Comment Number: 8-1

Submitted by Westinghouse Owners Group

Comment: In Chapter 8, Ongoing Work, it is noted that the LOCA elicitation results were for normal operating conditions only. The effects of Service Level D transients, of which seismic was found by NRC to be the most prominent, were not considered in the elicitation efforts. The reason seismic loading was not explicitly considered was that most of the expert panel did not believe that it would significantly change the LOCA frequencies for normal operation. Experience from probabilistic fracture mechanics calculations indicates that severe seismic loading, such as that from a design-basis safe shutdown earthquake, could increase the conditional probability of failure in flawed piping by one to two orders of magnitude. However, the probability of having the severe seismic loading during the worst time in life, such as the 40th or 60th year of operation, would be a maximum of 0.001 and would likely be much less. Thus, the maximum effect of this severe seismic loading would be to increase the LOCA frequency during normal operating conditions by 10 percent. This increase was deemed to be insignificant relative to the other uncertainties that were considered by the expert panel in the elicitation process for LOCA frequencies.

Response: The results from a separately-sponsored NRC-led seismic LOCA study (Reference: Chokshi, N. C., Shaukat, S. K., Hiser A. L., DeGrassi, G., Wilkowski, G., Olson, R., and Johnson, J. J., "Seismic Considerations For the Transition Break Size," NUREG-1903, U. S. Nuclear Regulatory Commission, February 2008) tend to support this comment. In this study, both unflawed and flawed piping analyses were conducted in order to ascertain the magnitude of any potential adjustments to the baseline TBS for the proposed rule change to 50.46a due to failures associated with seismic loading.

The principal findings from this study are that the critical flaws associated with the stresses induced by seismic events of 10-5 and 10-6 annual probability of exceedance are large. When considering the effects of mitigation strategies to preclude large flaws in service, the probabilities of pipe breaks larger than the TBS are likely to be less than 10-5 per year. Similarly, for the cases studied, the probabilities of indirect failures of large RCS piping systems are less than 10-5 per year.

These findings tend to support the contention of the commenter that seismic loading would not significantly change the LOCA frequencies under normal operation. As a result of this and related comments, the NRC report on Seismic Considerations for the Transition Break Size is now referenced in Section 2 of NUREG-1829. In addition, a summary of the seismic LOCA analysis and results is provided in the Executive Summary and in Section 7.2 of the report. Additionally, Section 2 clearly identifies that the elicitation LOCA frequency estimates do not consider rare event loading from seismic, severe water hammer, and other similar sources.

Comments Related to Section 9 of NUREG-1829

None

Comments Related to Appendix A of NUREG-1829

None

Comments Related to Appendix B of NUREG-1829

None

Comments Related to Appendix C of NUREG-1829

None

Comments Related to Appendix D of NUREG-1829

Comment Number: D-1

Submitted by Joseph Conen of the BWR Owners Group

Comment: Figure D.7 in Appendix D shows two through-wall IGSCC cases for 22 inch and 28 inch stainless steel pipe field history data. This reviewer is not aware of any through-wall IGSCC cracks in large diameter (>20-inch) BWR stainless steel pipes. A primary reason for this is the presence of mid-wall compressive weld residual stresses in such pipe that tend to retard deep cracks.

Response: Figure D.7 in the draft NUREG only shows selected IGSCC data points (only weld flaws for which detailed sizing data are available). In actuality there are have been other leaks in 22-inch and 28- inch diameter recirculation lines in BWRs. According to the expanded OPDE database used as the basis of this query (currently 1,215 records on IGSCC), there have been 10 instances of circumferential through-wall cracking in large diameter (D=22 inch to 28 inch recirculation system piping, 8 of which were leaks. Three of the leaks were in 22-inch diameter piping: a cap-to-manifold leak at Hatch-1 (LER 82-089, November 1982) and two welds at Monticello (LER 82-013, October 1982). The other five leaks were associated with the 28-inch diameter recirculation line at Brunswick-1 (LER 85-026, July 1985): Weld 1B32-RR-28-A-4, Weld 1B32-RR-28-A-14, Weld 1B32-RR-28-B-4, Weld 1B32-RR-28-B-8, and Weld 1B32-RR-28-A-15.

Comments Related to Appendix E of NUREG-1829

None

Comments Related to Appendix F of NUREG-1829

None

Comments Related to Appendix G of NUREG-1829

None

Comments Related to Appendix H of NUREG-1829

None

Comments Related to Appendix I of NUREG-1829

None

Comments Related to Appendix J of NUREG-1829

None

Comments Related to Appendix K of NUREG-1829

None

Comments Related to Appendix L of NUREG-1829

None

NRC FORM 335 (9-2004) NRCMD 3.7	1. REPORT NUMBER (Assigned by NRC, Add Vol., Supp., Rev., and Addendum Numbers, if any.)
U. S. NUCLEAR REGULATORY COMMISSION

BIBLIOGRAPHIC DATA SHEET

NUREG-1829, Vol. 2

(See instructions on the reverse)

3. DATE REPORT PUBLISHED MONTH YEAR April 2008 4. FIN OR GRANT NUMBER ___________ N6360 6. TYPE OF REPORT

Technical

7. PERIOD COVERED (Inclusive Dates)

N/A

8. PERFORMING ORGANIZATION — NAME AND ADDRESS (If NRC, provide Division, Office or Region, U. S. Nuclear Regulatory Commission, and mailing address; if contractor, provide name and mailing address.)

Division of Engineering Battelle-Columbus

Office of Regulatory Research 505 King Avenue

U. S. Nuclear Regulatory Commission Columbus, OH 43201

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION — NAME AND ADDRESS (If NRC, type "Same as above"; if contractor, provide NRC Division, Office or Region, U. S. Nuclear Regulatory Commission, and mailing address.)

