U-602369, Forwards Amend to Util 940924 Response to GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs

Forwards Amend to Util 940924 Response to GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs

ML20080C986
Person / Time
Site:	Clinton
Issue date:	12/14/1994
From:	Jackie Cook ILLINOIS POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-94-03, GL-94-3, JGC-394-94, L30-94(12-14)LP, U-602369, NUDOCS 9412220134
Download: ML20080C986 (13)

Text

Ithnois Power Company 1

Chnton Power Station P o Box 678 Chnton, IL 61727 l

Tel 217 935 5623 Fax 217 935-4632 John G. Cook Vce President P6WER U402369 L30-94( 12-14 )LP 80.120 JGC-394-94 i

Docket No. 50-461 December 14, 1994 l

i Document Control Desk Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Clinton Power Station: Revised Response to Generic Letter (GL) 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors"

Dear Sir:

The purpose of this letter is to provide an amendment to Illinois Power's (IP's) original response to the Nuclear Regulatory Commission's (NRC's) request for information under Generic Letter 94-03.

The information contained within Attachment 2 revises the information originally submitted for the generic letter in IP letter U-602364, dated September 24,1994, for:

(a)

A schedule for inspection of the core shroud, (b)

A safety analysis supporting continued operation untilinspections are conducted.

Revision bars are included indicating where changes to the original U-602364 letter have been made. Drawings of the core shroud configuration showing details of the geometry, and the history of core shroud inspections including date, scope, methods and results are unchanged and not resubmitted.

This revision is based upon the "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" presented to the NRC during the September 21,1994 meeting with members of the BWR Vessel and Internal Project. Those guidelines place ple.nts in one of three categories. CPS is a category "A" plant and will not approach the threshold of being a category "B" plant until the sixth refueling outage (RF-6) scheduled for October 1996. Therefore, appropriate core shroud inspections will be performed during RF-6 rather than during RF-5 which is scheduled to begin in March 1995.

8' 9412220134 941214 l

ADOCK0500g1 PDR

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U-602369 Page 2 i

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Attachment I provides an affidavit supporting the facts set forth in this letter and l

its attachments.

Sincerely yours, i

i hv G. Cook Vice President i

JSP/csm l

Attachments cc:

NRC Clinton Licensing Project Manager NRC Resident Office, V-690 j

Regional Administrator, Region III, USNRC i

Illinois Department of Nuclear Safety i

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Attachment I to U 602369 J. A. Miller, being first duly sworn, deposes and says: That he is Manager-Nuclear Station Engineering Department of the Nuclear Program at Illinois Power (IP), Clinton Power Station (CPS); that this letter supplying IP's amended response to Generic Letter 94-03 has been prepared under his supervision and direction; that he knows the contents thereof; and that to the best of his knowledge and belief said letter and the facts contained therein are true and correct.

Date: This /Y day ofDecember 1994.

Signed:

d4Mt.Ydtal h

J. A. Miller STATE OF ILLINOIS l SS.

• 0FFEM.8ER*

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g s.mma 361w-COUNTY J

g Subscribed and sworn to before me this 14^ day of December 1994.

/awfia d1'Yken M ]/flotany Public) '

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Page1 of10 l

IP's Amended Response to GL 94-03 Thirty-Day 1

Beporting Requirements and Requested Licensee Actions l

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NRC Reporting Requirement 1.(a): A schedule for inspection of the core shroud.

Requested Licensee Action #1 requests that the Licensee " Inspect the core shroud in their BWR plant no later than the next scheduled refueling outage "

IP Response:

The core shroud at CPS will be inspected during the sixth refueling outage, currently scheduled to start in September 1996. This inspection plan considers recommendations from the BWR Vessel and Internals Project (VIP) Assessment Committee. Their guidance contained in General Electric - Nuclear Energy report GENE-523-113-0894 entitled "BWR Core Shroud Inspection and Flaw Evaluation Guidelines" [ GENE 0894] recommends the extent of inspection required to ensure continued safe operations, based upon three susceptibility factors of hot operating time, coolant conductivity, and core shroud material. Review of these parameters for CPS shcws that CPS is a Category A plant; thus immediate inspection is not required. Category B status is expected to be reached during the seventh fuel cycle, based upon reaching the eight hot operating years criterion. Therefore, appropriate core shroud inspections will be performed daring RF-6.

