -listed below or.the appropriate Office of Nuclear Reactor Regulation (NRR) project manager.
l Thomas T. Martin Director Division of Reactor Program Management Office of Nuclear Reactor Regulation Technical contacts:
Keith R. Wichman (301) 415-2757 e-mail: krw@nrc. gov James Medoff i
(301) 415-2715 e-mail: jxm@nrc. gov Lead Project Manager:
C. E. Carpenter. Jr.
i (301):415-2169 I
e-mail: cec @nrc. gov Attachments:
1.
References 12.
List of Recently Issued NRC Generic Letters I
i i
e i
f s
a
> ?
I
% (d) to the Combined Minutes of CRGR Meeting No. 299 and 300 Red-line/ strike-out version of the CRGR Review Package for Proposed Generic Letter " Degradation of Control Rod drive Mechanism and Other Vessel Head Penetration" revised by the staff to make it-consistent with the generic letter as endorsed by the CRGR.
CRGR REVIEW PACKAGE PROPOSED ACTION:
Issue a generic letter.on the deifiditl6ii primary water strc: corrction cracking of c'6nifo1~F6d" drive mechanism and other vessel head penetrations.
' CATEGORY:
2 RESPONSE TO REQUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW (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 proposed generic requirement or' staff position can be achieved by setting a readily quantifiable standard that has an unambiguous relationship to a readily measurable quantity and is enforceable, the proposed requirement should merely specify the objective or result to be l
attained, rather than prescribing to the licensee how the objective or result is to be attained.~
.The information requested in items 1 and 2. below, is needed by the NRC l
staff to verify compliance with 10 CFR 50.55a and 10 CFR Part 50.
Appendix A. GDC 14. and to determine Midthed if the 1 ^0 ition # an j
augmented inspection program, pursuant ~to"10' CFR 50.55a(g)(6)(ii). is required, while the information requested in item 23 relates to the McurFbndEidfpotentialfordecsticresinbiadintrusionsindomestic l
PWRs.~ isch is occurred at Zorita.
" ~ ~ ~
l Within 120 days of the date of this generic letter, bildi addressees is ar-e requested to provide hTyHt46]~ rep ~6ftithit@M~Udes~the followiny" informationf6MfsjfipjlipJ~~
i
~~~~
~~
u 1
l 1.
Regarding. inspection activities:
1 1
1.1 A descriation of all inspections of CRDMs and other VHPf ve :cl clecureiced^cnetratiens$erformedtothedateofthis" generic letter incluEling the resu ts of these inspections 3 IfyouhavedevelopedaplanhisWshdWil to periodically inspect the CRDM and other vees 5Fc16]:5?5~'hpl~ penetration:
1.2 M5:
a.
PFosidsithi yew schedule for first, and subsequent.
idipicti6nsoftheCRDM'andotherVHPsvc:clc10:ureScad penetrations, including the techniEiTbasis for this~ yew schedule.
I b.
P'f69Tdifths Vo w scope for the CRDM and other VHP vessel Oddi??fh~clid penetration ins)ections, includin@*The total number of penetrations (and low many will be inspected).
4 4
need;M, iresubeW@it?butim6Bhsteadfreferencelthe7ap'prLopd~ ate-
' ~ ~ ~
C00CespondenceMM gresponsgtOhikGenerggetwjr i
CRGR REVIEW PACKAGE "
aM which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
1.3 If you have not develo cd a 31an hihiWbi6h~fdd9hl6pid to seriodically inspect t1e CRDi and*6ther VHPs75553Fc15:ure
- ^;d p.cnc*rgt10n';._p.rovide thEisnajjiMthat))pyiT5@hptyij augmentsddnsp#5PibdiESTTf"ilif;55cting your "HP ; Or. your ectionRisinecessary? your techmca. Or
- :..
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schedul sc.~nA..,e for dcvc'Oping such a phn and the bric for that u_..
