ML081760042

Meeting Minutes of the ACRS Subcommittee on the Safety Evaluation Report Related to the License Renewal Application for McGuire Nuclear Station Units 1 and 2, Catawba Nuclear Station Units 1 and 2, October 8, 2002

ML081760042
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	10/11/2002
From:	Apostolakis G Advisory Committee on Reactor Safeguards
To:
References
Download: ML081760042 (95)

Text

October 11,2002 MEMORANDUM TO: John T. Larkins Executive Director FROM:

George E. Apostolakis, Chairman

SUBJECT:

CERTIFICATION OF THE MINUTES OF THE PLANNING AND PROCEDURES SUBCOMMIITEE MEETING - OCTOBER 9, 2002 I hereby certify that, to the best of my knowledge and belief, the minutes of the sUbject meeting, issued October 11, 2002, are an accurate record of the proceedings for that meeting.

George E. Apostolakis, Chairman Planning and Procedures Subcommittee October 11, 2002

CERTIFIED 10/11/02 G. Aposto1akis, Chmn INTERNAL USE ONLY G:PlanPro(ACRS):ppmins.496 October 10, 2002

SUMMARY

MINUTES OF THE ACRS PLANNING AND PROCEDURES MEETING OCTOBER 9, 2002 The ACRS Subcommittee on Planning and Procedures held a meeting on October 9, 2002, in Room T 2 B3, Two White Flint North Building, Rockville, Maryland. The purpose of the meeting was to discuss matters related to the conduct of ACRS business. The meeting was convened at 1:30 p.m. and adjourned at 4:10p.m.

ATTENDEES MEMBERS M. Bonaca T. Kress ACRS STAFF J. T. Larkins S. Bahadur H. Larson S. Duraiswamy R. P. Savio J. Gallo S. Meador NRC STAFF I. Schoenfeld, OEDO

1)

Review of the Member Assignments and Priorities for ACRS Reports and Letters for the October ACRS meeting Member assignments and priorities for ACRS reports and letters for the October ACRS meeting are attached (pp. 10-13). Reports and letters that would benefit from additional consideration at a future ACRS meeting were discussed.

RECOMMENDATION The Subcommittee recommends that the assignments and priorities for the October 2002 ACRS meeting be as shown in the attachment (pp. 10-13).

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2)

Anticipated Workload for ACRS Members The anticipated workload for ACRS members through December 2002 is attached (pp.

10-13). The objectives are to:

Review the reasons for the scheduling of each activity and the expected work product and to make changes, as appropriate Manage the members' workload for these meetings Plan and schedule items for ACRS discussion of topical and emerging issues During this session, the Subcommittee also discussed and developed recommendations on the item included in Section II of the Future Activities List (p. 14).

RECOMMENDATION The Subcommittee recommends that the members provide comments on the anticipated workload. Changes will be made, as appropriate. The Committee should decide on the Subcommittee's recommendations on the item in Section II of the Future Activities List.

3)

Foreign Travel Update Since last ACRS meeting, the final travel arrangements for the October foreign travel have been made and a detailed itinerary has been put together by Tanya Winfrey. For the Germany leg of the trip (Quadripartite Meeting), the cost of the sleeping rooms and an ancillary fee have been prepaid by the Government in the form of a registration fee for both the members and staff. Therefore, members should not have to pay any rooming charge for their stay in Germany during the week of the Quadripartite Meeting as it has been prepaid.

The technical papers for the Quadripartite Meeting have been sent to the RSK and the Commissioners. The slides for presenting those papers are done. Sherry Meador is coordinating the translation so that both English and German versions of the presentations and overheads are available at the meeting in Germany.

RECOMMENDATION The Subcommittee recommends that each member review a copy of the detailed travel itinerary. Comments or changes should be provided to Jenny Gallo immediately.

The slides for the Quadripartite Meeting have been completed and forwarded to be translated. The PowerPoint presentation will be forwarded to the RSK with a backup copy on CD to be available.

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4)

Celebration of the 500th ACRS Meeting As agreed to by the members, invitations were sent to the NRC Commissioners to participate at the 500th ACRS meeting ceremony, which is scheduled for March 4-5, 2003. (This is also coincidental with the Committee's 50th Anniversary.) NRC Chairman Meserve and all Commissioners as well as Bill Travers, EDO, have agreed to participate. Invitations were also sent to those who are expected to serve as panel members.

Drs. Hal Lewis, Robert Seale, Bill Stratton, J.Ernest Wilkins, Stephen Hanauer, and Mr.

Dave Ward, have agreed to serve on the panels. Because of his health problems, Dr.

Okrent has decided not to be a lunch time speaker. Dr. Remick has agreed to be a lunch time speaker.

As decided by the Committee at the June 2002 meeting, invitations were sent to Mr.

Ralph Beedle, NEI, and Mr. Bert Wolf, GE, to participate in the celebration. They agreed to participate as Panel members.

A letter has been forwarded to all of the Panel participants and speakers thanking them for their willingness to participate in the 500th meeting ceremony. Additionally, accommodations are being arranged for all of the guests to stay in the area along with developing transportation from the hotel to the White Flint building.

RECOMMENDATION The Subcommittee recommends that the Executive Director keep the Committee informed of further developments, including the status of obtaining funding from the agency. The possibility of the Committee sponsoring a reception at the end of the first day was discussed and it was recommended that the Full Committee consider this matter after the staff collects information on cost.

5)

Role and Use of PRA in the Regulatory Decisionmaking Process Mr. Karl Fleming of Technology Insights has started to plan for the Committee's white paper" addressing the role and use of PRA in the regUlatory decisionmaking process.

He is in Rockville, Maryland, and plans to conduct interviews with NRC staff on October 10-11, 2002, hearing their views on what needs to be done to enhance PRA submittals.

Mr. Fleming will present a draft plan for researching and compiling the information during the Saturday session of the October full Committee meeting. To the extent time permits, he will meet with members individually during the October meeting. Dr.

Hossien Nourbakhsh has been designated as the Project Manager for this activity and will work closely with the contractor to guide this effort.

RECOMMENDATION The Subcommittee recommends that members who would like to provide individual input into this project contact Dr. Nourbakhsh to coordinate a conference call with the contractor.

4 During Mr. Fleming's briefing at the October ACRS meeting regarding his plans for developing a draft 'White Paper," the members should provide feedback to Mr. Fleming on his plans.

6)

Meeting with the EDO The Planning and Procedures Subcommittee plans to meet with the EDO and the Deputy EDOs (DEDOs) on Friday, October 12, 2002, between 12:00 and 1:00 p.m. to discuss several issues, including the following:

ACRS/NRC staff coordination Adequacy of the NRC staff's review of power uprate applications License Renewal Issues High Burnup Fuel Issues Revision 1 to Reg. Guide 1.174 Significant issues that the NRC staff expects to submit to the ACRS for review in the next two years.

DPV regarding proposed 10 CFR 50.69 (pp. 15-16)

As suggested by the Committee during its September meeting, a formal meeting between the EDO/DEDOs/Office Directors and ACRS will be scheduled for a future ACRS meeting.

RECOMMENDATION The Subcommittee recommends that the Chairman report to the Committee on the results of the above meeting.

7)

Quad Cities Unit 2 - Damage to Steam Generator Dryer Resulting from Power Uprate Operation (Open)

On July 11, 2002, Quad Cities Unit 2 was shut down to investigate the irregularities in the steam flow, reactor pressure and level, and moisture carryover in the main steamlines. The results of the investigation revealed that a cover plate of the steam dryer was missing. Subsequent investigation led to the identification of pieces of the cover plate. A large piece was found in the Separator and there were indications that pieces had been transported into "A" main steamline and the vessel. Most of the pieces have been retrieved.

The cause of the damage is believed to be due to high-cycle fatigue induced by the cover plate natural frequency, nozzle chamber standing wave acoustic frequency, and vortex shedding frequency -- all coinciding at 180 Hz.

5 During the September 2002 ACRS meeting, the Committee asked Drs. Ford and Ransom to review this matter and propose a course of action, noting any generic implications (pp. 17-19).

RECOMMENDATION The Subcommittee recommends that Drs. Ford and Ransom provide a progress report on their assignment at the October meeting. They should develop a course of action for consideration by the Subcommittee and the full Committee during their November 2002 meetings.

8)

ACRS Senior Fellow A contract for one of the ACRS Senior Fellow positions has been awarded to Link Technologies, Inc. The company will provide the equivalent of one full time employee (FTE) over the course of next year. The FTE is equivalent to 2087 hours0.0242 days <br />0.58 hours <br />0.00345 weeks <br />7.941035e-4 months <br /> of manpower.

The company has proposed the use of eight individuals who offer a wide array of expertise. The individuals are Dr. John Austin, Ali Tabatabai, Dr. Spyros Traiforos. Dr.

Bernard Snyder, Phillip McKee, Jeff Woody, Charles Haughney, and Peter Kiang. In order to effectively utilize this contract, a work plan is being developed. The work plan will coincide with high priority topics under review by the Committee. Since the contract has been executed effective September 30, a plan needs to be put in place as soon as possible preferably by October 30,2002. Dr. Savio, in consultation with Dr. Bahadur, has been designated as the Project Manager. Suggested topics for the work plan should be provided to Dr. Savio.

RECOMMENDATION The Subcommittee recommends that members identify topics that they would like to have included in the work plan. The Executive Director in consultation with the Planning and Procedures Subcommittee will prioritize the Tasks for Link Technologies, Inc. and provide periodic status reports to the Full Committee.

9)

Proposed Tasks for Dr. Nourbakhsh, ACRS Senior Fellow During the September meeting, the Subcommittee proposed the following tasks for Dr.

Nourbakhsh and recommended that members propose other tasks:

Review NUREG-1150 to see if parameter and model uncertainties can be extracted from the overall uncertainty in order to have an estimate of just the model uncertainty contribution. This estimate could then be used in regulatory decisions involving PRA results that include only parameter uncertainty. [May need to use current PRAs with parameter uncertainty quantified for the NUREG-1150 reference plants.]

The ACRS has proposed that frequency-consequence (F-C) curves could replace or supplement CDF and LERF as a risk-acceptance metric that would capture the full range of potential radioactivity releases and be

6 made consistent with the safety goals. A White paper is needed to flesh out" this proposal:

What Consequence to Use? Would it be TEDE?