NRC Reporting Requirement 1.(b): A safety analyses, including a plant specific safety 2.

assessment, as appropriate, supporting continued operation of the facility until inspections are conducted. Additionally, Requested Licensee Action #2 requests the following from the Licensee:

Perform a safety analysis supporting continued operation of the facility until inspections are conducted. The safety analysis should consider, but not be limited to the following factors:

Details of the conditions that would miiuence the probability of occurrence of a.

the cracking and rate of crack growth (e. g., material types sad forms, water chemistry, fluence, carbon contents, welding materials and crucedures).

b.

A plant specific assessment accounting for the uncertainties in the amount of l

cracking, which should include but not be limited to the following:

(1)

An assessment of the shroud response to the structural loadings resulting from design basis events. If asymmetrical loads can affect the shroud response, these should also be considered.

1 Gf An assessment of the ability of plant safety features to perform their i

function considering the shroud response to structural loadings.

Att:chment 2 to U-602369 Page 2 of 10 i

IP Response:

Analysis Approach The original CPS response reasoned that as a BWR-6 CPS should demonstrate the absence of safety significant crack (s). " Safety significant cracks" were defied as crack (s) large enough to degrade the shroud's ability to function as designed, under all normal. off-normal and accident conditions, by reduction ofits structural margin. GENE 0894 places plants into one of three categories when considering the likelihood of any cracking. This letter compares the values of five stress corrosion cracking parameters of BWRs inspected to those values at CPS, going beyond the three parameters GENE 0894 uses. This comparison demonstrates that in addition to clearly being a Category A plant through the planned shutdown for RF-6, the likelihood of safety significant crack (s)in the CPS shroud is too remote to be a threat to safe plant operations before inspection in RF-6.

The following is an assessment of stress corrosion cracking parameters including those identified in the generic letter, which may influence the probability ofinitiating cracks or the rate of crack growth resulting from stress corrosion cracking in the core shroud and attachment of the shroud to the reactor pressure vessel. In letter dated August 5,1994, BWROG-94097, " Revision 1 of BWR Shroud Document" from R. A. Pinelli, Chairman BWROG, August 5,1994, [BWROG-94097] the BWROG provides the source of these parameters for comparison to other BWRs. In addition, BWROG-94100, " Response to NRC Questions on Core Shroud and Reactor Internals" from R. A. Pinelli, Chairman BWROG, August 5,1994, [BWROG-94100], was used for both comparison and reference.

Shroud parameters are grouped into the following categories for assessment and comparison:

A.

Materials of Construction B.

Reactor Water Chemistry C.

Neutron Fluence D.

Fabrication E.

Hot Open Sg Time Assessment of thm parameters at CPS is as follows:

1 A. Materials ofConstruction.

l The materials used to construct the core shroud geometry at CPS are shown in Table 1.

This table is a compilation ofinformation contained in on-site records and in the Shroud Composite Drawing (Attachment 3 of the originalletter) The horizontal rings of the shroud (pieces No. 3 and 8) were cut frnm plate as indicated in BWROG-94100.

The shroud support assembly material idw

- ' as the support legs in GL 94-03, Fgure 1, is Inconel.

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l to U-602369 Page 3 of10 On-site records are copies of documents provided by the manufacturer of these components, Sun Shipbuilding and Dry Dock Company, to General Electric (GE), as l

required by purchase specifications. The records of Chicago Bridge and Iron Nuclear l

and Reactor Controls Incorporated, which made the reactor pressure vessel including the shroud suppon assembly (pieces No.10 through 12) and performed on-site assembly, are not on site. GE purchase specification 21 A9496 required that manufacturing of the principle components (shroud, core plate, top guide and the shrot, support assembly) be in accordance with subsection NG of the ASME Section l

III, while the balance of the components use the ASME Code as a guide.