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m cMnputsstolthese;modelsHndshow:these; mode s substahtistelthetsssceptibl11tfWVa10ati6fiMissinffsh providesalde..t_ai_l'_7' dest.~pectiondrogramjistbeingi!! relie integratsdRindust 1ns
~~
- riptiohlofsthis. ipr 6.grami~ ~ ~ ~ ~
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9.
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factor used to deter-ine this ranming. Other t. :n or in addition to the b0ric acid visual c=ination (500 Cercric Letter 88 05. B0rie i
Acid Corr,osion of Carbon Steel Rea,ctor Prc ure Boundary Co ponents i
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- 23. Pf6Mdd a description of any resin SEid intrusions in your mnt, as dssdFibed in IN 96-11. that have exEE6ded the current EPRI
)WR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels. including the following information:
23.1 Were the intrusior.s cation, anion, or mixed bed?
23.2 What were the durations of these intrusions?
23.3 D6eilthylihtM 00 your RCS water chemistry Technical Spscificatiohs~ follow the EPRI guidelines?
~
23.4 Identify any RCS chemistry excursions that exceed tbs yee plant administrative limits for the following speciEsIithium.
sulfates, chlorides or fluorides, oxygen, boron and
^
CRGR REVIEW PACKAGE j 26.5 Identify any conductivity excursions which may be indicative of resin intrusionshiPr6VideEd. provide youe technical assessment ofeachexcursion~arkiany'9saefollowupactions.
23.6 Provide sd yeae 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 Committee member may request CRGR 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)
NUREG/CR-6245 " Assessment of Pressurized Water Reactor Control Rod Drive Mechanism Nozzle Cracking." dated October 1994.
(4)
Information Notice 96-11. " Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations." dated February 14, 1996.
(iii) Each proposed requirement or staff position shall contain the sponsoring office s position as to whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce existing requirements or staff positions.
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 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 ex)erience. the NRC staff has determined that ItIisihfobsbld there is a ligh probability that VHPs it other plants c6ntsin simitar axial cracks.
Further., if any significant resin intrusions have occurred at U.S. PWRs sucn 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 depending on their individual susceptibility to this form of degradation.
In the long term, however, degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation.
The vessel closure head provides the vital function of maintaining reactor pressure boundary.
1
9 CRGR REVIEW PACKAGE '
Cracking in the VHPs has occurred and is expected to continue to occur e
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 de)th for plant safety.
Therefore. to verify that the margins required )y 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 continues to believe that an integrated, long-term program, which includes periodic inspections and monitoring of VHPs. is necessary.
This was the ccoclusion of the staff's November 19. 1993.
safety evaluation, which stcted. in part. "...the staff recommends that you consider enhanced leakage detection by visually examining the
{
reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area... noadestructive examinations should be Jerformed to ensure there is no unexpected cracking in domestic PWRs. T1ese examinations do not have to be conducted immediately... As the surveillance walkdowns proposed by NUMARC are not intended for detecting small leaks, it is conceivable that some affected PWRs could potentially operate with small undetected leakage at CRDM/CEDM penetrations.
In this regard, the staff believes that it is prudent for NUMARC to consider the implementation of an enhanced -leakage detection method for detecting small leaks < during l
plant operation."
In addition, 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.
l The NRC staff recognizes that individual PWR licensees may wish to i
determine their inspection activities based on an integrated industry inspection program (i.e.. B&WOG CE0G. WOG. or some subset thereof) to i
take advantage of inspection results from other plants that have similar susceptibilities. The NRC staff.does not with tc discourage such group actions but notes that such an integrated industry inspection program must have a well-founded technical basis that justifies the relationship between the plants and the planned implementation schedule.
(iv) The proposed method of implementation with th3 concurrence (and any comments) of OGC on the method proposed. The concurrence of affected program offices or an explanation of any nonconcurrences.
See attached concurrence page.
(v)
Regulatory analyses conforming to the directives and guidance of NUREG/BR-0058 and NUREG/CR 3568.
(This does not apply for backfits that ensure compliance or ensure define. o: redefine adequate protection.
i
CRGR REVIEW PACKAGE In these cases a documented evaluation is-required as discussed in IV.B.(ix).)