What are the F-C values produced by PAAs?

How are these related to CDF and LEAF?

What would be a reasonable 3-region set of curves to use as risk acceptance values?

On September 29,2002, Dr. Wallis suggested that Dr. Nourbakhsh look at the history of the "momentum equation" in RELAP and other codes and advise the Committee about what should be done.

Dr. Nourbakhsh has provided his initial thoughts on the feasibility of extracting parameter and model uncertainties from NUAEG-1150 overall uncertainty (pp. 20-28).

Also attached are e-mails from Drs. Kress and Apostolakis regarding Hossein Nourbakhsh's work (pp. 29-30)

RECOMMENDATION The Subcommittee agrees with the conclusion reached by Dr. Nourbakhsh that separating out parameter and model uncertainties, while possible in principle, is somewhat an overwhelming task and the direct application of the results to any specific plant would be highly questionable. The Subcommittee recommends the following:

Dr. Nourbakhsh need not have to pursue this task.

Pursue the alternative tasks proposed in the Feasibility Memorandum by Dr. Nourbakhsh.

Pursue the task on F-C curves.

The priority recommended by the Subcommittee for Dr. Nourbakhsh's tasks is as follows:

AP1000 PRA F-C Curves Alternative tasks proposed by Dr. Nourbakhsh History of the "momentum equation" in AELAP and other codes

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10)

ACRS Retreat for 2003 Due to budget limitations, the ACRS retreat is scheduled to be held in Rockville on January 23-25, 2003. The topics below have been proposed by the Planning and Procedures Subcommittee and Dr. Powers.

Topics proposed by the Planning and Procedures Subcommittee include Dr. Fleming's draft 'White Paper" on PRA.

Mr. Rosen's report on Davis-Besse Certification process for advanced reactors Member issues (process, ACRS practice)

SUbsequently, Dr. Powers suggested that the following issues associated with ACRS process and procedures be discussed at the retreat (pp. 31-42):

ACRS effectiveness and self-assessment Proliferation of Subcommittee meetings Compensation process ACRS strategy for reviewing technical issues -- proactive vs. reactive Member assignments and Subcommittee responsibilities Discussion and resolution of differing technical views among ACRS members in areas such as risk-informed regulatory process at the meetings instead of via electronic medium.

As requested by the ACRS Chairman, Dr. Bonaca, ACRS Vice Chairman, discussed the issues raised by Dr. Powers during the Subcommittee meeting.

RECOMMENDATION The Subcommittee recommends the following:

Since the agency is currently operating under "continuing Resolution" and is expected to operate under this condition until at least the first quarter of next year, the Committee should defer its retreat until 2004.

The Committee should start the February 2003 meeting a day earlier to discuss the process issues raised by Dr. Powers, other member issues, as well as the ACRS self-assessment.

11 )

ACRS Meeting Dates for CY 2003 Proposed ACRS meeting dates for CY 2003 were provided to the members during the September ACRS meeting. Changes proposed by the members (pp. 43-47) are included in the attached calendar (pp. 48-59). The Committee needs to decide on these dates during the October ACRS meeting. [Note: In response to members' comments the June meeting has been moved from June 4-6 to June 11-13, and the September meeting from 3-5 to 11-13.]

8 RECOMMENDATION The Subcommittee recommends that the Committee discuss the changes proposed by the members and approve the dates for CY 2003 ACRS meetings.

12)

Tour of Vender Facilities Members and Consultants of the Thermal-Hydraulic Phenomena Subcommittee toured the GE facilities in San Jose, CA, on September 23-24 and the Framatome ANP Richland facilities in Richland, WA on September 25-26, 2002, to obtain information on the details of fuel and reactor core design methodology for use by the Committee in reviewing core power uprate applications. Dr. Apostolakis suggested that Dr. Wallis provide a report to the Committee on this matter.

RECOMMENDATION The Subcommittee recommends that, as suggested by Dr. Apostolakis, Dr. Wallis provide a brief report to the full Committee at the October 2002 meeting regarding tour of the GE and Framatome facilities.

13)

Tour of Global Nuclear Fuel Cycle Facility The members of the joint Subcommittee of ACRS/ACNW (Kress, Garrick, and Levenson) are scheduled to tour the Global Nuclear Fuel Cycle Facility in Wilmington, NC, on November 5, 2002. The ACRS members of the joint Subcommittee Drs. Kress and Powers decided not to attend this tour. Since only two ACNW members plan to attend this tour, the Subcommittee needs to decide whether it is worthwhile having this tour.

RECOMMENDATION Because of budget constraints due to the agency being under a continuing resolution, the Executive Director has recommended that this tour be deferred to some future time when funds become available. The Subcommittee agreed with the recommendation of the Executive Director.

14)

Staff Response to Dr. Kenneth D. Bergeron Regarding TVA's License Amendment Request to Produce Tritium at the Watts Bar Nuclear Power Plant In a letter of September 13, 2001, Dr. Kenneth Bergeron, formerly associated with the Sandia National Laboratories, and now a member of the public, expressed concern about the NRC staff's review of TVA's license amendment request to irradiate tritium producing burnable absorber rods at the Watts Bar Nuclear Power Plant. As suggested by the Committee at the October 2001 meeting, Dr. Larkins sent a memorandum to the EDO on October 18, 2001 (pp. 60-65) transmitting Dr. Bergeron's letter and requesting that the EDO keep the ACRS informed of the staff's disposition of Dr. Bergeron's concerns.

(

9 The staff sent a letter to Dr. Bergeron on September 6, 2002 (pp. 66-91) responding to his concerns. The staff also sent a memorandum to Dr. Larkins informing him of the staff's response to Dr. Bergeron.

15)

Meeting with Laurence Williams. Nil As requested previously by the Planning and Procedures Subcommittee, Dr. Larkins has contacted the Nil Chief Inspector's office in the UK concerning a meeting with Laurence Williams, Chief Inspector for Nil, and the ACRS (p. 92). We have been notified by the Nil Chief Inspector's Technical Support Staff that Mr. Williams will be in the U.S. in December and more specifically in the Washington area on the 5th of December, and it may be possible to spend some time with some of the ACRS members. Dr. Larkins will e-mail the Chief Inspector's office and try to set up a time for a meeting with some of the ACRS members during the 498th meeting, December 5-7,2002. These discussions will center around the certification of the AP1 000 design. As we understand it, the UK has considered building two advanced reactor plants and is currently focusing on the Westinghouse AP1000. As such, they are interested in the certification process required by the NRC, including ACRS involvement.

RECOMMENDATION The Subcommittee recommends that Dr. Larkins keep the members informed as to the scheduling of this matter.

16)

Member Issues Implications of Boric Acid Dr. Kress sent an e-mail (p. 93) stating that the Davis-Besse event raises broader concerns than the safety culture issue and he believes that is "boric acid" issue. He states that boric acid is used simply for convenience and flexibility in controlling the reactivity as burnup proceeds during a fuel cycle. He believes it is not really needed.

BWRs do very well without boric acid and by using burnable poisons (gadolinium), and PWRs could do just as well. He suggests that the Committee urge the staff to take a much broader look at the implications of boric acid than just what happened at Davis Besse.

Mr. Sieber does not agree with Dr. Kress' idea about eliminating the use of boric acid in PWR coolant systems (p. 94). His views are included in the attached e-mail (pp. 93-94).

ANTICIPATED WORKLOAD October 10-12, 2002 AVAIL.

LEAD ENGINEER!

BASIS FOR LEAD ISSUE PRIORITY OF BACKUP REPORT PRIORITY MEMBER BACKUP DRAFTS Guidance for Performance-Based To meet the CTM Regulation Larson A

Apostolakis schedule License Renewal Application for Catawba To meet the CTM KobetzlDuraiswamy A

Bonaca Leitch

[Interim schedule letter as needed]

and McGuire Units 1 and 2 Overview of ESBWR, SWR 1000, and Savio ACR 700 Pre-application review Kress Policy issues related to advanced reactor licensing EI-Zettawy Significant recent operating events Weston Leitch A

Response to a and Cancellation of revision 4i of SPAR LPSD SPAR Model Development Plan Duraiswamy Powers Commissioner's Model request B

-~

To provide ACRS EI-Zeftawy Confirmatory Research Program on High-Kress views Burnup Fuel

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ANTICIPATED WORKLOAD November 7-9, 2002 LEAD BACKUP MEMBER Apostolakis Ford Kress Leitch Bonaca Powers LEAD ENGINEER!

BACKUP Savio Savio Weston EI-Zeftawy Savio EI-Zeftawy/Boehnert Weston KobetzlDuraiswamy KobetzlDuraiswamy ISSUE Safeguards and Security activities Status of annual ACRS report to the Commission on the NRC Safety Research Program Proposed Resolution of GSI-189, "Susceptibility of the Ice Condenser and Mark III Containments to Early Failure From Hydrogen Combustion During a Severe Accident" Early Site Permit Process Advanced Reactor Research Plan Westinghouse AP1000 Design Significant recent operating events Peach Bottom License Renewal application Draft Final ANS External Events Methodology Standard [Tentative]

PRIORITY Report to be complet edin December A

B A

[Interim letter as needed]

A BASIS FOR REPORT PRIORITY AVAIL.

OF DRAFTS To meet the CTM schedule To provide early feedback To meet the CTM schedule To meet the CTM schedule

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ANTICIPATED WORKLOAD November 7-9, 2002 LEAD BACKUP MEMBER Apostolakis Ford Kress Leitch Bonaca Powers LEAD ENGINEER!

BACKUP Savio Savio Weston EI-Zeftawy Savio EI-Zeftawy/Boehnert Weston KobetzlDuraiswamy KobetzlDuraiswamy ISSUE Safeguards and Security activities Status of annual ACRS report to the Commission on the NRC Safety Research Program Proposed Resolution of GSI-189, "Susceptibility of the Ice Condenser and Mark III Containments to Early Failure From Hydrogen Combustion During a Severe Accident" Early Site Permit Process Advanced Reactor Research Plan Westinghouse AP1000 Design Significant recent operating events Peach Bottom License Renewal application Draft Final ANS External Events Methodology Standard [Tentative]

PRIORITY Report to be complet edin December A

B A

[Interim letter as needed]

A BASIS FOR REPORT PRIORITY AVAIl.