Table 1 - CPS Core Shroud Material of Construction Drawing Manufacturer BWROG 94100 Plate Carbon GE Piece Piece Name Piece Name Material Content Purchase No.

Specifications 1

Grid Flange Top Guide 304L 0.020 21 A9496.4.9 Head Flange 2

Grid Shell Top Guide 304L 0.022 21 A9496.4.9 Cylinder 3

Grid Plate Top Guide 304L 0.014 21 A9496.4.9 Flange 4

Top Flange Ring Shroud Head 304L 0.017 21 A9496.4.9 Flange 5

Upper Shroud Shell Central Upper 304L 0.024 21 A9496.4.9 l

Cslinder 6

Intermediate Cc itral Middle 304L 0.023 21 A9496.4.9 Shroud Shell Plate Cylin.:er i

7 Lower Central Lower 304L 0.019 21 A9496.4.9 j

Shroud Shell Plate Cylinder 8

Lower Flange Ring Core Plate 304L 0.025 21 A9496.4.9 Flange 9

Bottom Shroud Lower 304L 0.019 21 A9496.4.9 Shell Plate Cylinder 10 Shroud Support Shroud Support inconel 21A9477 Baffic Plate Bame Plate (1) 11 Shroud Support Shroud Support inconel 21A9477 Cylinder C31inder (1) 12 Shroud Support Shroud Support Inconel 21A9477 Leg (s)

Leg (s)

(1) i Table i Note:

GE vessel purchase specification 21 A9477 R/9 of 12/19/76. indicates the material is

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nickel-chrome-iron, ASME SB-168 or SB-166 with weld material to meet ASME SFA 5.14 or 5.11.

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• Information is not available on the carbon content of piece number 10, II and 12.

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The maximum carbon content in a plate used to construct the core shroud geometry (pieces No. I through 9) is 0.025% with an average value of 0.0199% by plates used to as.' mble the shroud (see individual plate carbon content in Attachment 3 of the original letter). BWROG-94100 states that crack indications have been detected in core shroud materials in the range of 0.023% to 0.06% carbon. Weld materials used for assembly of the shroud geometry components are either 308 or 308L, with a maximum carbon content of 0.072%. The weld material for the shroud support assembly is nickel-chrome-iron (Table 1 Note).

BWR-94100, Table 3 identifies the extent of circumferential cracking detected in twenty-six shrouds inspected. Of these twenty-six shrouds, thirteen are constructed of 304L with a maximum carbon content allowed of 0.04%. No cracks have been found in eleven of these thirteen 304L shrouds. The two shrouds found with cracks each had a single weld with crack indications between 100 to 200 inches in length. Only four 304 (not low carbon) shrouds were found to have cracks exceeding 200 inches. In all four of these 304 cases and the two 304L cases, carbon content exceeds 0.06 percent or on-line years were high (>10.0 years). These values exceed corresponding values at CPS by a considerable margin. Reasonablejudgment can conclude that based upon the low carbon content of the CPS shroud and the projected hot operating years at RF-6 (estimated to be 7.96), the likelihood of cracks existing that would degrade the structural margin of the shrcud is extremely remote. This agrees with the conclusion contained in BWROG-94097. The amount of carbon in the CPS shroud is well below the threshold value for being a Category B plant.

B. Reactor Water Chemistry:

Extensive stress corrosion cracking (SCC) testing as well as theoretical modeling demonstrate that SCC initiation and growth are strongly dependent upon reactor water chemistry. The key parameters of water chemistry are coolant impurity levels as indicated by average coolant conductivity, specific impurity ionic species, and local environmental oxidizing power as defined by the electrochemical corrosion potential (ECP) at the surface of the component.

These values are for steady state conditions with the reactor at power levels at or above seventy-five percent thermal. ECP measurements are not taken at CPS.

However, since reactor coolant oxygen values are normal for a " normal water chemistry" (i. e., no hydrogen injection) plant, ECP values are assumed to be about

+200 mV standard hydrogen electrode (SHE), as shown in BWROG-94100. The effective full power years (EFPY) are used to show cycle length or time at a specific value.

