ML20216C872

Forwards RAI Re Implementation of ASME Code Case N560 at Plant

ML20216C872
Person / Time
Site:	Vermont Yankee
Issue date:	03/11/1998
From:	Croteau R NRC (Affiliation Not Assigned)
To:	Reid D VERMONT YANKEE NUCLEAR POWER CORP.
References
TAC-M99389, NUDOCS 9803160222
Download: ML20216C872 (7)

Text

March 11, 1998 Mr. Donald A. Reid Senior Vice President, Operations Vermont Yankee Nuclear Power Corporation 185 Old Ferry Road Brattleboro, VT 05301

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING IMPLEMENTATION OF ASME CODE CASE N560 AT VERMONT YANKEE NUCLEAR POWER STATION (TAC NO. M99389)

Dear Mr. Reid:

By letter dateu August 6,1997, you informed the NRC that Vermont Yankee had agreed to be the lead plant for the NRC review and approval of ASME Code Case N560. Subsequent information was provided by letters dated August 15 and October 23,1997.

We have determined that additional information is needed to continue our review. Enclosed is our request for this additional information.

Sincerely, Original signed by Richard P. Croteau, Project Manager Project Directorate 1-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-271 -

Enclosure:

Request for Additional Information cc: See next page DISTRIBUTION Docket File" CThomas PUBLIC RCroteau Vermont Yankee r/f TClark JZwolinski WHehl, RI i ACRS OGC #

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I DOCUMENT Name: G:\CROTEAU\VYM99389.RAI To ::ceive a copy of this document, Indicate in the box: "C" = Copy without attachment / enclosure "E" =

Copy with attachment / enclosure "N" = No copy OFFICE PDI 3/PM PDl-3/L4 m l E PDI-3/D [A/ l NAME RCroteau 9/6' TClark.D ET)6tneer DATE 03/// /98 03////98 Os/ / / /98 OFFICIAL RECORD COPY 9803160222 900311 PDR ADOCK 05000271 P DR . g '

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D. Reid Vermont Yankee Nuclear Power Station cc:

Regional Administrator, Region 1 Mr. Raymond N. McCandless U. S. Nuclear Regulatory Commission Vermont Division of Occupational  !

475 Allendale Road and Radiological Health King of Prussia, PA 19406 l Administration Building l Montpelier, VT 05602 Mr. David R. Lewis Shaw, Pittman, Potts & Trowbridge Mr. Gautam Sen 2300 N Street, N.W. Licensing Manager Washington, DC 20037-1128 Vermont Yankee Nuclear Power Corporation _

q Mr. Richard P. Sedano, Commissioner 185 Old Feny Road l Vermont Dopartment of Public Service Brattleboro, VT 05301 '

120 State Street,3rd Floor i

Montpelier, VT 05602 Resident inspector l Vermont Yankee Nuclear Power Station Public Service Board U. S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 120 State Street Vemon, VT 05354 Montpelier, VT 05602 i Mr. Peter LaPorte, Director Chairman, Board of Selectmen ATTN: James Muckerheide  !

Town of Vemon Massachusetts Emergency Management j P.O. Box 116 Agency Vernon, VT 05354-0116 400 Worcester Rd.

P.O. Box 1496 Mr. Richard E. McCullough Framingham, MA 01701-0317 Operating Experience Coordinator Vermont Yankee Nuclear Power Station Jonathan M. Block, Esq.

P.O. Box 157 Main Street Governor Hunt Road P. O. Box 566 Vernon, VT 05354 Putney, VT 05346-0566 G. Dana Bisbee, Esq.

Deputy Attomey General 33 Capitol Street Concord, NH 03301-6g'47 Chief, Safety Unit Office of the Attomsy General One Ashburton Place,19th Floor Boston, MA 02108 Ms. Deborah B. Katz Box 83 She!'burne Falls, MA 01370 l

RFOUFET FOR ADblTIONAl INFORMATION The following information needs to be provided by the licensee in order for the staff to complete its review.

1. N-560 requires an evaluation ofindirect e#ects of postulated pressure boundary failures, including the spatial effects of flood, spray and pipe whip on mitigating systems, and these effects have generally been addressed by the licensee. However, for breaks outside the drywell, the effect of a steam environment on Motor Ccatrol Centers (MCCs) and other electrical components should be evaluated. Show that the analysis of the result of breaks outside containment has considered the effect of a steam environment on MCCs or other electrical switchgear.

2. The example in Section 2.2 for the initiating Event impact Group Assessment includes the I

probability of failure to isolate a break in the reactor water cleanup system due to failure of  !

the inner isolation valve to close on low reactor pressure vessel (RPV) water level (4E-3), but  !

note 3 of Table 2-1 indicates that the initiating event frequency for loss-of-coolant accidents (LOCAs) outside containment already includes the probability of inner isolation valve failure.

This would change the ranking based on conditional core damage probability (CCDP) to high rather than medium. Provide further clarification and an evaluation to determine any changes in the ranking of LOCA-OC events.