OF DRAFT'S To meet the CTM schedule To provide early feedback To meet the CTM schedule To meet the CTM schedule

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ANTICIPATED WORKLOAD DECEMBER 5-7, 2002 LEAD MEMBER Apostolakis BACKUP LEAD ENGINEER!

BACKUP Larson ISSUE Draft Final Regulatory Guide DG-1122, "Determining the Technical Adequacy of PRA Results for Risk-Informed Activities" PRIORITY A

BASIS FOR REPORT PRIORITY To meet the CTM schedule AVAIL.

OF DRAFTS Ford Kress Leitch Sieber Bonaca Savio Weston Savio EI-Zeftawy KobetzlDuraiswamy Weston Safeguards and Security Activities Vessel head penetration cracking and Vessel head degradation Draft annual ACRS report to the Commission on the NRC Safety Research Program Policy issues related to advanced reactor licensing North Anna and Surry license Renewal Application [Final Review)

Significant recent operating events A

A A

A To provide early advice to the Commission To meet the CTM schedule To meet the CTM schedule Ransom Wallis Wallis Boehnert Boehnert Resolution of GSI-185, "Control of Recriticality Following Small Break LOCAs in PWRs" Framatome ANP Richland, Inc.

S-RELAP5 Realistic Large-Break LOCA Code A

A To meet the CTM schedule To meet the CTM schedule

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II.

ITEM REQUIRING COMMITIEE ACTION

1.

Draft Final ANS External Events Methodology Standard (Open) (DAPITJK/SD)

ESTIMATED TIME: 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

Purpose:

Determine a Course of Action Review schedule specified in CTM [B. Budnitz, ANS/N. Chokshi, RES]. The NRC staff previously requested the American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) to develop Standards for use by industry in standardizing and upgrading their PRAs to facilitate risk-informed decisionmaking. ASME developed a Standard on internal events which the Committee reviewed and provided comments in letters dated March 25, 1999, and July 20. 2000.

The Committee reviewed the draft ANS Standard on external events PRA and /

provided comments in a letter dated February 9. 2001. The public comment /

period for the draft ANS External Events Methodology Standard closed in April 2001. Significant comments have been received during the public comment period and ANS is in the process of resolving these comments. The revised Standard, which reflects incorporation of pubic comment, will be resubmitted to the ANS Committee for approval. ANS has reconciled all issues resulting from the resolution of public comments and requested to brief the full Committee in November 2002.

[The Planning and Procedures Subcommittee recommends that Dr. Powers propose a cause of action. ]

1

October 3, 2002 MEMORANDUM TO: John Larkins FROM:

August W. Cronenberg

SUBJECT:

Synopsis Concerning Differing Professional View (DPV) Regarding Proposed 10CFR50.69 Rulemaking ("Risk Informed Categorization and Treatment of os Systems, Structures, and Components for Nuclear Power Plants")

Synopsis: On Sept. 26, 2002 three (3) NRC/NRR staff engineers filed DPVs concerning the proposed rulemaking on 10CFR50.69, sometimes referred to as Option-2. The three parties are Mr. David C. Fischer, Mr. Thomas Scarbrough, and Mr. John R. Fair, all senior engineers with the Mechanical & Civil Engineering Branch, Div. of Engineering, within NRR. All three participated in the development of Option 2 and thus quit familiar with the issues involved.

The central focus and commonality of their concerns relate to the treatment requirements for components classified in the RISC-3 class (Safety-Related/Low Safety Significance). Although each provide somewhat different arguments and supporting documentation for their case, commonality is evident in each of the DPV's. Each essentially asserts that the treatment requirements for components classified as RISC-3 under the current language of 10CFR50.69 are at such a high level that they are vague, therefore not sufficient to reasonably assure the functionality of RISC-3 classified components. Each DPV also asserts that the original language of the rule was altered in the "Concurrence Process", so that staff input to assure RISC-3 component functionality was largely eliminated.

In a one-page memo attached ("Changes to 10CFR50.69"), staff engineer Cronenberg alerted members of the Reliability and PRA Subcommittee as to redline/strikeout changes to the proposed 10CFR50.69 rule resulting from the Concurrence Process. That memo indicated that the intent of these changes was to relax requirements to meet consensus standards and control measures in the draft rule. A prior E-mail of 8-28 also indicated that the ACRS might have a problem with such changes. A copy of that Memo is attached.

cc:

S. Duraiswamy S. Bahadur P&P Members...Apostolakis, Bonaca, Kress

UNITED STA"rES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C. 20555-0001 urnrs Aug. 29,2002 (Overnight FEDX)

MEMORANDUM TO:

Reliability and PRA Subcommittee, Apostolakis, Bonaca, Powers, Rosen. Ford, Shack, Kress FROM:

August W. Cronenberg, Senior Staff Engineer

SUBJECT:

Changes to 10CFR50.69 The purpose of this memorandum is to forward to PRA subcommittee members recent redline/strikeout changes to the proposed 10CFR50.69 rulemaking on Risk-Informed Categorization of SSCs. Such changes result from the Concurrence Process, per input from the LT (Leadership Team) and ET (Executive Team). The intent of these changes is to relax somewhat requirements to meet consensus standards and control measures in the draft rule.

Per my prior E-mail of 8-28

• the ACRS might have a problem with such changes.

Attached are the following:

a) Rulemaking Issue Notation Vote (note: strikeout and vertical lines indicating additions) b) Federal Register Notice of Rulemaking for 10CFR50 (only pages with strikeout are attached) c) Note From Chris Grimes on Rule Changes (also sent as attachment in prior e-mail of 8-28) cc:

Bahadur, Duraiswamy. Larkins, Larson

/b

Page 1.

• • .~-~X~_*_**.*~*_*.*.

"~.,'.."', "y'W" Received: from igate.nrc.gov by nrcgwia.nrc.gov; Sun, 06 Oct 2002 13:57:20 -0400 Received: from nrc.gov by smtp-gateway ESMTPce id NAA15750; Sun, 6 Oct 2002 13:55:03 *0400 (EDT)

From: FPCTFord@aol.com Received: from FPCTFord@aol.com by imo-m07.mx.aol.com (maiLout_v34.13.) id 6.f2.23026960 (4446);

Sun, 6 Oct 2002 13:58:11 -0400 (EDT)

Message-ID: <f2.23026960.2ad1 d3b3@aol.com>

Date: Sun, 6 Oct 200213:58:11 EDT

Subject:

Quad Cities Steam Dryer To: apostola@MIT.edu, mvbonaca@snet.net, PAB2@nrc.gov, TSKress@aol.com, JTL@nrc.gov, dapower@sandia.gov, ransom@ecn.purdue.edu, historyart@computron.net, GMLeitch@aol.com, wjshack@anl.gov, JDSIEBER@aol.com, Graham.B.Wallis@Dartmouth.edu MIME-Version: 1.0 Content-Type: texVplain; charset="ISD-8859-1*

Content-Transfer-Encoding: quoted-printable X-Mailer: AOL 5.0 for Windows sub 138 Subject; Quad Cities Unit 2 Steam Dryer Failure At the September P & P meeting I was asked to follow up on the flow induced vibration (FIV) failure of the Quad Cities Unit 2 steam dryer which occurred soon after the plant had initiated power uprate operations.

I requested that this particular FIV topic be added to the agenda of a meeting that was held in San Jose, September 22-23. This meeting was hosted by GE Nuclear Energy for our thermal-hydraulics subcommittee in order to discuss various thermal-hydraulic codes relevant to power uprate applications. Consequently there was not time for a detailed discussion of the steam dryer design features, etc. but, nevertheless, some useful information was given. The salient points are given below.

1. The steam dryer is regarded as a non-safety related component. This status was determined in the industry report VIP-06 "Safety Assessment of BW R

Internals" which was approved by the NRC in an SER in September 1998. In tha t

report it was determined that even if various sub-assemblies in the dryer cracked (e.g. the support and hold down brackets) structural integrity would be maintained even if there was a pressure transient due to a main steam lin e

break. It was assumed that cracking would be discovered during inspection when the dryer was removed during refueling outages and that this frequency was sufficient to minimize the danger of component failure during operation.

Even if a loose part was created during operation the NRC-approved VIP*06 report stated there was no danger to reactor safety; this latter assumption was challenged at our meeting (for instance the consequence of loose part damage to the MSIV or the jet and recirculation pumps), with no definitive resolution..

2. Cracking has been observed in steam dryer assemblies "numerous times";

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these have included the dryer hood, drain line, tie bar, support bracket, dryer cover plate. The cracking mechanisms have been a mixture of intergranular stress corrosion cracking and transgranular fatigue. Where loose parts have been involved, their disposition has been in agreement with GE's prediction and they have given no safety concern. In no case have thes e

instances led to plant shutdown, until the incident at Quad Cities Unit 2.

3. In many cases the utilities affected by the cracking have asked GE to instrument the plant subassemblies so that mechanical responses may be measured during subsequent operation; these tests have included pressure, strain and acceleration measurements. Thus there is a reasonable measuremen t

data base on the actual dryer to supplement the "paper"analyses that have been conducted, and these have led to recommendations to the BWR owners from GE regarding various operational procedures; e.g. attention to reassembly procedures. It was further pointed out that many of the steam dryer designs were modified based on these in-reactor tests; however, such changes were no t

made to the steam dryer at Quad Cities.

4. There were no changes recommended to the dryer design specifically becaus e

of power uprate (PU) conditions. This decision was based not on vibration tests on model assemblies (since GE does not have such a facility), but on the basis of analyses and data from operating plant mentioned above. Reasons for not changing the dryer design with PU included the following;

• No dome pressure change with PU.

• Even though the stress on the component will increase with flow ra te (squared) it was decided that there was not a significant change in IGSCC poter-tial

• It was recognized that the potential for fatigue cracking would be increased basec, for instance, on strain measurements conducted on the Susquehanna support lugs, but it was thought that this was manageable given the inspection frequency and the fact that the dryer was not a safety-relate d

component.

5. Obviously some of the conclusions above relating to changes in steam dryer design associated with PU can be challenged in the light of Quad Cities, but the actions being taken at GE (admittedly reactively) include th e

evaluation of the cracking probability for all steam dryers in the BWR fleet and the consequence of dryer failure, should it occur. The former will rely primarily on vibrational analyses, calibrated by the field test measurements (in item 3), and the latter will focus primarily on a loose parts analysis.