Not applicable (vi)
Identification of the category of reactor plants to which the generic requirement or staff position is to apply (that is, whether it is to apply to new plants only, new OLs only, OLs after a certain date, OLs before a certain date, all OLs, all plants under construction, all i
plants, all water reactors, all PWRs only, some vendor types, some vintage types such as BWR 6 and 4 jet pump and nonjet pump plants, i
etc.).
All holders of operating licenses for pressurized water reactors (PWRs).
except those Gih6thsVsidertifisd?t6?sTheMhentFde55#tT6676f@~sF5tibh5 l
1icence that"E30F~SEh*5fnd6dt635565M5h~~5nW^btstBEl~
~ ~ ~ ~ " ~
(vii) For backfits other than compliance or adequate protection backfits, a backfit analysis as defined in 10 CFR 50.109..The backfit analysis shall include, for each category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The backfit analysis shall document for consideration information available concerning any of the 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.
F (c)
Potential change in the risk to the public from the accidental release of radioactive material:
Not applicable.
(d)
Potential impact on radiological exposure of facility employees and other onsite workers:
Not applicable.
(e)
Installation and continuing costs associated with the action, including the cost of facility downtime or the cost of j
construction delay:
Not applicable.
l l
CRGR REVIEW PACKAGE *
)
i (f)
The potential safety impact of changes in plant or operational complexity, including the relationship of proposed and existing i
regulatory requirements and staff positions; j
Not applicable.
(g)
The estimated resource burden.on the NRC associated with the 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 interim or final, and if interim, the justification for imposing the proposed action on an interim
- basis, i
Not applicable.
(j)
How the action should be prioritized and scheduled in light of other ongoing regulatory activities. The following information ma he appropriate in this regard:
1.
The proposed priority or schedule, 2.
A summary of the current backlog of existing requirements awaiting implementation,
{
3.
An assessment of whether implementation of existing requirements should be deferred as a result, and 4.
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 50.109(a)(2) (i.e.,
not adequate protection backfits and not compliance backfits), the 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 public health and safety or the common defense and security to be derived from the proposal: and (b)
The direct and indirect costs of implementation, for the facilities affected, are justified in view of this increased protection.
Not applicable.
t
CRGR REVIEW PACKAGE (ix) For adequate protection or compliance backfits evaluated pursuant to 10 CFR 50.109(a)(4)
(a) a documented evaluation consisting of:
(1) the objectives of the modification (2) the reasons for the modification (3) the basis for invoking the compliance or adequate protection exemption.
(b) in addition, for actions that were immediately effective (and theiBre issued without prior CRGR review as discussed in III.C) the evaw-Hon shall document the safety significance and appropriateness nf the action taken and (if applicable) 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 the PWR licensees: rather. the proposed generic letter is requesting the PWR licensees to ]rovide to the NRC information that the PWROGs has already told the 4RC staff it has gathered, but has not shared with the NRC to date.
(x)
For each evaluation conducted for proposed relaxations or decreases in current requirements or staff )ositions, the pro)osing Office Director's determination, together with t1e rationale for t1e determination based on the considerations or paragraphs (1) through (vii) above, that:
(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 (b)
The cost savings attributed to the action would be substantial enough to justify taking the action.
Not applicable.
(xi) 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 at least the following elements:
(a)
A problem statement that describes the need for the information in j
terms of potential safety benefit.
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 l
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CRGR REVIEW PACKAGE given to each licensee.
In addition, the NRC staff was informed that Framatome Technologies. Inc. [FTI. formerly Babcock & Wilcox (B&W)].
also developed a susceptibility model for CRDM 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, none of the PWR Owners Groups (i.e.. WOG. B&WOG or CEOG) 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 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 ex)erience, the NRC staff has determined that itiisiprobabls there is a ligh probability that VHPs at other plants c6nt~ain~siihilsF axial cracks.
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 depending on their individual susceptibility to this form of degradation.