For this example (and the analysis that considered isolskn success), provide an evaluation showing that the reactor water clean-up (RWCU) motor-operated valve (MOV) would be able to close in the event of blowdown due to the break.

3. Item 2 under the System impact Group Assessment regarding determination of the number of unaffected backup systems / trains includes a confusing statement: "When considering the consequences, given an isolation failure, the number of available backup trains includes isolation." A similar statement appears in the Combination impact Group Assessment discussion. Clarify what is meant by these statements.

' 4. Table 2-2 of the submittal, which was used as a guideline for assigning consequence categories to pipe failures that result in a loss of systems / trains without an initiating event, differs from the version in N-560 (Table I-5) and the one in the EPRI methodology. In some cases, using the consequence categories of Table 2-2 would result in less conservative rankings than using the corresponding table in N-560 (e.g., for a system responding to an infrequent event with a long AOT and only 1 backup train, the consequence category would be high in the N-560 table but medium in Table 2-2). Although the basis for Table 2-2 is explained in RAI response Ill.8, this would seem to be a departure from the original guidance.

Please explain why it was necessary to depart from the original guidance.

a) With regard to your response to RAI Ill.8 (10/23/g7), explain how the unavailability limits were chosen for each number of backup trains available, and whether the resulting mean values that were used are conservative with respect to values obtained from the W PRA as shown in Table 3 2 of the submittal.

Enclosure

5. Table 2-2 is not based on CCDP from the PRA calculations, but instead relies on estimates based in part on Unavailability Mean Values corresponding to the number of backup trains. It l is not clear how the quantitative estimates account for dependent failures between trains of i the same system or different systems. Please clarify how common cause failures and spatial  ;

impacts (if any) are accounted for. Please also clarify how support system dependencies are addressed. For example, Figure 3-1, Note 2 (for the medium LOCA success criteria) indicates use of the condensate system as a low pressure makeup source is dependent upon l feedwater success. Similarly, residual ht.at removal (RHR) train "A" for decay heat removal is dependent upon LPCI train "A". Please explain how dependencies are treated while generating the quantitative estimates and provide some example calculations.

6. Several questions remain regarding determination of the number of backup trains based on the critical safety functions success criteria as depicted in Figure 3-1 of the submittal:

a) Please clarify the relationship between the " Qualitative Basis"in Table 4-1 and the success criteria in Figure 3-1. Please verify that the success criteria in Figure 3-1 reflect all the entries in 4-1, or provide a set of figures similar to Figure 3-1 with the appropriate success criteria.

b) Note 3 of Figure 31 indicates that the W PRA does not credit the recovery of the main condenter for MLOCA. Identify why recovery of the main condenser is credited as a backup in this figure.

c) Success criteria for a small LOCA indicate reactor core isolation cooling (RCIC) is sufficient backup for high pressure makeup. Most BWR studies consider that there are some small LOCAs that RCIC alone would be insufficient to make up. Clarify why this success criteria conflicts with the discussion for LOCAs in Section 4.2 of the submittal since, in that Section, "small LOCA" is defined as the pipe size that, if broken, still allows normal makeup capabilities from 1 CRD pump Ed RCIC to maintain reactor waterlevel.

7. In Table 4-1, it is not clear how the corresponding CCDP estimate for the Table 2-3 qualitative basis evaluations was obtained from Table 3.1 and 3.2. Please describe how this analysis is done and provide some example calculations.

8. The discussion on containment perfomance in Sections 2.2 and 3.2 of the submittalis based upon the EPRI methodology rather than being an N 560 requirement. The analysis in Section 3.2 regarding containment performance for the high pressure makeup function is confusing.

Further clarification of the basis for the containment performance analysis is requested,

g. You state that the in.sirect effects of pipe ruptures inside the drywell we not explicitly considered since the equipment inside the drywell is environmentally qualified for the LOCA i steam and flooding environment which is within the design basis. This is correct, but the j proper operation of the pressure suppression and heat removal systems is required to 1 maintain the design basis conditions, in particular, the drywell to wet well vent lines, the l drywell wetwell vacuum breakers, the ADS piping, and the containment atmospheric vents, i are to varying degrees relied upon to mitigate LOCAs and maintain conditions within the drywell within the design basis. Please include an evaluation of welds within these systems.

Improved inspection of these items may be usefulin offsetting any poientialincreases in .;sk  ;

due to the reduced number of inspection of Class 1 welds.

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10. Section 3.1 of the Vermont Yankee August 6,19g7 submittal, states that the PRA prepared to support the Vermont Yankee individual Plant Examination (IPE) was used to support the PRA portions of the submittal. Section 2.4 of the W IPE states that the IPE reflects the l plant configuration as of December 1,1993. The staff notes that the recent maintenance rule program inspection report for Vermord Yankee (Inspection Report Number 50-271/97-

81) states that the Core Damage Frequency (CDF) value in the W PRA has increased from i the IPE value of 4.3E 6 to a current CDF value of 5.0E-6.