The vibrational analyses are being backed up by a limited mechanical testin 9

/~

Carol Rowe - Mime.822 Page 3

'ST*****

-...... (".0w. '.".

program on a 1/16 scale model dryer Utilities will then be advised as to whether modifications to their steam dryers should be made based on the resultant frequency/consequence assessment.

6. When challenged as to the likelihood of FIV incidences similar to Quad Cities occurring again, GE admitted that it could not be dismissed, but deemed that the risk was minimized by the actions being taken in item 5 and the fact that the consequence was relatively low.

l.---.

---

!l---

UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C.120555*0001 October 2, 2002 MEMORANDUM TO:

George Apostolakis. Thomas Kress ACRS Members FROM:

J/.J Hossein Nourbakhsh, Senior Fellow

SUBJECT:

EXTRACTING PARAMETER AND MODEL UNCERTAINTIES FROM THE NUREG 1150 OVERALL UNCERTAINTY Attached, for your comments, are my initial thoughts on feasibility of extracting parameter and model uncertainties from the NUREG*1150 overall uncertainty.

CC: John Larkins Sher Bahadur

I Feasibility of Using the NUREG-1150 Uncertainty Analyses in RE'gulatory Decisions Involving PRA Results by Hossein P. Nourbakhsh ACRS Senior Fellow U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

1. Introduction Probabilistic Risk Assessment (PRA) is being used increasingly as an important element in regulatory decision making. A concern associated with the results ofPRAs stems from their susceptibility to model uncertainties. These uncertainties are often of such a magnitude that they make the decision making process difficult.

There are two classes of epistemic uncertainty that impact the results ofPRAs: parametric uncertainty and model uncertainty. Parameter uncertainties are those associated with the values of the fundamental parameters of the PRA model, such as initiating event frequencies and equipment failure rates that are used in quantifying the accident sequence frequencies.

Model uncertainties are those associated with the use of models for specific events or phenomena used in the development of the PRA model. Examples include approaches to modeling reactor coolant pump seal behavior and containment pressurization at reactor vessel breach.

The l\\TUREG-ll SO study [Ref. IJwas a major effort to put into a risk perspective the insights into system behavior and ph~nomenological aspects ofsevere accidents. An important characteristic ofthis study was the inclusion of the uncertainties in the calculations of core damage frequency and risk that exist because ofincompJete understanding of reactor systems and severe accident phenomena.

Five specific commercial nuclear power plants were analyzed in NUREG-1150: Surry Power Stations, a 3-loop Westinghouse PWR with a subatmospheric containment; Zion, a 4.Loop Westinghouse PWR with large dry containment; Sequoyah,a 4-loop Westinghouse PWR with ice-condenser containment; Peach Bonom, a BWR*4 reactor with a Mark I containment; and Grand Gulf, a BWR-6 reactor with a Mark III containment.

Assessing modeling uncertainties associated with the results of a plant-specific PRA is very resource intensive. The purpose of this paper is to review NUREG-1150 to see if parameter and model uncertainties can be extracted from the overall uncertainty in order to have an estimate ofjust the model uncertainty contribution. This estimate could then be used in regulatory decisions involving PRA results that include only parameter uncertainty.

The NUREG-ll SO constituent analyses are briefly summarized first in order to provide a framework for discussion presented later on feasibility of utilizing the NUREG-1150 information on model uncertainty associated with each of the key elements of risk analysis.

1

</

2. Oven'iew of the NUREG-1150 Study figure 2.1 displays schematically the components ofthe NUREG-l ISO analytical process which consists ofthe following key elements:

(1)

Systems analysis and models of plant response to various initiating events, quantification of accident sequences leading to core damage; (2)

Analysis of the accident progression and containment performance to determine various possible ways the accident could evolve given core damage; (3)

Source term analysis, the release of radioactive material to the environment for various outcomes of the accident progression; (4)

Consequence analysis, the health impacts of each of the source terms.

~1 Core Damage Accident

S

.~

Frequency Analysis Plant Damage States

....-r.--..

~

I

_._*...*,_M'._

z¥.,!j i

~

Accident Progression and Containment Performance j

.ii Analysis tAccident Progression Bins

"............~

-. i'i i

.~

Source Term Analysis 1

11

~ Source Term Group..s ii:

"I Offsite Consequences Analysis J

f YConsequence Measu res

__i.'WW.

'1S

~

f"-"

R_i_Sk_I_D_t_eg_r_a_t_io_n li Figure2.1 Elements of The NUREG-I 150 Risk analysis Process 2

Integrated risk was obtained by combining the frequency of core damage, the conditional probability of the release paths, and the value of the consequences of each source tenns conditional on the release into a single risk measure. By repeating the calculation several times with different input values (over specific ranges) of key parameters, a distribution of offsite risk estimates was obtained from which the uncertainties in the risk were estimated.

2.1 Core-Dama2e Accident Frequencv Analysis The core damage frequency analysis considered accidents initiated by events occurring during nonnal full power operation ofthe plants (internal events). The analysis ofaccident frequencies for the Surry and Peach Bottom also included the consideration of accidents initiated by external events (e.g. earthquake, floods, fires). The analysis consists offault trees and event trees delineating the accident sequences leading to core damage.

The calculations of core damage frequency and risk included the quantitative analysis of uncertainties.

This analysis was perfonned using the Latin hypercube sampling technique. Probability distributions for many parameters for which the uncertainties were estimated to be large and important to risk were developed. For example 48 variables were sampled in accident frequency analysis for the Surry plant.

Probability distributions for many of the most important accident frequency variables were generated using statistical analyses of plant data or data from other published sources (Ref. I) For certain key issues in the uncertainty analysis, the elicitation of expert judgment was used to develop the needed probability distributions. An example ofthe accident analysis issues evaluated by the expert panels was the frequency and size of reactor coolant pump (Rep) seal failures before the onset of core damage in PWRs.

The outcome of the frequency analysis was a group of accident sequences leading to core damage and their associated f:-equencies. The accidents were then grouped into plant damage states (PDSs), based on similarity of plant conditions, to define the ent!)' points for the subsequent accident progression analysis.

2.2 Accident Pro~ression and Containment Performance Analvsis For each general type of accident, defined by the plant damage states, an analysis was performed to develop Phenomenological conditions and containment response for each accident progression path which determine the timing and failure mode of containment and influence the transport and release of radionuclides.

Accident progression was analyzed using a single accident progression event tree (APET) developed for each plant which was evaluated with the EVNTRE code [Ref. 2]. The specification of each PDS defines the entry conditions to the APET. The accident progression event trees developed for this study made extensive use of the available severe accident computer code calculations and experimental observations.

The elicitation of expert judgment was used to develop probability distributions for fourteen accident progression, containment loading, and structural response issues. Probabilit)1 distributions for many other issues believed to be of less importance to risk were also developed by analysts on the project staff or by phenomenologists from national laboratories using techniques like those employed with the expert panels.

3

I The APET developed for Surry and Zion in NUREG-1150 had over 70 event questions and many of the questions had several (more than two) outcomes; there are thus far too many paths through the tree to allow consideration of each individual path in terms ofthe subsequent source term and consequence analysis. The outcomes of the paths were then grouped into accident progression bins (APBs) which have similar characteristics and define the entry conditions for the source term analysis.

2.3 Source Term Analysis The magnitude and composition of radioactive materials released to the environment with associated energy content, time, initial elevation and duration of release together are termed the "source term".

The source term analysis tracks the release and transport of radioactive materials from the core, through the reactor coolant system, then to the containment and other buildings, and finally into the environment.

The removal and retention of radioactive materials by natural processes, such as deposition on surfaces, and by engineered safety systems, such as sprays, are accounted for in each location.

For the NUREG-1150 risk analysis, the source terms were calculated using simplified parametric algorithm. The parametric equations describe the source terms as the product of release fractions and transmission factors at successive stages in the accident progression for a variety ofrelease pathways, a variety of accident progressions, and nine classes of radionuclides. This approach led to development of separate computer codes for each plant, i.e., the XSOR codes [Ref. 3]. The parametric models used in the XSOR codes are not time dependent. These codes generate source terms only in terms of early and delayed releases. The timing of release is particularly important for the prediction of early health effects.

None of the basic parameters used in the XSOR codes are internally calculated. The values for the parameters must be specified by the user or chosen from a distribution of values by a sampling algorithm.

The input data on the more important parameters were constructed in the forms of probability distributions. Such distributions were developed using the elicitation of expert judgement. For a few parameters that were judged of lesser importance or not considered as uncertain, single-valued estimates were used in XSOR models.

The source term analysis resulted in characterizing thousands of source terms (20,000 for Surry and 75,000 for Grand Gulf) associated with tens of plant damage states, hundreds of accident progression bins, and the variation in source term phenomenological issues which were included in the propagation of uncertainties.

For the risk analysis, radioactive releases were grouped using the PARTITION program [Ref. 4]

according to their potential to cause early or latent cancer fatalities and warning time. Through this "partitioning" process, the large number of radioactive releases calculated with the XSOR codes were collected into a small set of source term groups (30 to 60 in number for each plant). This set of groups was then used in the offsite consequence calculations.

4

2.3 Offsite Conseg uence Analvsis The severe accident radiological releases are of concern because of their potential for impacts on the surrounding environment and population. The impact of such releases to the atmosphere can manifest themselves in a variety of early and delayed health effects, loss of habitability of areas close to the plant site, and economic losses [Ref. 1]. In NUREG-1150 study, the consequence measures, early fatalities, population dose (person-rem), and latent cancer fatalities, were calculated for each source term group by the MAACS code [Ref 5]. The output ofMACCS for each source term group is a distribution of the consequences, conditional on occurrence ofthe source term, which incorporates the uncertainty (variabilit),) due to weather as well as the uncertainty in the underlying health (dose-response) models.

3. Feasibility of Utilizing the NUREG*1150 Information on l\\fodel Uncertainty As stated in the introduction, an important characteristic ofNUREG -1150 study was the inclusion of quantitative estimates of the uncertainties in the calculations of core damage frequency and risk. Both t)'pes of epistemic uncen.ainties (parametric uncertainty and model uncertainty) were included in the NUREG-1150 study, and no effon. was made to differentiate between the effects of the two types of uncertainties.