In the long term, however.
degradation of the CRDM and other VHPs is an important safety consideration that warrants further evaluation.
The vessel closure 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
)lant safety. Therefore, in order to verify that the margins required
)y 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 t
(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 continues to believe that an integrated long-term program, which includes periodic inspections and monitoring is necessary. This was the conclusion of the staff's November 19. 1993, safety evaluation.
I which stated, in part. "...the staff recommends that you consider I
enhanced leakage detection by visually examining the reactor vessel head until either inspections have been completed showing absence of cracking or on-line leakage detection is installed in the head area...
nondestructive examinations should be performed to ensure there is no unexpected cracking in domestic PWRs.
These examinations do not have to be conducted immediately... As the surveillance walkdowns proposed by NUMARC are not intended for detecting small leaks, it is conceivable
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.CRGR REVIEW PACKAGE
- that some affected PWRs could potentially operate with small undetected leakage at CRDM/CEDM penetrations.
In this regard, the staff believes that it is prudent for NUMARC to consider the implementation of an enhanced-leakage detection method for detecting small leaks during plant operation." In addition, 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.
Therefore, this request does not increase nor reduce existing requirements.
It is a request to obtain information to confirm compliance with existing requirements.
The NRC staff recognizes -that individual PWR licensees may wish to 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-te discourage such group actions but notes that such an integrated industry inspection program 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 in items 1 and 2. below, is needed by the NRC staff to verify compliance with 10 CFR 50.55a and 10 CFR Part 50.
Appendix A. GDC 14. and to determine Wiethsi4 if the i=^0cition of an augmented inspection program, pursuant ^~t6"10 CFR 50.55a(g)(6)(11), is required, while the information requested in item 23 relates to the 6bcdffsricsI6f potential for do estic resin SEid intrusions in do c: tic PWRs. such as occurred at Zorita.
Within 17
.f the date of this generic letter, sidfi addressees is are reque.
provide 65ifitt information Igf3t5Hsbi]itF'W'~~ rep ~6ht][thit!3ndlUdes~the followiny^~
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1.
Regarding inspection activities:
1.1 A descri) tion of all inspections of CRDMs and other VHPs vc :cl closure iced ^cnctration: performed to the date of thii~ generic letter, including the results of these inspections 4 1.2 If you-have developed a plan hisibsshidsVsW5sd to periodically inspect the CRDM and other ve:55T~dT6sUf5~152' penetration:
VHPs:
a.
Pf6VfdsRhs yee schedule for first, and subsequent.
Id5pectibns of the CRDM and other VHPi vc :cl 10:ure head Th6ssMibenesssWEhWEihiiiVsipiisEduilyy subslEEsi1REhs?fs40^sWEsa infsmationRnesdinetE@yyckpn$}1tWsQmapEissteadl&efeVendefths sshbmit spp$chMjpy[c$$N tjs @fges M sj h Qf % g psfid
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CRGR REVIEW PACKAGE
- penetrat4 ens, including the technical basis for thM yee schedule.
b.
PFbVideiths Ve w scope for the CRDM and other VHP ve::cl 5155HF5'"h55d penetration ins)ections, includind"the total number of penetrations (and low many will be inspected),
and which penetrations have thermal sleeves, which are spares, and which are instrument or other penetrations.
i 1.3 If you have not devc10)cd a plan lii5MbEshTdsVilbpEd to seriodically inspect t1e CRDM and"other VHPHE5bTc15:urc 1 cad ^cnetrations, provide thi7h'n#1y5Tsith6tiid "^MiiWhyEh6 m_gment,_edu.nspectiondsi,ne,cessary; year-te,g.
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n II4"In'119 t'of the'degradatio'n of'CRDM and 'other VHPs 'desc^ribed h
~ 'above, provide the analysis that supports the' selected course j
of, action as listed in either l'.2'or 1.3, above. / In ~
i particular, provide a< description of all relevant dataland/or tests used to develop crack initiation and crack growth models';
the methods and data used to validate these models the' lant-l l
specific inputs to these models, and how these models" p~ '^
substantiate'the susceptibility evaluation. Also, if an integrated' industry inspection program is being relied on provide, a, detailed descriptionjof this programj 2.
