ML20205K768
| ML20205K768 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 04/06/1999 |
| From: | Nick Hilton NRC (Affiliation Not Assigned) |
| To: | Hutchinson C ENTERGY OPERATIONS, INC. |
| References | |
| TAC-MA2023, NUDOCS 9904140010 | |
| Download: ML20205K768 (8) | |
Text
p-if Mr. C. R:ndy Hutchinson April 6, 1999 Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING APPLICATION TO USE A RISK-INFORMED ALTERNATIVE TO CERTAIN REQUIREMENTS OF ASME CODE SECTION XI, TABLE IWB-2500-1 AT ARKANSAS NUCLEAR ONE, UNIT 1 (TAC NO. MA2023)
Dear Mr. Hutchinson:
By letter dated June 3,1998, you requested that the NRC approve a risk-informed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)Section XI, inspection requirements for Class 1, Category B-J piping welds (excluding socket welds) contained in Taole IWB-2500-1. Your application noted that, in generai, the risk analysis was performed consistent with the requirements of Code Case N-560.
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Adoitional information is necessary for the staff to independently conclude that the proposed
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alternative would provide an acceptable level of quality and safety.- During our review, we l
determined that various areas of your request require further justification, including the application of risk-informed principles, certain applications of Code Case N-560 methodology, and departures from Code Case N-560 methodology. As a pilot plant application, such a request for additional information (RAl) is typically necessary.
Enclosed is the RAI regarding your request. The RAI was discussed with Mr. Steve Bennett and others of your staff on April 1,1999, in order to support your request, end our target completion date of July 1999, we agreed that your response would be provided by May 17, j
- 1999, Sincerely, ORIGINAL SIGNED BY:
j Nicholas D. Hilton, Project Manager, Section 1 i
Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-313 DISTRIBUTION Docket File OGC
Enclosure:
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NUCLEAR REGULATORY COMMISSION E
E wAsMINGToN, D.C. 20seMo01 April 6, 1999
. Mr. C. Randy Hutchinson Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801 l
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING APPLICATION TO USE A RISK-INFORMED ALTERNATIVE TO CERTAIN REQUIREMENTS OF ASME CODE SECTION XI, TABLE IWB-2500-1 AT ARKANSAS NUCLEAR ONE, UNIT 1 (TAC NO. MA2023)
Dear Mr. Hutchinson:
By letter dated June 3,1998, you requested that the NRC approve a risk-informed altemat e to t
the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code)Section XI, inspection requirements for Class 1, Category B-J piping welds (excluding socket welds) contained in Table IWB-2500-1. Your application noted that, in general, the risk analysis was performed consistent with the requirements of Code Case N-560.
i Additional information is necessary for the staff to independently conclude that the proposed attemative would provide an acceptable level of quality and safely. During our review, we determined that various areas of your request require further justification, including the application of risk-informed principles, certain applications of Code Case N-560 methodology, and departures from Code Case N-560 methodology. As a pilot plant application, such a request for additional information (RAl) is typically necessary, i
Enclosed is the RAI regarding your request. The RAI was discussed with Mr. Steve Bennett and others of your staff on April 1,1999. In order to support your request, and our target l
completion date of July 1999, we agreed that your response would be provided by May 17, 1999.
Sincerely, s/ //
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Nicholas D. Hilton, Project Manager, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-313
Enclosure:
As stuted cc w/ encl:. See next page
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Mr. C. Randy Hutchinson Entergy Operations, Inc.
Arkansas Nuclear One, Unit 1 cc:
Executive Vice President Vice President, Operations Support
& Chief Operating Officer Entergy Operations, Inc.
l Entergy Operations, Inc.
P. O. Box 31995 P. O. Box 31995 Jackson, MS 39286-1995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway Director, Division of Radiation P. O. Box 651
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Control and Emergency Management Jackson, MS 39205 Arkansas Department of Health l
l 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 i
Manager, Rockville Nuclear Licensing Framatone Technologies 1700 Rockville Pike, Suite 525 Rockville, MD 20852 Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County Pope County Courthouse Russellville, AR 72801
1 REQUEST FOR ADDITIONAL INFORMATION ARKANSAS NUCLEAR ONE. UNIT 1 RISK-INFORMED INSERVICE INSPECTION ALTERNATIVE TO ASME CODE SECTION XI. TABLE IWB 2500-1 Based on the review of the licensee's submittals, a number of items were identified for which additional information is required to clarify a response or to permit further evaluation, as appropriate. These items are listed as follows:
1.