Impon.ant source of uncertainties exist in all four stages of risk analysis sho\\\\'Jl in Figure 2.1. In order for uncertainties in accident phenomena to be included in the probabilistic risk analyses conducted for the NUREG*1150 study, they had to be expressed in terms of uncertainties in the "high level" or summary parameters that were used in the study.

The NUREG-1 ]50 analytical procedure for risk analysis is a cumbersome process which involves numerous computer codes and data transfer. Therefore, an effon. to extract parameter and model uncertainties from overall uncertainty may become very involved and resource intensive. In addition, any such evaluation of model uncertainty should reflect the more recent technical knowledge and understanding of severe accident phenomena. For example, since the completion ofNUREG* I150 study, advances have been made in the ability to predict the early containment failure phenomena of direct containment heating (DCH) and liner meltthrough that should be reflected in the model uncertainty. It should be noted that, as a pan of a study to assess the risk significance of containment and related engineered safet)' features (ESF) system performance requirements [Ref. 6], the accident progression event trees (APETs) for lion and Peach Bottom that had been used forNUREG-1150 were modified to reflect the current knowledge of early containment failure phenomena (DCH and Liner meltthrough).

The direct applicability of overall model uncertainty calculated for NUREG-1150 plants to other plants of similar NSSS (nuclear steam supply system) and containment design may be questionable. There are plant specific features and operational practices that may influence the likelihood and the severity of specific events or phenomena during the progression of severe accidents. For example, the results of individual plant examinations (lPEs)[Ref. 7] indicate that, specific containment features lead to unique and significant failure modes. For instance, the large probability values of early containment failures found in the IPEs for both Palisades and Davis Besse, do not result from the high pressure loads associated with DCR Instead, the values are attributed to the special features of the pan.icular designs of 5

the plants. The location of the (ESF) sump in Palisades cause's the flow of molten core debris from the reactor cavity into this sump and subsequently into the ESF recirculation piping. In the IPE analysis, the debris is assumed to eventually melt through the pipe wall and enter the auxiliary building, resulting in a large containment failure area. For Davis Besse, one of the few PWR plants that have large dry containments of steel construction, the largest fraction of early containment failure is associated with the potential failure of the containment wall via direct contact with core debris.

In spite of plant specific nature ofNUREG -1150 quantitative results (e.g., core damage frequency, containment performance, risk), this study provides valuable insights on severe accident phenomenological issues and associated state-of-knowledge uncertainties which are very useful to the study of plants of sim ilar NSSS and containment design.

While a formal propagation of the uncertainty is the best way to account for model uncertainties, under certain circumstances, it can be demonstrated that the model uncertainties associated with many phenomenological issues are not important to the overall risk. For example, it can be demonstrated that the bulk of risk significant contributing scenarios do not involve highly uncertain phenomenological issues. For instance, in a PWR plant with a high frequency of containment bypass sequences (i.e., Event V and SGTR) and with a high probability of depressurization of reactor coolant system (RCS), the model uncertainties associated with the thermally induced failure of steam generator tubes and direct containment heating are not important to the overall uncertainty in the early fatality risk.

NUREG-] ]50 and the results ofthe IPE Insights Program [Ref. 7] provide important sources of information that may be used to develop a simplified systematic methodology for assessing the plant specific importance of individual phenomenological issues and their model uncertainties to the overall uncertainty. The feasibility and options for developing such assessment methodology should be further explored. Development of such assessment methodology can greatly reduce the effort necessary for a formal propagation of the uncertainty to account for model uncertainty.

It may also be desirable to assign some ranking of"risk importance" among the various phenomenological issues that are considered in a plant PSA model. For example, a risk importance measure of "Risk Significance Worth", somewhat similar to Risk Achievement Worth (RAW) used for risk importance ranking of various plant components, can be defined as:

Risk Significance Worth =R (issue IflRo Where R (issue 1r is the calculated risk, using the bounding (conservative) assumptions in quantifying the phenomenological issue -], and Ro is the base-case risk. Examples of phenomenological issues include containment pressurization due to DCH, containment failure pressure, and probability of temperature* induced steam generator tube rupture. Other risk importance measures somewhat similar to Fussel-Vessly (F-V) or Risk Reduction Worth (RRW) can also be defined. It should be noted that various risk metrics (e.g. LERF) could be used for defining these risk importance measures. Development of importance measures for phenomenological issues should also be further explored. Such risk importance measures for phenomenological issues can be useful for directing any needed sensitivity analysis (in the absence of any formal model uncertainty analysis), as well as for developing research priorities to reduce the overall model uncertainty.

6

(I

4. Summary and Conclusions Assessing modeling uncertainties associated with the results of a plant-specific PRA is very resource intensive. NUREG-I ISO constituent analyses was reviewed to see if parameter and model uncertainties could be extracted from the overall uncertainty in order to have an estimate ofjust the model uncertainty contribution. This estimate could then be used in regulatory decisions involving PRA results that include only parameter uncertainty.

The NUREG-llS0 analytical procedure for risk analysis was found to be a cumbersome* process which involves numerous computer codes and data transfer. Therefore, an effort to extract parameter and model uncertainties from overall uncertainty may become very involved and resource intensive. In addition, any such evaluation of model uncertainty should reflect the more recent technical knowledge and understanding of severe accident phenomena.

The direct applicability of overall model uncertainty calculated for NUREG-l1S0 plants to other plants of similar NSSS (nuclear steam supply system) and containment design may be questionable. There are plant specific features and operational practices that may influence the likelihood and the severity of specific events or phenomena during the progression of severe accidents. In spite of plant specific nature of 1I.TUREG *1 ISO quantitative results, this study provides valuable insights on severe accident phenomenological issues and associated state* of-knowledge uncertainties which are very useful to the study of plants of similar NSSS and containment design.

While a formal propagation ofthe uncertainty is the best way to account for model uncertainties, under certain circumstances, it can be demonstrated that the model uncertainties associated with many phenomenological issues are not important to the overall risk. NUREG*l ISO and the results ofthe IPE Insights Program [Ref. 7] provide important sources of information that may be used to develop a simplified systematic methodology for assessing the plant specific importance of individual phenomenological issues and their model uncertainties to the overall uncertainty. The feasibility and options for developing such assessment methodology should be further explored. Development of such assessment methodology can greatly reduce the effort necessary for a formal propagation ofthe uncertaimy to account for model uncertainty.

It may also be desirable to assign some ranking of "risk importance" among the various phenomenological issues that are considered in a plant PSA model. This paper has presented examples for such risk importance measures. Development of importance measures for phenomenological issues should also be further explored. Such risk importance measures for phenomenological issues can be usefu1for directing any needed sensitivity analysis (in the absence ofany formal model uncertainty analysis), as well as for developing research priorities to reduce the overall model uncertainty.

7

4. References

1.

u.s. Nuclear Regulatory Commission, " Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants," NUREG*1150, December 1990.

2.

J. M. Griesmeyer and L. N. Smith, "A Reference Manual for the Event Progression Analysis Code (EVNTRE)," Sandia National Laboratories, NUREG/CR*5174, SAND88*1607, September 1989.

3.

H. N. Jow, W. B. Murfin and J. D. Johnson, "XSOR Codes User's Manual," Sandia National Laboratories, NUREG/CR-5360, December 1993.

4.

R. L. Iman et aI., *'PARTITION: A Program for Defining the Source Term/Consequence Analysis Interface in the NUREG*11 50 Probabilistic Risk Assessments," Sandia National Laboratories, NUREG/CR*5253, SAND88-2940, May 1990.

5.

D. I. Chanin et aI., "MELCOR Accident Consequence Code System (MACCS)," Sandia National Laboratories, NUREG/CR-4691, Vo1s. 1-3, SAND86-1562, February 1990.

6.

H. P. Nourbakhsh, A. L. Hanson, and W. T. Pratt, "Risk Importance of Containment and Related ESF System Performance Requirements," Brookhaven National Laboratory, NUREG/CR-6418, BNL*NUREG*52489, November 1998.

7.

U.S. Nuclear Regulatory Commission, *'Individua1 Plant Examination Program: Perspectives on Reactor Safet)' and Plant Performance", NUREG-1 560, November 1996.

8

ICarol Rowe - P&P MEETING *....

Page*

From:

George Apostolakis <apostola@mit.edu>

To:

<jtl @nrc.gov>

Date:

10/6/02 10:54AM

Subject:

P&P MEETING John:

Item 9: I strongly support the recommendation that Hossein work on the F-C curves. This is a forward-looking subject that could be of great importance to the licensing of future plants.

Hossein's conclusions regarding NUREG-1150 do not surprise me. It is indeed very difficult to produce an overall estimate of model uncertainty. I don't think it's worth pursuing this issue further at this time. Let's wait until Fleming gives us his input.

Item 10:

I disagree that we need a facilitator. The ACRS chairman has moderated all the retreats (except the one in Boston). I have not heard any complaints that efficiency has suffered. Even in Boston, a man of Neil Todreas's stature was uncomfortable moderating our sessions. So-called "professional" moderators are very annoying because

.they don't understand the issues under discussion and all they care about is keeping time.

The Subcommittee cannot just recommend that "adequate time" be provided. The Subcommittee should actually specify the time.

still think that four hours would be sufficient, although some flexibility would be required. I agree that starting the retreat with this item is a good idea. I do believe that Mario's wishes on this matter should be respected, even though the election has not taken place yet.

George cc:

<sxd1 @nrc.gov>

Page Item 9: I strongly support the recommendation that Hossein work on the F-C curves. This is a forward-looking subject that could be of great importance to the licensing of future plants.

Hossein's conclusions regarding NUREG-1150 do not surprise me. It is indeed very difficult to produce an overall estimate of model uncertainty. I don't think it's worth pursuing this issue further at this time. Let's wait until Fleming gives us his input.

Item 10:

I disagree that we need a facilitator. The ACRS chairman has moderated all the retreats (except the one in Boston). I have not heard any complaints that efficiency has suffered.

Even in Boston, a man of Neil Todreas's stature was uncomfortable moderating our sessions. 50 called "professional" moderators are very annoying because they don't understand the issues under discussion and all they care about is keeping time.