Same as above

10. SUPPLEMENTARY NOTES

A. Csontos, NRC Project Manager_____________________________________________________________________

11. ABSTRACT (200 words or less)

The NRC is developing a risk-informed revision of the design-basis pipe break size requirements in 10 CFR 50.46, Appendix K to Part 50, and GDC 35 which requires estimates of loss-of-coolant-accident (LOCA) frequencies as a function of break size. Separate BWR and PWR piping and non-piping passive system LOCA frequency estimates were developed as a function of effective break size and operating time through the end of license extension. The estimates were based on an expert elicitation process which consolidated service history data and insights from probabilistic fracture mechanics studies with knowledge of plant design, operation, and material performance.

The elicitation required each member of an expert panel to qualitatively and quantitatively assess important LOCA contributing factors and quantify their uncertainty. The quantitative responses were combined to develop BWR and PWR total LOCA frequency estimates for each contributing panelist. The individual estimates were then aggregated to obtain group estimates, along with measures of panel diversity. Sensitivity studies were conducted to examine the effects of distribution shape, correlation structure, panelist overconfidence, measures of panel diversity, and aggregation method. The group estimates are most sensitive to the method used to aggregate the individual estimates.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)	13. AVAILABILITY STATEMENT
piping	unlimited
risk-informed	14. SECURITY CLASSIFICATION
emergency core cooling system (ECCS)	(This Page)
loss-of-coolant accident (LOCA)	unclassified
break frequencies	(This Report)
design-basis break size	unclassified
LOCA frequency estimates
expert elicitation	15. NUMBER OF PAGES
aging
	16. PRICE

[1] Each panelist’s quantitative elicitation responses can be found through the “Electronic Reading Room” link on the NRC’s public website (http://www. nrc. gov/) using the Agencywide Documents Access and Management System (ADAMS). The document is found in ADAMS using the following accession number: ML080560005.

[2] The nomenclature for the table and figure numbers is such that the letter B refers to Appendix B, the first number (1 or 2) refers to a figure associated with either the first or second panel meeting, and the second number refers to the numerical sequence of that particular table or figure in the text for the applicable meeting, i. e., either first or second.

[3] See for example the report EPRI NP-2472 (The Growth and Stability of Stress Corrosion Cracks in Large-Diameter BWR Piping, July 1982).

[4] The figure is reproduced courtesy of K. N. Fleming (Technology Insights, Inc., San Diego, California).

[5] Details on Bayesian reliability analysis is found in text books on statistical analysis of reliability data; e. g., Martz and Waller (1991): Bayesian Reliability Analysis, Krieger Publishing Company, Malabar (FL), ISBN 0-89464-395-9. For conjugate functions like the gamma and beta distributions a Bayesian point estimator for the failure rate is the mean of respective posterior probability density function, or:

X = (8 + r)/(p + T) — gamma X = (8 + r)/(8 + p + n) — beta

Where, (8, p) are the parameters of respective distribution and (r, T, n) correspond to new evidence (i. e., ‘r’ failures in ‘T’ hours, or ‘r’ failures in ‘n’ tests).

[6] See SKI 98:30 [D.15] for details.

[7] This table includes active leaks (= leaks detected during routine power operation) and ‘non-active’ leaks (= leaks discovered during

change of plant mode of operation), but it excludes ‘ISI-leaks.’ Appendix A, Table A-5 includes details on the through-wall cracks in NPS12, NPS22 and NPS28 Reactor Recirculation piping as included in Table D.4 above._________________________________________________________

[8] The mean of weld count in NPS20-, 22- and 24-piping.

[9] NPS30 is used to characterize the CL — and HL-piping.

[10] Table 2-3, page 2-10; Aj = 1.34E-05 (RF = 100). TR-111880: Piping System Failure Rates and Rupture Frequencies for Use in Risk Informed In-Service Inspection (September 1999).

[11] The term ‘non-active leak’ is taken to mean a through-wall flaw without visible leakage or with a small, detectable leakage that stays relatively constant over time.

[12] The database includes a single event involving the fracture of a small-diameter steam line due to seismic event (Fukushima-Daiichi Unit 6 on 07-21­2000).

[13] See for example T. V. Vo et al (1991). “Estimates of Rupture Probabilities for Nuclear Power Plant Components: Expert Judgment Elicitation,” Nuclear Technology, 96:259-270.

[14] At the time of the pipe break the primary system was at 278 C (533 F) and 15.3 MPa (2,225 psi) primary system pressure with a secondary system pressure of 6.2 MPa (900 psi)

[15] The effects of applying a load-controlled overload stress at a specified time were studied. This is called a design-limiting stress, and represents an overload event, such as water hammer or a seismic event even larger than the 5 SSE already considered for this component.

[16] This information needed to be supplied because the transient experience for the Naval Nuclear program is a) confidential and b) not applicable to commercial plants.

[17] Failures are classified using four categories: partial through-wall cracks, through-wall cracks without a significant leak rate (typically indicated by a boric acid deposit), leaks, and joint failures (i. e., non-welded connection).

[18] A half failure (0.5) was added to all degradation mechanism (DM) totals to force the representation of all DMs.

[19] Lydell, B., “Independent Review of SKI 96-20 Database,” Technical Note 1996-01, SKI Ref. No. 14.2-940477, February 1996.

[20]

Letter from Frederick P. Schiffley, II, Chairman to USNRC, “Westinghouse Owners Group Comments on Draft NUREG-1829, ‘Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Expert Elicitation Process’ (MUHP-3062)”, WOG-05- 517, dated November 28, 2005, ADAMS Accession # ML0503340274.