A description o,4.+,,,f the eva',uaticr method: and results used,,to :5000
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PWSCC including the susceptibility ranking of your phnt and the factor: used to determine thi; ranking. Other than Or in addition to the bcric acid visual examination (500 Cencric Letter 88 05 Boric Acid Corr 4, no..n, n,osion of Carbon Stec' Rea,noo,sctor Prc ure Boundary Co ponent a_na, u, ~ s,
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relevant dat: and/
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=0dc's to the "HP cracking issue. Ah0 if you are relying On any i
integrated industry inspection program, provide detailed dcccription of this progrcm i
- 26. PF6Mds a description of any resin bsid intrusions in your p' ant, as dei 6Fibed in IN 96-11. that have exEEEded the current EPRI PWR Primary Water Chemistry Guidelines recommendations for primary water sulfate levels, including the following information:
26.1 Were the intrusions cation, anion, or mixed bed?
26.2 What were the durations of these intrusions?
23.3 06s5ftheTplaiit[5 00 your RCS water chemistry Technical
~
S$cifiEhtions"followtheEPRIguidelines?
a CRGR REVIEW PACKAGE 23.4 Identify any RCS chemistry excursions that exceed Efii yeae plant administrative limits for the following specie ~5:
sulfates, chlorides or fluorides, oxygen, boron, and lithium.
23.5 Identify any conductivity excursions which may be indicative of of each excursio @n and any yeaf followup actions.F6Vidijivprov resin intrusions 26.6 Provide sd yeae assessment of the potential for any of these
^'
intrusions to result in a significant increase in the probability for IGA of VHPs and any associated plan for i
inspections.
Withihi!30Td595?bf2thsidat#1bfIthiMyEnbhiW1sttsfRiibhFiddFds5EEffi requ frsditoj subsni ttaiWittsn fresponsefi ndidatingiGUMiethsE66h6t theirequestediinfohmat16niWilhbeMubmittedMd1(2Mthe$orinotsth#
tsquest.edlinformationMlhbsisubijittted;Withintthelre$sstediti.me ~~ ~
perfod MAddresseesavhoichooselnotitoisubmittthe> requestbd ji jjformati oR.
ordare iuhabl eitoDati sfylthss@0s$tedicompl etiohidateWmustEdesbrib6%1ij i
i thsi r@e'sponssf anytalternati veicoufselbffacti onsthatsisiptoposedstoi bs'~
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eltetnativegmoseafactis@he3?eptabjl1}j{oMhelproposs'~~ ~
NRCf5ta ffIsilKrbViEGIthEIF&sp6 hie 57t67thfirs6HsFiED stWW6d]~f posernsgelidentMisdgiffecteHaddrjsgeesgilRbghotgledf
.11 addrc :cc: Chall submit in 'criting the information identified above
^
within 120 day; from the date cf thic letter Ano 4nenne&4nn enciel+c
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'1. _. 4.".1. C3. ' C..."+ L. ~3""..a. ~s" J..s'7 7.. ""
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nno m
Chculd be reported to the NRC st:ff prior te plant restart.
The public reporting burden for this collection of information is estimated to average 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and l
maintaining the data needed, and completing and reviewing the collection of information, t
The cost estimated for the collection of information is estimated to l
average 58000.00 ($100/ hour expended).
l (c)
An anticipated schedule for NRC use of the information.
The NRC staff plans to make immediate use of the requested information to verify that licensees are monitoring vessel head penetration cracking so as to provide reasonable assurance that existing regulations are being satisfied and to determine if augmented inspection rules need to be developed.
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i (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 by the licensees (as stated by the PWR Owners Groups to the NRC during the meeting on August 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.
I The NRC staff feels that the proposed Generic Letter has no impact on the Commission's Safety Goal Policy Sta.tement since it is only requesting information.
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