The Subcommittee cannot just recommend that "adequate time" be provided. The Subcommittee should actually specify the time. I still think that four hours would be sufficient, although some flexibility would be required. I agree that starting the retreat with this item is a good idea. I do believe that Mario's wishes on this matter should be respected, even though the election has not taken place yet.

George 3U

i Sam Duraiswamy - Members' Issues for the ACRS Retreat Page 1 From:

"Powers, Dana A" <dapower@sandia.gov>

To:

"APOSTOLAKIS, George" <apostola@mit.edu>, "BONACA,Mario"

<mvbonaca@snet.net>, "FORD, F. Pete~' <FPCTFord@aol.com>, "KRESS, T.S." <TSKress@aol.com>,

"LEITCH, Graham" <gmleitch@aol.com>, "ROSEN, Steve" <historyart@computron.net>, "SHACK, Bill"

<wjshack@anl.gov>, "sieber, JACK" <jdsieber@aol.com>, "WALLIS, Graham B."

<Graham.B.Wallis@Dartmouth.edu>

Date:

9/17/0210:55AM

Subject:

Members' Issues for the ACRS Retreat The current plan for the ACRS retreat relegates issues of ACRS process and procedures to a four hour period - undoubtably at the end of the ordeal. The rest of the time is to be spent on technical issues that really seem to be more appropriate for regular committee meetings. Perhaps it is the case that there are no real members' issues to discuss. But, it does seem worthwhile to discuss this and see if there are issues that might be worthy of more than the cursory treatment allowed in the current agenda for the retreat.

Perhaps we should compile a list of issues of process and procedure that could be the SUbject of discussion at the retreat to see if the list will fit comfortably within the allotted 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Some issues that I can well imagine include the following:

- ACRS effectiveness and self-assessment There seems to be a general degradation in the efficiency of the ACRS. We continue to have meetings of prolonged duration that we had when we were generating 6 to 7 letters per meeting. Now we generate about 2-3. When we were producing a larger number of letters we even had an hour for members to work on letters that has now disappeared to support the lower productivity.

This seems strange. We hail self-assessment by licensees as a valuable method for improving quality, but we are avoiding application of this technique to ourselves as a full committee or to our subcommittees.

- Proliferation of subcommittee meetings We seem to have entered an era involving a lot of subcommittee meetings some of which are only marginally longer than a presentation to the full committee. This comes at a time when the engineering support for subcommittee meetings by experienced ACRS staff engineers is limited because of the transfer of Markely and DUdley. 80ehnart keeps threatening to retire which would further detract from the ability of the staff to support multiple subcommittee meetings. Can we continue to operate as we have been?

- Compensation process A new compensation process has been inflicted on the members that is completely opague. We have no idea what numbers to bill for what work and cannot ascertain what charges have been made.

- Limited evidence of ACRS strategy There is limited evidence of any strategy by the ACRS. We are back on what Prof. Wallis has so poignantly called the reactive treadmill. We detected weakness of a severe nature in the way NRC staff is doing power uprate reviews, but we did not follow up and, in fact, simply let the staff get away with inadequate reviews for fear of penalizing licensees. We see the staff ignoring technical evidence of high burnup fuel vulnerability, but we f

J;tT.

/0

!JI

'Sam Du, aiswamy - Members' Issues for the ACRS Retreat Pa~e 2 let them go ahead with this. We are not holding the staff to high technical standards. There is little evidence that the ACRS is searching for issues that others within the agency are overlooking.

- Confusion in Issue Assignments and Subcommittee responsibilities Issues that based on title seem to belong to one subcommittee are showing up in the domain of other subcommittees. Members who cannot attend subcommittee meetings are being found assigned responsibility for draft letters.

- Differing opinions within the ACRS are not getting aired There are some areas within risk-informed regulatory plans that there appear to be splits within the ACRS, but these issues get debated electronically rather than within the context of meetings.

I am sure others issues can be identified and it would be of interest to see what these issues are.

Dana cc:

"'sxd1@nrc.gov'" <sxd1@nrc.gov>

!'Sam Dl,;raiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page' I

From:

"Mario V. Bonaca" <mvbonaca@snet.net>

To:

<RANSOM@ECN.PURDUE#062#EDU>. "Powers, Dana A" <dapower@sandia.gov>,

"FORD, F. Peter" <FPCTFord@aol.com>, "KRESS, T.S." <TSKress@aol.com>, "LEITCH, Graham"

<gmleitch@aol.com>, "ROSEN, Steve" <historyart@computron.net>, "SHACK, Bill" <wjshack@anl.gov>,

"sieber, JACK" <jdsieber@aol.com>, "WALLIS, Graham B." <Graham.B.Wallis@Dartmouth.edu>, "George Apostolakis" <apostola@MIT.EDU>

Date:

9/20102 12:47PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge

George, Since we haven't yet elected a new Chaiman, I assume that you are asking me to take care of these proposals because I am currently Vice-Chairman, and as such, I will be glad to do it.

I believe that the issues Dana raises are significant and central to the functioning of the committee. I think that this planning session should have priority on all other business, if any, so it should be held at the beginning of the retreat and last for as long as it takes.

On the other hand, we don't want it to become just a complaining and venting session. So we need to plan an agenda that should start with a strategy

.session. I propose that at the end of the next P&P, you, John, Tom and I get together in executive session and work on this agenda, to identify topics and assign adequate time to each topic. We will then bring this agenda to the committee and finalize it. Once decided how much time we need for this session, we will fit into the retreat other topics as time allows.

Members are requested to provide us with some of the issues, in addition to those raised by Dana, that should be discussed during this session.

Regarding the location for the retreat, J am open to staying in Rockville if necessary. On the other hand, I don't see why meeting in Las Vegas and visiting Yucca Mountain would be a boondagle. This is not an unnecessary trip; we will work in retreat in Las Vegas as we will in Rockville. Steve is correct when he says that Yucca Mountain is important for this committee, and we will have to visit the site at some point. On the other hand, if it is a controversial issue as it appears to be, we should bring it to the

. membership and vote on it in October. After all, John may tell us that we have no money for this trip and that will close the issue.

Mario

-- Original Message From: "George Apostolakis" <apostola@MIT.EDU>

To: "Powers, Dana A" <dapower@sandia.gov>; "BONACA,Mario"

<mvbonaca@snet.net>; "FORD, F. Peter" <FPCTFord@aol.com>; "KRESS, T.S."*

<TSKress@aol.com>; "LEITCH, Graham" <gmleitch@aol.com>; "ROSEN, Steve"

<historyart@computron.net>; "SHACK, Bill" <wjshack@anl.gov>; "sieber, JACK"

<jdsieber@aol.com>; "WALLIS, Graham B." <Graham.B.Wallis@Dartmouth.edu>

Cc: <sxd 1@nrc.gov>

Sent: Tuesday, September 17, 2002 2:16 PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge

> Mario:

~<..I()

J3

Sam Duraiswamy - Re: Members' Issues for the ACRS Retreat Mario's Charge Page 2

>1 defer to you on what to do with these proposals.

> George

> At 08:56 AM 9/17/2002 -0600. Powers. Dana A wrote:

> >The current plan for the ACRS retreat relegates issues of ACRS process and

> >procedures to a four hour period - undoubtably at the end of the ordeal.

The

> >rest of the time is to be spent on technical issues that really seem to be

> >more appropriate for regular committee meetings. Perhaps it is the case that

> >there are no real members' issues to discuss. But, it does seem worthwhile

> Dr. G.E. Apostolakis

> Professor of Nuclear Engineering

> Professor of Engineering Systems

> Room 24*221

> Massachusetts Institute of Technology

> Cambridge, MA 02139-4307, USA

> e*mail: apostola@mit.edu

> tel: +1-617-252-1570

> fax: +1-617-258-8863 cc:

<sxd1@nrc.gov>, "John Larkins" <JTL@nrc.gov>

Sam DUiaiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 1 From:

George Apostolakis <apostola@MIT.EDU>

To:

"Mario V. Sonaca" <mvbonaca@snet.net>, <RANSOM@ECN.PURDUE#062#EDU>.

"Powers, Dana A" <dapower@sandia.gov>, "FORD, F. Peter" <FPCTFord@aol.com>, "KRESS, T.S."

<TSKress@aol.com>, "LEITCH, Graham" <gmleitch@aol.com>, "ROSEN, Steve"

<historyart@computron.net>, "SHACK, Bill" <wjshack@anl.gov>, "sieber, JACK" <jdsieber@aol.com>.

"WALLIS, Graham B." <Graham.B.Wallis@Dartmouth.edu>

Date:

9/201022:16PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge Mario:

I don't think that we need more than four hours to discuss the issues that Dana has raised. Most of them we have discussed in the past. We have started to publish reports and papers on issues of particular concern to the ACRS (safety culture, did. etc) and, with Dr. Nourbakch joining us, we will continue to do so. Gus and Tom are going to PSA '02 to present papers on our behalf. There are always new issues that require attention (Dana mentioned a few) and we will certainly debate them at the retreat. The differing opinions on risk-informed regulatory plans will be aired when we have our session with Karl Fleming.

We always get excited when we hear that a member complains that the ACRS is not functioning well. Then, reality sets in. We are an advisory committee that must respond to the Commission's and staffs requests for reviews and comments. This does take most of our time. We have looked for ways to reduce the burden, but we have been unable to do anything drastic. The members can always propose new ideas without waiting for the retreat. Furthermore, we are part-time advisors. Several members keep complaining that they are spending too much time on committee business.

Frankly, I don't know what new ideas will be proposed to justify allotting more than four hours to this item. I am afraid that the result will be "just a complaining and venting session." Of course, I am willing to be convinced otherwise and I agree with you that the P&P subcommittee should talk about it in October.

Regarding the location of the retreat: Apparently, some members feel that the ACRS members should not enjoy themselves after hours (surely they don't mean that we will neglect our business and go to the gambling tables). Yucca mountain is not within our charter, yet it is of such major importance to the nuclear enterprise that I don't see how it would harm the committee getting a first-hand look. Besides, you never know what we will be asked to do in the future (see the joint ACNW/ACRS SUbcommittee).