CPS reactor water chemistry has historically achieved better (lower) values than the j

" achievable values" specified in the 1986 BWR Water Chemistry Guidelines. The one exception was conductivity during the first fuel cycle. Subsequent to the 1

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to U-602369 Page 5 of10 first fuel cycle, reactor water chemistry continually improved such that by cycle four, even the more strict 1993 BWR Guidelines were being met, with the exception of conductivity. In regard to conductivity, values for all cycles were increased by the presence of chromate. According to the 1993 BWR Water Chemistry Guidelines, the chromate levels experienced by CPS are of only marginal SCC concern.

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The historical values for key reactor water chemistry parameters at CPS are provided in Table 2.

Table 2 -- CPS Reactor Water Chemistry Operating EFPY Conductivity C1(ppb)

SO (ppb)

CrO (ppb) 4 4

Cycle (pmho/cm) j 1

0.99 0.277 5.66 5.8 2

0.82 0.142 1.65 4.5 25 3

0.87

(,.'170 1.30 2.6 47 4

1.22 0.180 1.10 1.4 33 5

1.12(l) 0.160 0.80 3.9 29 l

Total or Cycle 5.0 0.186 2.10 3.64 33.5(4)

Average l

1986 EPRI(2) 0.2 15 15 N/A(5) l 1993 EPRI(3) 0.11 1.0 2.0 N/A l

Table 2 Notes:

1.

The parameter variables for cycle five are through 10/31/94 and the EFPY is through 03/12/95 (projected end of cycle five).

2.

1986 EPRI are the achievabic values from BWR Normal Water Chemistry Guidelines: 1986 Revision, EPRI NP-4946-SR.

3.

1993 EPRI are the median values from BWR Noimal Water Chemistry Guidelines: 1993 Revision, Normal and Hydrogen Water Chemistry, EPRI TR-103515.

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Data based upon excluding cycle one during which chromate was not l

measured.

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Not available.

Two domestic plants with shrouds fabricated of 304L stainless steel (like the CPS shroud) were found to have cracks. The first five-year average conductivity of their reactor water as reported in BWROG-94100 is 0.459 and 0.695 uS/cm. These conductivity values are significantly higher than CPS's first five-year average of 0.186 l

i uS/cm.

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The quality of reactor water at CPS, compared to BWRs that have detected significant l

core shroud cracks, is clearly excellent, one of the lowest conductivities reported in l

BWROG-94100 According to BWROG-94100, plant coolant condutivity correlates I

well with the extent of cracking reported at BWRs. Since SCC likely initiates relatively I

early in life, a plant parameter called the First Five Cycle Mean Conduedvity was reported for all domestic BWRs. For plants exhibiting conductivity value.? below 0.22 uS/cm, cracking detected to date has been limited to less than 30 inches in.'ength.

CPS's First Five Cycle Mean Conductivity has been calculated to be 0.186 u3/cm, well l

below the threshold value for being placed in Category B. This low conductivity, l

which would be even lower without the chromate contribution, represents strong technical evidence supporting the conclusion that there are no safety significant cracks in CPS's shroud.

C. Neutron Fluence:

The neuvon flux value for the peak location at the inner vessel wall, as reported in

" Implementation of Regulatory Guide 1.99 Revision 2 for CPS," SASR 89-59, by General Electric, October 1989, at the end of cycle one (EFPY=0.99), was used to scale the spatial distribution. This scaling factor was applied to the flux profile found in EPRI NP-152, PWR and BWR Radiation Environments for Radiation Damage Studies by Science Applications Inc., September 1977 [EPRI NP-152]. This method of neutron fluence estimation shows that the upper shroud shell, piece number five in, will experience the greatest neution fluence, approximately,1.98 E+20 nyt. This value is based upon EFPY through October 12,1996, (projected end of cycle six). GE Service Information Letters (SIL) 572R1 and 574 indicate that fluence-induced damage begins at approximately 3 5 E+20 nyt in non-sensitized material. The highest CPS fluence levels are a factor of 1.5 to 2.5 below that required for the l

inception ofirradiation assisted stress corrosion cracking as reported in BWROG-94097 and BWROG-94100.