Regulatory Guide 1.174, as well Standard Review Plan (SRP) 3.9.8, identifies five principles of risk informed (RI) regulations. They are:
(a) Meet current regulations unless explicitly related to a requested exemption or ru!e
- change, (b) Consistent with defense in depth philosophy, (c) Maintain sufficient safety margins, (d) Proposed increase in core damage frequency (CDF) or risk are small and are consistent with the Commission's Safety Goal Policy Statement (See also Question 9), and (e) Use performance measurement strategies to monitor the change.
Please explain how the proposed change to the Arkansas Nuclear One, Unit 1 (ANO-1),
inservice inspection (ISI) program meets the listed five principles. Principle (d) on the potential risk increase can be answered in Question 9.
2.
The licensee should identify or provide a cross-reference between the risk segment identifiers in Tables 3-1,3-2, and 3-3 of Report No. SIR-98-055, " Risk Evaluation and Element Selection in Support of ASME Code Case N-560, ANO-1," and the line number / segment identifiers in Table 4-1 for the consequence category rankings in Calculation No. NSD-024, "ANO-1 N560 Consequence Evaluation," to facilitate review of the final eum locations that were selected.
3.
Section 5.0,"Results," of NSD-024,"ANO-1 N560 Consequence Evaluation," indicates that reactor coolant piping falls into the "High" consequence category because a pipe segment failure causes a loss-of-coolant accident (LOCA) with a conditional core damage probability (CCDP) in the high range. Section 2.2, " Impact Group Assessment,"
also states that "[t]he evaluation is performed such that pipe segments can be qualitatively binned using the bins in Table 2-4 (e.g., High) and/or by their quantitative CCDP value estimated in this evaluation. This allows additional ranking to be performed within a category (e.g., High)."
l Since both small and large LOCA's are ranked as having High consequences, does this statement imply that further ranking within the High consequence segments or during element selection was performed based on the quantitative determination of CCDP? If so, please describe how any additional ranking process was applied in the analysis.
Enclosure i
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, 4.
Code Case N-560, Section 2.6, requires that examination zones shall be selected starting with the structural elements !n e High risk group and working toward the Low risk group, until a total number of structural elements equal to 10 percent of the Category B-J welds, excluding socket welds, has been selected. Report No. SIR-98-055 indicates that there are 394 total elements in the Class 1 piping at ANO-1, and, thus,10 percent would be 40 welds. The examination locations proposed for the ANO-1 RI ISI program consists of 40 welds divided between 28 Highs and 12 Mediums. Since there were 84 High risk elements identified but only 28 selected, does this selection constitute a departure from the N-560 Code Case guidance? If so, the licensee should provide the rationale as to why examination locations were chosen based on belonging to the Medium risk group instead of the High risk group.
5.
In order to conclude that the resu'ts and conclusions reflect the as-built and as-operated plant, please provide a discussion on the following, a.
Code Case N-560 (a)(4) requires that the consequence ranking be performed by a multidisciplinary team of experts. Experience with the maintenance rule and other risk-informed applications has shown that the expert panel should be composed primarily of plant personnel to ensure that sufficient plant-specific experience is available to provide confidence that the results reflect the as-built, as-operated plant. ANO-1's submittal appears to have been primarily developed by two contractors and there was no mention of review and approval by a team of experienced plant personnel. The licensee should describe the makeup of the ISI selection team for the ANO-1 analysis and a description of the process and rationale that was used to classify pipe failure potential and final element exam location selection. In addition, please describe the ANO-1 plant team that reviewed the consequence analysis work and approved the results and conclusions.
b.
The licensee's submittal referenced a "Rev.1" of its probabilistic risk assessment (PRA), but the CDF referenced in Section 3.1 is from a table in the licensee's April
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1 1993 individual plant examination (IPE) submittal. Also, the IPE submittal cover letter indicated the large, early release frequency (LERF) was expected for 2.8 percent of the CDF, or 1.3x104 per year. Please supply the date of the last PRA model up-date and briefly describe how the up-date was reviewed and approved. If the last review of plant data and plant changes determined that a PRA up-date was not necessary, the date of this review may be considered the last PRA model up-date. The corresponding baseline CDF and LERF should also be provided to further define and characterize the PRA model used.
c.