The Government gets its money's worth from us. The rate of pay is laughable and we definitely put in more than eight hours per day when we are in Rockville. I, for one, am offended when someone hints that I am trying to exploit the government's resources. Finally, I believe that some people underestimate the value of the camaraderie that is created when we all (including spouses) get together in a relaxed environment. Perhaps doing so would contribute to the reduction of the acrimony that we sometimes witness in meetings.

We will make final decisions in October.

George

. Sam Duraiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 2 At 01 :04 PM 9/20/2002 -0500, Mario V. Bonaca wrote:

>George,

>Since we haven't yet elected a new Chaiman, I assume that you are asking me

>to take care of these proposals because I am currently Vice-Chairman, and as

>such, I will be glad to do it.

>1 believe that the issues Dana raises are significant and central to the

>functioning of the committee. I think that this planning session should have

>priority on all other business, if any, so it should be held at the

>beginning of the retreat and last for as long as it takes.

>On the other hand, we don't want it to become just a complaining and venting

>session. So we need to plan an agenda that should start with a strategy

>session. I propose that at the end of the next P&P, you, John, Tom and I get

>together in executive session and work on this agenda, to identify topics

>and assign adequate time to each topic. We will then bring this agenda to

>the committee and finalize it. Once decided how much time we need for this

>session, we will fit into the retreat other topics as time allows.

>Members are requested to provide us with some of the issues. in addition to

>those raised by Dana, that should be discussed during this session.

>Regarding the location for the retreat. I am open to staying in Rockville if

>necessnry. On the other hand, I don't see why meeting in Las Vegas and

>visiting Yucca Mountain would be a boondagle. This is not an unnecessary

>trip; we will work in retreat in Las Vegas as we will in Rockville. Steve is

>correct when he says that Yucca Mountain is important for this committee,

>and we will have to visit the site at some point. On the other hand, if it

>is a controversial issue as it appears to be, we should bring it to the

>membership and vote on it in October. After all, John may tell us that we

>have no money for this trip and that will close the issue.

>Mario

>---- Original Message

. >From: "George Apostolakis" <apostola@MIT.EDU>

>To: "Powers. Dana A" <dapower@sandia.gov>; "BONACA,Mario"

><mvbonaca@snet.net>; "FORD, F. Peter" <FPCTFord@aol.com>: "KRESS, T.S."

><TSKress@aol.com>; "LEITCH, Graham" <gmleitch@aol.com>; "ROSEN, Steve"

><historyart@computron.net>; "SHACK, Bill" <wjshack@anl.gov>; "sieber, JACK"

><jdsieber@aol.com>; "WALLIS, Graham B." <Graham.B.Wallis@Dartmouth.edu>

>Cc: <sxd1@nrc.gov>

>Sent: Tuesday, September 17, 20022:16 PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge

> > Mario:

> > I defer to you on what to do with these proposals.

> > George

> > At 08:56 AM 9/17/2002 -0600, Powers, Dana A wrote:

ISam Duraiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 3

> > >The current plan for the ACRS retreat relegates issues of ACRS process

>and

> > >procedures to a four hour period - undoubtably at the end of the ordeal.

>The

> > >rest of the time is to be spent on technical issues that really seem to

>be

> > >more appropriate for regular committee meetings. Perhaps it is the case

>that

> > >there are no real members' issues to discuss. But, it does seem

>worthwhile

> > Dr. G.E. Apostolakis

> > Professor of Nuclear Engineering

> > Professor of Engineering Systems

> > Room 24-221

> > Massachusetts Institute of Technology

> > Cambridge, MA 02139-4307, USA

> > e-mail: apostola@mit.edu

> > tel: +1-617-252-1570

> > fax: +1-617-258-8863 Dr. G. E. Apostolakis Professor of Nuclear Engineering Professor of Engineering Systems Room 24-221 Massachusetts Institute of Technology Cambridge, MA 02139-4307, USA e-mail: apostola@mit.edu tel: +1-617-252-~570 fax: +1-617-258-8863 cc:

<sxd 1@nrc.gov>, "John Larkins" <JTL@nrc.gov>

Sam DUiaiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 1 From:

"Mario V. Bonaca" <mvbonaca@snet.net>

To:

<RANSOM@ECN.PURDUE#062#EDU>, "Powers, Dana A" <dapower@sandia.gov>,

"FORD, F. Peter" <FPCTFord@aol.com>, "KRESS, T.S." <TSKress@aol.com>, "LEITCH, Graham"

<gmleitch@aol.com>. "ROSEN, Steve" <historyart@computron.net>, "SHACK, Bill" <wjshack@anl.gov>,

"sieber, JACK" <jdsieber@aol.com>, "WALLIS, Graham B." <Graham.B.Wallis@Dartmouth.edu>, "George Apostolakis" <apostola@MIT.EDU>

Date:

9/21/023:38PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge

George, I did not mean to prejudge how much time it will take. Four hrs may be enough, but we had to look at the list of issues being raised, estimate what it will take to cover them effectively and give this planning session the time needed. We should also conduct this session early during the retreat, so that it doesn't get pushed to the last minute, when members are already half way out of the door.

Mario

--- Original Message From: "George Apostolakis" <apostola@MIT.EDU>

To: "Mario V. Bonaca" <mvbonaca@snet.net>; <RANSOM@ECN.PURDUE#062#EDU>;

"Powers, Dana A" <dapower@sandia.gov>; "FORD, F. Peter" <FPCTFord@aol.com>;

"KRESS, T.S." <TSKress@aol.com>; "LEITCH, Graham" <gmleitch@aol.com>:

"ROSEN. Steve" <historyart@computron.net>; "SHACK, Bill" <wjshack@anl.gov>;

"sieber, JACK" <jdsieber@aol.com>: "WALLIS, Graham 8."

<Graham. B. Wallis@Dartmouth.edu>

Cc: <sxd1@nrc.gov>; "John Larkins" <JTL@nrc.gov>

Sent: Friday, September 20,2002 1:06 PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge

> Mario:

> I don't think that we need more than four hours to discuss the issues that

> Dana has raised. Most of them we have discussed in the past. We have

> started to publish reports and papers on issues of particular concern to

> the ACRS (safety culture, did, etc) and, with Dr. Nourbakch joining us, we

> will continue to do so. Gus and Tom are going to PSA '02 to present papers

> on our behalf. There are always new issues that require attention (Dana

> mentioned a few) and we will certainly debate them at the retreat. The

> differing opinions on risk-informed regulatory plans will be aired when we

> have our session with Karl Fleming.

> We always get excited when we hear that a member complains that the ACRS is

> not functioning well. Then, reality sets in. We are an advisory committee

> that must respond to the Commission's and staffs requests for reviews and

> comments. This does take most of our time. We have looked for ways to J~~ 10 38

'Sam Duraiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 2

> reduce the burden, but we have been unable to do anything drastic. The

> members can always propose new ideas without waiting for the

> retreat. Furthermore, we are part-time advisors. Several members keep

> complaining that they are spending too much time on committee business.

> Frankly, I don't know what new ideas will be proposed to justify allotting

> more than four hours to this item. I am afraid that the result will be

> "just a complaining and venting

> session." Of course, I am Willing to be convinced otherwise and I agree

> with you that the P&P subcommittee should talk about it in October.

> Regarding the location of the retreat: Apparently, some members feel that

> the ACRS members should not enjoy themselves after hours (surely they don't

> mean that we will neglect our business and go to the gambling

> tables). Yucca mountain is not within our charter, yet it is of such major

> importance to the nuclear enterprise that I don't see how it would harm the

> committee getting a first-hand look. Besides, you never know what we will

> be asked to do in the future (see the joint ACNW/ACRS subcommittee).

> The Government gets its money's worth from us, The rate of pay is

> laughable and we definitely put in more than eight hours per day when we

> are in Rockville. I, for one, am offended when someone hints that I am

> trying to exploit the government's resources, Finally, I believe that some

> people underestimate the value of the camaraderie that is created when we

> all (including spouses) get together in a relaxed environment. Perhaps

> doing so would contribute to the reduction of the acrimony that we

> sometimes witness in meetings.

> We will make final decisions in October.

> George

> At 01 :04 PM 9/20/2002 -0500, Mario V. Sonaca wrote:

. > >George.

> >Since we haven't yet elected a new Chaiman, I assume that you are asking me

> >to take care of these proposals because I am currently Vice-Chairman, and as

> >such, I will be glad to do it.

> >\\ believe that the issues Dana raises are significant and central to the

> >functioning of the committee. I think that this planning session should have

> >priority on all other business, if any, so it should be held at the

> >beginning of the retreat and last for as long as it takes.

> >On the other hand, we don't want it to become just a complaining and venting

> >session. So we need to plan an agenda that should start with a strategy

> >session. I propose that at the end of the next P&P, you, John, Tom and I get d

'd t'fy t

> >together in executive session and work on this agen a, to I en I OPICS

Sam Duraiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 3

> >and assign adequate time to each topic. We will then bring this agenda to

> >the committee and finalize it. Once decided how much time we need for this

> >session, we will fit into the retreat other topics as time allows.

> >Members are requested to provide us with some of the issues, in addition to

> >those raised by Dana, that should be discussed during this session.

> >Regarding the location for the retreat, I am open to staying in Rockville if

> >necessary. On the other hand, I don't see why meeting in Las Vegas and

> >visiting Yucca Mountain would be a boondagle. This is not an unnecessary

> >trip; we will work in retreat in Las Vegas as we will in Rockville. Steve is

> >correct when he says that Yucca Mountain is important for this committee,

> >and we will have to visit the site at some point. On the other hand, if it

> >is a controversial issue as it appears to be, we should bring it to the

> >membership and vote on it in October. After all, John may tell us that we

> >have no money for this trip and that will close the issue.

> >Mario

> >- Original Message

> >From: "George Apostolakis" <apostola@MIT.EDU>

> >To: "Powers, Dana A" <dapower@sandia.gov>; "BONACA,Mario"

> ><mvbonaca@snet.net>; "FORD, F. Peter" <FPCTFord@aol.com>; "KRESS, T.S."