The potential errors of this method of estimation include: 1) assumptions that the neutron flux distribution for subsequent cycles remains the same as in cycle one,2) errors introduced in interpolation of the graphs in EPRI NP-152 are no worse than the accuracy of the neutron dosimetry technique,3) the accuracy of neutron dosimetry on boat samples removed from damaged components is the same as for the Regulate.y Guide 1.99 sample, and 4) that the flux shape at this facility is approximately the same as that used in EPRI NP-152. Using engineering judgment, the accuracy of the estimated values to which CPS shroud pieces have been subjected should be accurate within a factor of two. It should be noted that this is the same accuracy as used in several of the original GE design specifications.

to U-602369 Page 7 of10 BWROG-94097 and BWROG-94100 report that cracking in low carbon stainless steel would nc,t be expected to initiate until the neutron fluence exceeded a threshold value of approximately 4E+20 nyt (E)) Mev). It is expected that approximately 12.7 EFPY l'

are required to produce this level of neutron fluence at CPS. The 12.7 EFPY are equivalent to approximately 16.0 hot operating years (assuming CPS maintains the i

historical EFPY to hot operating year ratio of 0.79). Therefore, it is reasonable to l'

conclude that low neutron fluence in the CPS shroud contributes to the very low likelihood that it contains or will contain significant cracks by the start of RF-6.

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This conclusion is substantiated by the data tables in BWROG-94097 and BWROG-i 94100, which show that the two 304L shrouds with~ up to 200 inches of crack have ten or more on-line years, and there are no cracks in 304L shrouds with up to 8.4 on-line j

years (with the exception of two 1 inch vertical indications at Fermi-2).

D. Fabrication:

CPS is a BWR-6 with a core shroud ',bricated by Sun Shipbuilding and Drydock Company of Chester, Pennsylvania.,he core shroud was fabricated and Code stamped t

to AShE Boiler and Pressure Vessel Code,Section III (Nuclear Power Plant-Components),1974 Edition with 1974 Summer Addenda.

i According to BWROG-94100, the CPS shroud is fabricated in two sections: Welds

- H1 and H2 are included in the top guide section and welds H3 through H7 are included in the lower shroud section. The significance of this design with regard to SCC susceptibility is the elimination of welding on both sides of the top guide ring. A bolted connection replaces the redundant welding on both the top and bottom surfaces of the

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top guide support ring, as shown in BWROG-94097, to be typical construction. This configuration may result in significantly less residual stress in the vicinity of the H2 and H3 welds that have demonstrated significant cracking in the older design.

Fabrication controls and processes were required by GE to minimize residual stresses.

All materials were purchased in the solution annealed condition with a maximum hardness of Rockwell B-90. Cold forming was permitt:ed on the solution heat treated material provided that the final haroness did not exceed Rockwell B-94. Based on heat treatment records, welds H1 through H6B, except for H5, received a low temperature stress relief at 750 degrees for a minimum of twenty-four hours. This heat treatment provides improved dimensional stability by minimizing peak. residual stress causes by i

welding and forming. Although not performed for enhanced SCC prevention, the heat treatment reduced peak stresses resulting in improved resistance to corrosion cracking aggravated by stress.

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to U-602369 Page 8 of 10 From BWROG-94097 and BWROG-94100, the worst cracking has been found in plants with cut plate support rings (in contrast to forged rings). CPS support rings were fabricated from cut plate. However, a review of those plants with greater than 100 inches of cracks and support rings fabricated from cut plate indicates an average of 12.5 on-line years and a first five cycle conductivity of 0.453 uS/cm. Both of these values are significantly greater than the corresponding values for CPS.

The combination of material control, fabrica ion controls and bolted connection at the top guide, combined with few on line years and good water chemistry, contributes to lowering the likelihood of safety significant cracks in the CPS shroud.

