LIC-03-0079, Supplemental Response to Request for Additional Information, Integrated Leak Rate Testing Surveillance Interval Amendment Request

Supplemental Response to Request for Additional Information, Integrated Leak Rate Testing Surveillance Interval Amendment Request

ML031490377
Person / Time
Site:	Fort Calhoun
Issue date:	05/21/2003
From:	Ridenoure R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-03-0079, TAC MB6473
Download: ML031490377 (12)

Text

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Omaha Public Power Distict 444 Soutt 16th Street fall Omaha NE 68102-2247 May 21, 2003 LIC-03-0079 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

References:

1. Docket No. 50-285

2. Letter from OPPD (D. J. Bannister) to NRC (Document Control Desk) dated October 8, 2002, Fort Calhoun Station Unit No. 1 License Amendment Request, LAR - Risk-Informed One Time Increase in Integrated Leak Rate Test Surveillance Interval (LIC-02-0108)

3. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated March 24, 2003, Request for Additional Information Related to Ft. Calhoun Station Integrated Leak Rate Test Surveillance Interval (TAC No.

MB6473) (NRC-03-055)

4. Letter from OPPD (R. T. Ridenoure) to NRC (Document Control Desk) dated April 11, 2003, Response to Request for Additional Information, Integrated Leak Rate Testing Surveillance Interval Amendment Request (TAC No. MB6473) (LIC-03-0055)

SUBJECT:

Supplemental Response to Request for Additional Information, Integrated Leak Rate Testing Surveillance Interval Amendment Request (TAC No.

MB6473)

This letter provides the response to Question 6 of the Nuclear Regulatory Commission's (NRC's) Request for Additional Information (Reference 3) with regard to Omaha Public Power District's (OPPD's) request, "Fort Calhoun Station Unit No. 1 License Amendment Request, LAR - Risk-Informed One Time Increase in Integrated Leak Rate Test Surveillance Interval" (Reference 2). The responses to Questions 1 through 5 were submitted in Reference 4.

I declare under penalty of perjury that the forgoing is true and correct. (Executed on May 21, 2003). No commitments are made to the NRC in this letter.

Employment vith Equal Opportunity

U. S. Nuclear Regulatory Commission LIC-03-0079 Page 2 If you have any questions or require additional information, please contact Dr. R. L. Jaworski of the FCS Licensing staff at (402) 533-6833.

Sincerely, c: E. W. Merschoff, NRC Regional Administrator, Region IV A. B. Wang, NRC Project Manager J. G. Kramer, NRC Senior Resident Inspector Winston & Strawn

LIC-03-0079 Attachment Page 1 Response to NRC Request for Additional Information Fort Calhoun Station, Unit 1 One Time Increase in Integrated Containment Leak Rate Test Interval NRC Question 6:

Inspections of some reinforced and steel containments (e.g., North Anna, Brunswick, and D. C.

Cook, Oyster Creek) have indicated degradation from the uninspectable (embedded) side of the steel shell and liner of primary containments. The major uninspectable areas of the Fort Calhoun Station (FCS) containment would include those at the liner concrete interface in the dome and the cylinder, and in the basemat liner embedded in the concrete. Please provide a quantitative assessment of the impact on large effluent release frequency (LERF) due to age related degradation in these areas, in support of the requested integrated leak rate test (ILRT) interval extension OPPD Response:

An analysis, using the Calvert Cliffs Nuclear Power Plant (CCNPP) method (Reference Al), was performed to determine how containment liner corrosion affects the risk associated with extending the ILRT for Fort Calhoun Station.

The following issues were addressed:

• Differences between the containment basemat and the containment cylinder and dome;

• The historical liner flaw likelihood due to concealed corrosion;

• The impact of aging;

• The liner corrosion leakage dependency on containment pressure; and

• The likelihood that visual inspections will be effective at detecting a flaw.

Assumptions A. Two corrosion events have been identified which could potentially result in liner corrosion.

It is assumed that these events may be precursors for a larger containment leakage.

B. A half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures (See Table 1, Step 1).

C. The success data was limited to 6 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection.

D. The liner flaw likelihood is assumed to double every five years.

E. The likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists is a function of the pressure inside the Containment. Even without the liner, the Containment is an excellent barrier. But as the pressure in Containment increases, cracks will form. If a crack occurs in the same region as a liner flaw, then the containment

LIC-03-0079 Attachment Page 2 atmosphere can communicate to the outside atmosphere. At low pressures, this crack formation is extremely unlikely. Near the point of containment failure, crack formation is virtually guaranteed. Anchored points of 0.1% at 20 psia and 100% at 200 psia were selected based on conservative representation of the failure probabilities and pressures provided in the FCS independent plant evaluation (IPE) structural analysis section.