The staff evaluation report on the IPE identified potential weakness in the human error evaluation process. Specifically, the report stated that plant-specific performance shaping factors were not applied and there were no walkdowns performed to support the response times obtained from operator interviews. Please I
clarify how these issues could affect the results and conclusions reported in the submittal, and describe any evaluations performed during the Al-ISI analysis to ensure that any affects are minimized such that the results and conclusions are valid.
. 6.
In order to conclude that there is reasonable assurance that the results and conclusions i
appropriately reflect the application of the methodology to the plant, please clarify the following issues, a.
NSD-024, Section 3.3, " Plant Level Assumptions," item 4, page 18, identifies that breaks in the low-pressure injection (LPI) and core flood piping are limited to l
9 inches by flow restricting orifices in the vessel nozzles. It also states that the ANO-1 IPE did not include a Medium LOCA, but only small and large LOCAs.
Item 5, page 18, indicates that Reference 3 included a me dium LOCA designation but the reference was not used and item 5 concludes that "the present analysis is conservative." lter.: 6, page 18, states that "different size LOCAs should be considered." Finally, the second paragraph under Section 4.3," Impact Group Assessment," page 27, refers to a table on page 3.1-13 in the IPE where medium LOCA success criteria are given (although, as stated earlier, not used in the IPE).
The text on page 27 and the note (4)(b) on page 36 both apparently use the medium LOCA success criteria from the IPE table on page 3.1-13. If success criteria different from the those used in the IPE are used, this should be clearly stated and consistently labeled, and all success criteria should be included in Figure 3-1, " Simplified Success Criteria," in the NSD-024 submittal.
l b.
NSD-024, page 25, credits the operators following the " Reactor Trip Procedure" to i
close CV-1009. Since the pipe rupture will cause a LOCA, will the operators still follow the Reactor Trip Procedure or will they be following another procedure?
Also, item 6, page 18, states that it appears the valve could be closed in time to affect the success criteria but further evaluation could increase confidence. Was this evaluation done? Conversely, the assumption that the letdown isolation valves CV-1213 and CV-1215 do not close automatically following a LOCA, leads to assigning the downstream segments a High consequence. Do these valves close automatically upon LOCA indication and are the operators directed by procedure to isolate letdown following a LOCA?
c.
Are the lines between the pressurizer and the safety valves PSV-1000, PSV 1001, and PSV-1002 included in the analysis?
7.
The treatment of breaks in the LPI lines upstream of DH-14A/B is not clear; please clarify the following issues:
j a.
NSD-024, page 28, states that ruptures in these lines do not fail both LPI trains through flow diversion since there are flow orifices " upstream of the break,"
meaning upstream of DH-14A/87 Flow orifices upstream of the break limit the s.
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of the resulting LOCA but not the LPl/ core flood flow. The figure on page C-17 in th > IPE indicates that it is not possible to isolate one LPI injection line from the other between DH-13A/13B/17/18 and DH-14A/B. Breaks downstream of DH-14A/B are High so the assumption is immaterial. However, for a break upstream of DH-14A/B, why are both LPI trains and one core flood train not lost out the break?
b.
Why is failure of the segment ups,tream of DH-14A/B labeled a Small LOCA in Table 4-1 when, even with the orifice, it should be at least a Medium LOCA?
i o
i l c.
In addition, the CCDP column for this entry uses 1x10 for passive failure of check 8
valve DH-14A or B, but note 5 to the table indicates this value is 1.4x10~8, which, if used, would result in a CCDP of 1.1x104 even if the LOCA CCDP of 7.5x10 from 8
Table 2-1 is used rather than 8x10'8. Therefore, the CCDP would result in a High
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consequence ranking rather than the Medium ranking as noted. Please explain why your plant-specific data was not used for this scenario.
8.
To fully address LERF considerations, use of the generic containment failure probability given core damage of 0.1, must be shown to be applicable to the core damage scenarios caused by these breaks. Please indicate what conditional containment failure
. probability is estimated for the types of core damage scenarios initiated by the ruptures of segments classified as Medium, and compare these to those scenarios which cause the estimated 2.8 percent of large early release following core damage.
9.