Please verify that all plant configuration and operational changes that have occurred since December 1,1993, have been reviewed and a determination made that no PRA model changes were needed to support the 181 submittal, or that model changes were made as needed to support the ISI submittal. Additionally, describe any changes between the PRA prepared for the IPE and the PRA being used to support the ISI program consequence  !

analysis evaluation.  ;

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11. Please clarify whether W intends to periodically modify its ISI program based on risk insight and some W inte' mal process, or whether an approved risk informed ISI program will not be changed without prior NRC approval.

12. In your response to our question ll-7 you referred to your Operational QA Program (YOQAP).

Please provide us the documented results of your independent review of the analysis in your submittal. This documentation is the documented review required by Section Ill, " Design Control," of your YoQAP (e.g., the review required by Section 6 of ANSI N45.2.11 as endorsed by RG 1.164). Also note that it is the licensee's responsibility to justify that the quality of the PRA is adequate for the proposed applicafjon. Attributes of PRA quality which should be addressed are illustrated in the Draft SRP Chapter 19 issued for public comment on 6/25/97. Processes to support a licensee claim of adequate quality include structured and documented peer review, cross comparison studies, and industry certification. Please describe the process you have employed (and provide all related documentation) or that you will employ to substantiate the quality of your PRA.

13. In RAI VI.1, an estimate of the impact on core damage frequency (CDF) and large early release frequency (LERF) of the proposed change is made based on bounding estimates.

The basis for these estimates requires further clarification, e.g., provide the basis for the statement that likelihood of pressure boundary failure (PBF) for a pipe location with no degradation mechanism present (xo) is expected to have a value lower than 1E-8. Clarify how these estimates address the frequency of LERF.

14. The licensee has argued that the probability of detecting flaws will be greatly enhanced if inspections are targeted at specific degradation mechanisms. To this end, W has proposed to provide further training to inspection personnel to increase their knowledge of the subject mechanisms, and the type of ultrasonic indications to be cognizant of when performing inspections under N-560. However, past experience with the intergranular stress corrosion cracking (IGSCC) phenomena would indicate that, while an increased level of causal knowledge is helpful, it is not sufficient to validate the reliability of the NDE method. It is unclear how the licensee intends to qualify the ultrasonic methods, procedures, and personnel used to perform an " inspection for cause" at W.

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15. The potential degradation mechanisms applied to W piping components are stated to be l

l the result of industry and plant-specific surveys. With respect to this statement provide the ;

following information:

a) It is unclear which databases have been reviewed to assess potential degradation at i W. What was the extent of the review of these databases? Are any of these l l

databases considered to be a consensus or industry standard for describing potential  !

degradation mechanisms in light water commercial nuclear facilities?

I b) N-560 requires that examinations be performed commensurate with expected  !

degradation mechanisms, however no analysis is required to determine an optimum inspection frequency - an art >itrary schedule of every ten years is listed. -it would appear that an N-560 degradation mechanism evaluation is incomplete with respect to initiation and growth parameters, and how these might impact inspection frequencies.  ;

i Provide information as to how those parameters are being addressed at W. '

c) In the licensee's response to RAI, it was stated that the frequency of occurrence of new or unknown damage mechanisms are bounded by that of known failure mechanisms, based on EPRI studies. Further, that EPRI will continue to monitor i service experience and, if new mechanisms are discovered, strategies can be  !

developed to address them. However, the response did not identify any program or l formal vehicle in place to impart newly acquired EPRI information pertaining to  !

degradation mechanisms to W. It is the licensee's responsibility to integrate new information that could affect piping inspection strategies into the W RI-ISl program.

The licensee has indicated that new information will be reviewed via a formal

" Operating Experience" program, however no description of how this program correlates industry failure history, nor how a documented revision to the bases for the N-560 attemative will be performed is included in the licensee's submittal. This information needs to be provided. I d) The N-560 evaluation requires that licensee's elevate the risk ranking of piping segments where a potential for water hammer exists. The licensee's statement that

" appropriate measures" are in-place to preclude water hammer is unclear. If this determination is being made as a result of administrative controls versus engineered safety functions, then some probability for water hammer should be considered. This may impact the ultimate risk ranking of several piping ssgments. The licensee should describe what appropriate measures have been put in place to eliminate water hammer as a potential in applicable systems.

16. Attachment 2 of the W submittal entitled " Mechanisms Specific Examination Volumes and Methods" documents the basis for the selection of examination volumes and methods that will be utilized at W. However, no information is provided pertaining to qualification of examination procedures, examination equipment and personnel. In order to ensure an adequate examination of the selected examination volumes the examination procedures, equipment and personnel must be property qualified. With respect to this area provide the following information:

a) Will W use its own examination procedures, equipment and personnel to perform ultrasonic examinations of selected examination volumes?

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1 b) If it does, are the procedures, equipment and personnel qualified in accordance with '

l the requirements of Appendix Vill of the ASME Code,Section XI?

l c) If contractors are used to perform these examinations, are the contractors required to use procedures, equipment, and personnel qualified to Appendix Vill of Secticn XI?  !

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