Intermediate failure likelihoods are determined through logarithmic interpolation.

Sensitivity studies are included that decrease and increase the 20-psia anchor point by a factor of 10 (See Table 4 for sensitivity studies).

F. The likelihood of leakage escape (due to crack formation) in the basemat region is considered to be 10 times less likely than the containment cylinder and dome region (See Table 1, Step 4).

G. A 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used.

H. All non-detectable containment over-pressurization failures are assumed to be large early releases.

I. Containment failure probabilities used for this calculation are based on the FCS specific Containment Fragility Curves.

J. The ILRT test pressure of 75 psia will bound the test failure condition for the Fort Calhoun Station.

K. The Base methodology uses continuous fits to estimate containment failure probabilities, whereas the CCNPP method used discrete addition.

Analysis Table 1 presents the results of the analysis of the likelihood of non-detected containment leakage because of liner corrosion. The analysis considers the inspectable portion of the liner and the uninspectable portion of the liner. Approximately 86% of the interior surface of the Unit 1 containment liner is accessible for visual inspection. The 14% that are inaccessible for visual inspection include the fuel transfer tube shielded area, the area under the concrete floor, and the area behind the elevator shaft. The area under the concrete floor accounts for almost all of the inaccessible area.

LIC-03-0079 Attachment Page 3 Table 1 Liner Corrosion Base Case Containiment Cylinder and CotimnBae t Step - Description -Dome 14%0

______ ~~~~~~~~~~~~~~~86

%/

1 Historical Liner Flaw Likelihood Events: 2 Events: 0 Failure Data: Containment location specific (Brunswick 2 and North Assume half a failure Anna 2)

Success Data: Based on 70 steel-lined 2/(70

• 6.0) = 4.76E-3 0.5/(70

• 6.0) = 1.19E-3 Contaimnents and 6.0 years since the 10 CFR 50.55a requirement for periodic visual inspections of containment surfaces.

2 Age Adjusted Liner Flaw Likelihood Year Failure Rate Year Failure Rate During 15-year interval, assumed failure rate 1 1.93E-3 1 4.83E-4 doubles every five years (14.9% increase per avg 5 - 10 4.76E-3 avg 5 - 10 l.19E-3 year). The midpoint for 5 to 10" year was set to the historical failure rate 15 1.35E-2 15 3.37E-3 15 year avg = 5.55E-3 15 year avg 1.39E-3 3 Increase in Flaw Likelihood Between 3 and 15 years Uses aged adjusted liner flaw likelihood (Step 7.87% 1.97%

2), assuming failure rate doubles every five years. See Tables 5 and 6.

4 Likelihood of Breach in Containment given Pressure Likelihood of Pressure Likelihood Liner Flaw (psia) Breach (psia) of Breach The upper end pressure is consistent with the 20 0.10% 20 0.01%

Fort Calhoun Station Probabilistic Risk 75 (ILRT) 0.83% 75 (ILRT) 0.083%

Assessment (PRA) Level 2 analysis. 0.1% is 80 1.0% 80 0.10%

assumed for the lower end. Intermediate 120 4.6% 120 0.46%

failure likelihoods are determined through logarithmically interpolation. The basemat 200 100% 200 10%

failure likelihood is assumed to be 1/10 of the cylinder/dome analysis 5 Visual Inspection Detection Failure 10% 100%

Likelihood 5% failure to identify visual Cannot be visually flaws plus 5% likelihood inspected.

that the flaw is not visible (not through-cylinder but could be detected by ILRT)

All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption.

6 Likelihood of Non-Detected Containment 0.0065% 0.0016%

Leakage 7.87%

• 0.83%

• 10% 1.97%

• 0.083%

• 100%

(Steps 3

• 4* 5)

LIC-03-0079 Attachment Page 4 The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome and the containment basemat.

Total Likelihood of Non-Detected Containment Leakage = 0.0065% + 0.0016% = 0.0081%

The non-large early release frequency (LERF) containment over-pressurization failures for Fort Calhoun Station are estimated, based on the PRA, at 1.07E-05 per year. The non-LERF frequency is obtained by adding the Class 1 (intact) and late releases contribution from Class 7 (severe accident). If all non-detectable containment leakage events are considered to be LERF, then the increase in LERF associated with the liner corrosion issue is:

Increase in LERF (comparing a 3 in 10 year ILRT to a one in 15 years ILRT)

= 0.0081%

• 1.07E-5= 8.7E-10 per year.

Please note that the current approved ILRT test interval at Fort Calhoun Station is one ILRT every ten years. The above increase in LERF is greater than it would be when comparing to the one in ten year frequency currently in effect.