SRP 3.9.8 requires that the licensee determine whether any risk increases would result from implementation of the proposed RI-ISI program, and that cumulative effects are small and do not exceed NRC safety goals. In order to conclude that the changes in 1
CDF and LERF are, with reasonable assurance, actual risk decreases or at most, a negligible increase, please include a discussion of the impact of the changed inspection program on these metrics. If a qualitative discussion based on simple and straightforward assumptions does not clearly characterize the change in risk, a i
quantitative bounding estimate or more detailed quantitative evaluation should be provided.
10.
In the original ISI Program Plan, Table 4.1, " Inservice inspection Summary Table," a i
total of 444 B-J componen's are identified. Of these 444 elements,29 are socket welds.
The RI-ISI program (Table 3-4 of SIR-98-055) list a total of 394 elements, not including socket welds. Describe the discrepancy and how the total of 394 was determined.
11.
Qualification of nondestructive examination (NDE) systems (personnel, procedure, and i
equipment)is an important element of the Al-ISI program. The reliability of i
examinations must be established to achieve the desired confidence levels for the i
risk-informed inspection process. Therefore, the technical basis for the inspection l
reliability inputs used in structural reliability calculations of estimated failure probabilities j
must be justified. Such a basis can be provided by NDE performance demonstration programs. It is unclear how NDE methods, procedures, and personnel will be qualified at ANO-1. With respect to this area, provide the following information:
(a) Provide a detailed technical discussion describing how the reliability of NDE performed in the RI-ISI program will be qualified.
(b) Will ANO-1 use the licensee's own examination procedures, equipment, and personnel to perform ultrasonic examinations of selected examination volumes?
(c) If so, are the procedures, equipment, and personnel qualified in accordance with l
the requirements of Appendix Vil of the ASME Code,Section XI?
i (d)' If contractors are used to perform these examinations, are the contractors required l
to use procedures, equipment, and personnel qualified to Appendix Vill of Section XI?
12.
The licensee has selected 40 B-J elements for examination based upon a consequence evaluation and a degradation mechanism evaluation. Appropriate examination methods
. should be selected to address degradation mechanisms, pipe sizes, and material of concern. It appears that the only method of examination to be performed on the subject elements is volumetric examinations. Provide the specific NDE methods to be used for each of the subject elements. Provide information describing why the particular NDE method was chosen as well as why other NDE methods were not chosen (i.e., surface examinations and visual examinations).
13.
The inservice inspection strategy used in the RI-ISI program must define when the inspections are to be performed. Specified inspection intervals must be consistent with relevant degradation rates, inspection intervals should be sufficiently short such that degradation too small to be detected during one inspection does not grow to an unacceptable rlze before the next inspection is performed. Provide a discussion regarding the inspection intervals contained in the RI ISI program, include an examination schedule, and confirm that the proposed examination frequency will not exceed the current Section XI inspection interval of 10 years.
14.
Considering that the implementation of the proposed RI program will significantly reduce the number of examinations, limited examinations could have a significant impact on the risk. The licensee has chosen multiple elements that have not received previous inspections. Have the elements been chosen such that examination limitations do not exist? In the case that examination limitations are encounte,ed on any elements, specify what alternatives will be used to ensure structural integrity.
15.
The proposed RI ISI program contains 40 elements for examination. Considering that j
evaluation of indications is required based on iSI findings, describe the acceptance criteria and the examination expansion philosophy that will be utilized.
16.
Typically, licensee's perform periodic updates to ISI programs. Updates typically consider:
Plant design feature changes, plant procedure changes, and equipment performance changes.
Leakage, flaws, or indications identified during scheduled RI Isl examinations.
Plant and industry failure information.
Describe the intentions of the licensee, concoming Rl-ISI program update frequencies and rN#9 cations (if any) to the RI-ISI plan.
17.
It is understood that the licensee has excluded vessel nozzle to safe-end welds classified as category B-F welds from the evaluation. Are there any examination category B-J dissimilar metal welds between combinations of (1) carbon or low alloy steels to high alloy steels, (2) carbon or low alloy steels to high nickel alloys, or (3) high alloy steels to high nickel alloys, that were used in the evaluation process? If so, how many, and how many have been chosen for examination?
18.
In the licensee's original ISI program, many augmented examinatiens have been listed for performance, it appears that some of the augmented examinations are related to Class 1 elements. However, it is unclear whether any of the augmented examinations are effected / eliminated by the RI-ISI program. Is it the licensee's intention to continue to perform all of the augraented examinations as described in the original ISI program?
.