E. Hot Operating Time:

i On-line years equate to hot operating time with coolant temperature above 200 degrees. As defined in BWROG-94097, Section 3.0, the GE PLEDGE model predicts rapid crack growth with exposure to hot coolant. The validity of this model is discussed in the same reference. Entering Figure 3-3 with the average conductivity for i

CPS of 0.lS6 rounded to 0.2 uS/cm and the predicted 7.96 hot operating years at the end of the sixth fuel cycle, the maximum crack depth is predicted to not exceed 1.0 inches. Since the nominal plate thickness of our shroud is 2.0 inches, the remaining ligament would be a minimum of 1.0 inches. An adequate strength ligament to ensure structural integrity is judged to be available even with a ninety percent through wall crack depth, which is the depth allowed in the extreme case of a 360 degree crack. A minimum ligament of ten percent of the nominal thickness, or 0.2 inches would be exceeded by 0.8 inches. (The original CPS response incorrectly reported that ten percent of the nominal two inch thickness is 0.4 inches.)

Therefore, it is reasonable to conclude that the likelihood of safety significant cracks occurring in the CPS shroud prior to inspection is extremely remote. The estimated 7.96 hot operating years when entering RF-6 plac s CPS at the threshold of Category B plants, indicating the appropriateness ofinspecting during that outage.

F. Safety Analysis Summary:

Information provided in the BWROG reference documents and the CPS specific details provided can be summarized as follows and shows that CPS is clearly a Category A plant until RF-6:

1.

The CPS parameters are well below the threshold values of those BWRs that have identified shroud geometry component cracks. Hot operating years have been accurately determined from a detailed review determined of all mode change checklists from initial criticality (February 26,19S7) through August 8,1994. For conservatism, hot operating years were determined assuming that the plant would operate at full power until RF-5.

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2.

The conditions at CPS that influence shroud susceptibility to stress corrosion cracking are equal to or better than BWR facilities that have i

inspected and found no shroud cracks.

3.

If crack initiation is conservatively assumed, the rate of growth would not cause a reduction in shroud structural margin before the scheduled inspection in RF-6. This refueling outage for CPS will begin in October 1996.

4.

BWROG-94097 and BWROG-94100 cite an " extremely low yearly l

probability" of a seismic induced loss of coolant accident (LOCA) or l

main steam line break in coincidence with a core shroud crack that is greater than ninety percent through wall and 360 degrees. The probability for the combination of a seismic event and a LOCA has been analyzed for CPS using our plant specific probability risk assessment (PRA) methodology. NUREG/CR-4792, " Probability of Failure in i

BWR Reactor Coolant Piping, Vol.1: Summary Report," Lawrence r

Livermore National Laboratory, March 1989, values are used for the l

seismic event with a main steam line break. The NUREG value for l

probability of a seismic event / main steam line break was qualified by l

determining that CPS has a similar seismic spectrum response as the j

l subject plant. The yearly prob' ability of a safe shutdown earthquake and j

a LOCA is 1.1 E-10, while the probability of a seismic event and a main steam line break is 4.54 E-13. Even when considering twenty-two l

months of operation these values constitute an extremely low probability l

ofoccurrence.

Based on engineering judgment; the information provided in Generic Letter 94-03; the references cited herein for structural margins, shroud displacement likelihood, the integrated shroud assessment, and the shroud cracking safety assessment; in conjunction with this submittal, provide ample justification that CPS can operate safely as a Category A plant until shroud inspection in RF-6.

3.

NRC Reporting Requircment Request 1.(c): A drawing or drawings of the core shroud configuration showing details of the core shroud geometry.

IP Response: The copies of these documents were provided in our August 24,1994, transmittal and have not changed.

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NRC Reporting Requirement Request 1.(d): A history of shroud inspections for the plant should be provided addressing date, scope, methods and results, if applicable.

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- Page 10 of 10 IP Response:

T1.e description of the VT-3 inspections performed during the March 1992 outage is unchanged from our August 24,1994, :etter response.

5.

NRC Requested Licensing Action (5): Work closely with the BWROG on j

coordination ofinspections, evaluations and repair options for all BWR internals l

susceptible to SCC.

IP Response:

CPS currently participates on the BWRVIP Integration, Inspection and Mitigation committees and intends to remain an active member of the BWRVIP until resolution of vessel internals corrosion issues.

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