Change in Risk The risk of extending the ILRT from 3 in 10 years to 1 in 15 years is small and estimated as being less than 1E-7 per year. It is evaluated by considering the following elements:

1. The risk associated with the failure of the Containment due to a pre-existing containment breach at the time of core damage (Class 3 events).

2. The risk associated with liner corrosion that could result in an increased likelihood that containment over-pressurization events become LERF events.

3. The likelihood that improved visual inspections (frequency and quality) will be effective in discovering liner flaws that could lead to LERF.

These elements are discussed in detail below.

Pre-existing Containment Breach The original submittal addressed Item 1. The submittal calculated values of the increase in risk using the Combustion Engineering Owners Group (CEOG) method (Reference A2) and a previously NRC-approved method (Reference A3). The ILRT interval for Fort Calhoun Station is currently once in ten years. Reference A3 indicated that extending this interval from once in ten years (i.e., the current interval) into once in fifteen years (the proposed interval) would result in a LERF increase from 1.226E-09 per year to 1.407E-08 per year. When measured from the original Appendix J inspection interval (i.e., three times in ten years), the increase in each risk metrics is somewhat larger. Table 2 lists the changes in risk metrics associated with extending the ILRT to a once in fifteen years interval as compared to the original Appendix J reference condition (i.e., three times in ten years).

LIC-03-0079 Attachment Page 5 Table 2 Original Submitted with Updated Values (from three times in ten ears (3/10) to once per 15 years) 44 i Method 0:0 d LERF:;;- inra X: -- iiPerson-rem/yr i Percentage Increase

i:;lIethod- E -Z: LERIFIncrease-yi : a-- i E increase - in Person-remlyr CEOG Method 2.86E-09 0.06 0.80%

NRC Approved 4.22E-08 0.01 0.16%

Method Liner Corrosion The original submittal did not fully address the risk associated with liner corrosion. Table 3, below shows an additional small increase in LERF of 0.87E-9 from going from a 3/10 to a 15 year ILRT interval. Thus, Table 2 is modified as follows:

Table 3 Updated Values with Corrosion Impact (from three times in ten yars (3/10) to once per 15 years)

Increase -- -- - Percentage Increase i-Person-rem/yr

- Xi:E: ;

Methodt

E RF E -i -iincrease in Person-rem/yr CEOG Method 2.86E-09 0.060 0.80%

CEOG Method with 3.73E-09 0.063 0.84%

Liner Corrosion NRC-Approved Method 4.22E-08 0.0140 0.160%

NRC-Approved Method 4.31E-08 0.0141 0.161%

with Liner Corrosion Visual Inspections The original submittal did not fully address the benefit of the Subsection IWE visual inspections.

Visual inspections following the 1996 change in the ASME Code are believed to be more effective in detecting containment liner flaws. In addition, the flaws that are of concern for LERF are considerably larger than those of concern for successfully passing the ILRT.

Integrated leakage rate test failures have occurred even though visual inspections have been performed. However, the recorded ILRT flaw sizes for these failed tests are much smaller than that for LERF. Therefore, it is likely that future inspections would be effective in detecting the larger flaws associated with a LERF.

LIC-03-0079 Attachment Page 6 IWL of Section XI of the ASME B&PV Code, provides detailed requirements for ISI of Containment Structures. Inspection (which includes examination, evaluation, repair, and replacement) of the concrete containment liner plate, in accordance with the 10 CFR 50.55a requirements, involves consideration of the potential corrosion areas. Although the improvement gained by this requirement varies from plant to plant, it is believed that this requirement makes the detection of flaws post-September 1996 much more likely than pre-September 1996 using visual inspections.

Visual inspection improvements directly reduce the A LERF increases as calculated in the CEOG method and NRC-approved method. A VT General inspection was completed on the liner during the 2001 Refueling Outage and will be repeated during the 2003 Refueling Outage scheduled to commence September 12, 2003. A more thorough detailed visual examination is performed on any area with evidence of degradation.

Table 7 illustrates the benefit of visual inspection improvements on the A LERF calculations:

If the improved inspections (additional inspection, improved effectiveness, and larger flaw size) were 90% effective in detecting the flaws in the visible regions of the containment (5% for failure to detect and 5% for flaw not detectable [not-through-wall]), then the increased ILRT LERF frequency could be reduced by 22.6%. See Table 7 for additional sensitivity cases. This would result in a LERF increase of less than 1E-7.

LIC-03-0079 Attachment Page 7 Sensitivity Studies The following cases were developed to gain an understanding of the sensitivity of this analysis to the various key parameters.

Table 4 Liner Corrosion Sensitivity Cases Conitainment Viulnseto Age (Ste 2)

P Breach & Non-Visual

'n~~~~~Faws Likelihood is LERF*Flaw LERF Increase (R

Ste: pt 4)

Ex -Srd

(Step 5) __________________

Base Case Baseline 10% 100% 8.69E-10 Doubles every 5 years Doubles every 2 years Same as Base Base Base 9.56E-09 Doubles every 10 years Same as Base Base Base 4.28E-10 Base Base point 10 times Base Base 1.76E-10 lower Base Base point 10 times Base Base 4.30E-09 higher Base Same as Base 5% Base 5.22E-10 Base Same as Base 15% Base 1.22E-09 Lower Bound Doubles every 10 years Base point 10 times 5% 10% 5.19E-12 lower Upper Bound Double every 2 years Base point 10 times 15% 100% 6.62E-08 Doubl evhigher I I I

• Probability that flaw is large enough to cause a large release.

LIC-03-0079 Attachment Page 8 Table 5 Flaw Failure Rate as a Function of Time

Year;

i ;FailureRate Success Rate (FR): (1-FR) 0 1.68E-03 9.98E-01 1 1.93E-03 9.98E-01 2 2.22E-03 9.98E-01 3 2.55E-03 9.97E-01 4 2.93E-03 9.97E-01 5 3.37E-03 9.97E-01 6 3.87E-03 9.96E-01 7 4.44E-03 9.96E-01 8 5.10E-03 9.95E-01 9 5.8613-03 9.94E-01 10 6.73E-03 9.9313-01 11 7.74E-03 9.92E-01 12 8.89E-03 9.911E-01 13 1.02E-02 9.90E-01 14 1.17E-02 9.88E-01 15 1.35E-02 9.87E-01 Table 6 Failure Rate Success Rate Failure Rate Years (SR) (1-SR)

I to 3 9.94E-01 0.63%

1 to 10 9.64E-01 3.64%

1 to 15 9.1513-01 8.50%

A = 8.50% - 0.63% = 7.87% (delta between 1 in 3 years to 1 in 15 years)

LIC-03-0079 Attachment Page 9 Table 7 Benefit of Visual Inspection Improvements CEOG Factor :NRC ONRC Approved C Gethod Itt mprovement Approved Method vlinerv Mthod t Lner due to Visual i! Method Corrosion A LE  : Corrosion:

Inspections ALERF 'Considered A LERF Considered

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~~~A L E RF Pre-1996 Inspection 0.00% 4.22E-08 4.31E-08 2.86E-09 3.73E-09 Approach (Base Case)

Post- 1996 with 86.00% 5.9E-09 6.OE-09 4.0E-10 5.2E-10 Visual Inspections Perfectly Accurate Post-1996 with 81.7% 7.7E-09 7.9E-09 5.2E-10 6.8E-10 Visual Inspections 95% Accurate Post-1996 with 77.40% 9.5E-09 9.7E-09 6.5E-10 8.4E-10 Visual Inspections 95% Accurate and 5% chance of Undetectable Leakage Post-1996 with 64.5% 1.5E-08 1.5E-08 1.OE-09 1.3E-09 Visual Inspections 80% accurate and a 5% Chance of Undetectable Leakage

LIC-03-0079 Attachment Page 10 Conclusion Considering the benefit of improved visual inspections post-1996, the increase in risk is considered to be less than E-7 per year for LERF even when measured from the initial Appendix J baseline of inspection (i.e., three times in ten years). Changes less than 1.0E-7 per-year are considered insignificant according to Regulatory Guide 1.174. The one-time extension of the ILRT interval from 1-in-10 years (the currently approved interval) to 1-in-15 years is considered an acceptable risk increase.

References Al.Letter, C. Cruse, Calvert Cliffs Nuclear Power Plant (CCNPP) to Document Control Desk, "Response to Request for Additional Information Concerning the License Amendment Request for a one-time Integrated Leak Rate Extension," Calvert Cliffs Nuclear Power Plant (Docket No. 50-317), March 27, 2002.

A2.WCAP-15691, Revision 3, "Joint Applications Report for Containment Integrated Leak Rate Test Interval Extension," CEOG Task 2027, September 2002. Transmitted via Letter from D. J. Bannister (OPPD) to NRC, "Fort Calhoun Station Unit 1 License Amendment Request, Risk-Informed One-Time Increase in Integrated Leak Rate Test Surveillance Interval," dated October 8, 2002.

A3.LIC-02-01 08, "Fort Calhoun Station Unit No. 1 License Amendment Request, Risk Informed One-Time Increase in Integrated Leak Rate Test Surveillance Interval," D. J Bannister (OPPD) to USNRC, October 8, 2002