> ><TSKress@aol.com>; "LEITCH, Graham" <gmleitch@aol.com>; "ROSEN, Steve"

> ><historyart@computron.net>; "SHACK, Bill" <wjshack@anl.gov>; "sieber, JACK"

> ><jdsieber@aoJ.com>; "WALLIS. Graham B." <Graham.B.Wallis@Dartmouth.edu>

> >Cc: <sxd1@nrc.gov>

> >Sent: Tuesday. September 17, 2002 2:16 PM

> >Subject Re: Members' Issues for the ACRS Retreat: Mario's Charge

> > > Mario:

> > > I defer to you on what to do with these proposals.

> > > George

> > > At 08:56 AM 9/17/2002 -0600, Powers. Dana A wrote:

> > > >The current plan for the ACRS retreat relegates issues of ACRS process

> >and

> > > >procedures to a four hour period - undoubtably at the end of the ordeal.

> >The

> > > >rest of the time is to be spent on technical issues that really seem to

> >be

> > > >more appropriate for regular committee meetings. Perhaps it is the case

• Sam Duraiswamy - Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 4

> >that

> > > >there are no real members' issues to discuss. But, it does seem

> >worthwhile

> > > Dr. G.E. Apostolakis

> > > Professor of Nuclear Engineering

> :- > Professor of Engineering Systems

> > > Room 24-221

> > > Massachusetts Institute of Technology

> > > Cambridge, MA 02139-4307, USA

> > > e-mail: apostola@mit.edu

> > > tel: +1-617-252-1570

> > > fax: +1-617-258-8863

> Dr. G.E. Apostolakis

> Professor of Nuclear Engineering

> Professor of Engineering Systems

> Room 24-221

> Massachusetts Institute of Technology

> Cambridge, MA 02139-4307, USA

> e-mail: apostola@mit.edu

> tel: +1-617-252-1570

> fax: +1-317-258-8863 cc:

<SXD1@nrc.goY>, "John Larkins" <JTL@nrc.goY>

~a~_ Duraiswamy

• Re: Members' Issues for the ACRS Retreat: Mario's Charge Page 1 From:.

George Apostolakis <apostola@MIT.EDU>

To:

"Mario V. Bonaea" <mvbonaea@snet.net>, <RANSOM@ECN.PURDUE#062#EDU>,

"Powers, Dana A" <dapower@sandia.gov>, "FORD, F. Peter" <FPCTFord@aol.eom>, "KRESS, T.S."

<TSKress@aol.eom>, "LEITCH, Graham" <gmleiteh@aol.eom>, "ROSEN, Steve"

<historyart@eomputron.net>, "SHACK, Bill" <wjshaek@anl.gov>, "sieber, JACK" <jdsieber@aol.eom>,

"WALLIS, Graham B." <Graham.B.Wallis@Dartmouth.edu>

Date:

9/21/023:51 PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge Mario:

I agree with you.

George At 03:56 PM 9/21/2002 -0500, Mario V. Bonaea wrote:

>George,

>1 did not mean to prejudge how much time it will take. Four hrs may be

>enough, but we had to look at the list of issues being raised, estimate what

>it will take to cover them effectively and give this planning session the

>time needed. We should also conduct this session early during the retreat,

>so that i~ doesn't get pushed to the last minute, when members are already

>half way out of the door.

>Mario

>----- Original Message -

>From: "George Apostolakis" <apostola@MIT.EDU>

>To: "Mario V. Bonaea" <mvbonaca@snet.net>; <RANSOM@ECN.PURDUE#062#EDU>;

>"Powers, Dana A" <dapower@sandia.gov>; "FORD, F. Peter" <FPCTFord@aol.com>;

>"KRESS, T.S." <TSKress@aol.com>; "LEITCH, Graham" <gmleitch@aol.com>;

>"ROSEN, Steve" <historyart@eomputron.net>; "SHACK, Bill" <wjshack@anl.gov>;

>"sieber, JACK" <jdsieber@aol.com>; "WALLIS, Graham B."

><Graham. B. Wallis@Dartmouth.edu>

>Cc: <sxd1@nrc.gov>; "John Larkins" <JTL@nrc.gov>

>Sent: Friday, September 20, 2002 1:06 PM

Subject:

Re: Members' Issues for the ACRS Retreat: Mario's Charge cc:

<SXD1@nrc.gov>, "John Larkins" <JTL@nrc.gov>

JjiJA{ /1

~~

• Sam Duraiswamy - Comments on the proposed meeting dates Page i From:

"Powers, Dana A" <dapower@sandia.gov>

To:

"'sxd1@nrc.gov'" <sxd1@nrc.gov>

Date:

9/17/0210:29AM

Subject:

Comments on the proposed meeting dates Sam, You asked for comments on the proposed meeting dates for CY2003. I note that the meeting for June conflicts with the ANS meeting in San Diego.

I have to attend these meetings for Sandia now. Shifting the meeting to the next week would be particularly good. If not that, starting the meeting on Thursday would be most helpful.

Dana cc:

"APOSTOLAKIS, George" <apostola@mit.edu>, "BONACA,Mario"

<mvbonaca@snet.net>, "FORD, F. Peter" <FPCTFord@aol.com>, "KRESS, T.S." <TSKress@aol.com>,

-"LEITCH, Graham" <gmleitch@aol.com>, "ROSEN, Steve" <historyart@computron.net>, "SHACK, Bill"

<wjshack@anl.gov>, "sieber, JACK" <jdsieber@aol.com>, "WALLIS, Graham B."

<Graham. B.Wallis@Dartmouth.edu>

yJ1i/l1~ I(

'15

~am Durais.....amy - 2003 Meeting Schedule Page 1 From:

<GMLeitch@aol.com>

To:

<mvbonaca@snet.net>, <JDSIEBER@aol.com>, <dapower@sandia.gov>,

<g raham.b. wallis@dartmouth.edu>, <wjshack@anl.gov>, <historyart@computron.net>,

<TSKress@aol.com>, <FPCTFord@aol.com>, <apostola@mit.edu>, <ransom@ecn.purdue.edu>

Date:

9/18/02 11 :37AM

Subject:

2003 Meeting Schedule I notice that there is an ACRS Meeting scheduled for Sept. 3, 4, and 5 2003.

Assuming that we would not have a meeting Labor Day week, I scheduled a foreign trip that conflicts with these dates. I could do Sept. 10, 11, 12.

and 13 (if Saturday is necessary). Would any of the members have a problem moving back a week? We can discuss this at our October 2002 meeting at which time we plan to finalize the schedule. Thanks Graham L.

cc:

<MWW@nrc.gov>, <JNS@nrc.gov>, <AXS3@nrc.gov>, <RPS1@nrc.gov>.

<MTM@nrc.gov>, <HJL@nrc.gov>, <JTL@nrc.gov>, <MME@nrc.gov>, <SXD1@nrc.gov>,

<NFD@nrc.gov>, <AWC@nrc.gov>, <PAB2@nrc.gov>, <SXB@nrc.gov>

Page 1

_~~m ~~~aiswamy - Re: Comments on the proposed meeting dates From:

<GMLeitch@aol.com>

To:

<dapower@sandia.gov>, <sxd1@nrc.gov>

Date:

9/18/02 11 :39AM

Subject:

Re: Comments on the proposed meeting dates Either of Dana's proposals for the June 2003 meeting dates is OK with me.

Graham L.

cc:

<apostola@mit.edu>, <mvbonaca@snet.net>, <FPCTFord@aol.com>,

<TSKress@aol.com>, <historyart@computron.net>, <wjshack@anl.gov>. <JDSIEBER@aol.com>,

<Graham.B.Wallis@Dartmouth.edu>

Sam Duraiswamy - Re: 2003 Meeting Schedule Page 1 From:

"Steve Rosen" <historyart@computron.net>

To:

<GMLeitch@aol.com>, <mvbonaca@snet.net>, <JDSIEBER@aol.com>,

<dapower@sandia.gov>, <graham.b.wallis@dartmouth.edu>, <wjshack@anl.gov>. <TSKress@aol.com>,

<FPCTFord@aol.com>, <apostola@mit.edu>, <ransom@ecn.purdue.edu>

Date:

9/18/02 2:20PM

Subject:

Re: 2003 Meeting Schedule I would also prefer September 10.11 and 12.

-- Original Message From: GMLeitch@aol.com To: mvbonaca@snet.net. ; JDSIEBER@aol.com ; dapower@sandia.gov; graham.b.wallis@dartmouth.edu ; wjshack@anl.gov ; historyart@computron.net ; TSKress@aol.com ;

FPCTFord@aol.com ; apostola@mit.edu ; ransom@ecn.purdue.edu Cc: MWW@nrc.gov ; ~INS@nrc.gov ; AXS3@nrc.gov ; RPS1@nrc.gov ; MTM@nrc.gov ; HJL@nrc.gov ;

JTL@nrc.gov ; MME@nrc.gov ; SXD1@nrc.gov ; NFD@nrc.gov ; AWC@nrc.gov ; PAB2@nrc.gov ;

SXB@nrc.gov Sent: Wednesday, September 18, 2002 10:36 AM

Subject:

2003 Meeting Schedule I notice that there is an ACRS Meeting scheduled for Sept. 3, 4, and 5 2003. Assuming that we would not have a meeting Labor Day week, I scheduled a foreign trip that conflicts with these dates. I could do Sept.

10, 11, 12, and 13 (if Saturday is necessary). Would any of the members have a problem moving back a week? We can discuss this at our October 2002 meeting at which time we plan to finalize the schedule.

Thanks Graham L.

cc:

<MWW@nrc.gov>, <JNS@nrc.gov>, <AXS3@nrc.gov>, <RPS1@nrc.gov>,

<MTM@nrc.gov>, <HJL@nrc.gov>, <JTL@nrc.gov>, <MME@nrc.gov>, <SXD1@nrc.gov>.

<NFD@nrc.gov>. <AWC@nrc.gov>, <PAB2@nrc.gov>, <SXB@nrc.gov>

~11 II tf0

~am D~raiswamy - Re: 2003 Meeting Schedule Page' From:

George Apostolakis <apostola@MITEDU>

To:

"Steve Rosen" <historyart@computron.net>, <GMLeitch@aol.com>,

<mvbonaca@snet.net>, <JDSIEBER@aol.com>, <dapower@sandia.gov>,

<graham.b.wallis@dartmouth.edu>, <wjshack@anl.gov>, <TSKress@aol.com>, <FPCTFord@aol.com>,

<ransom@ecn.purdue.edu>

Date:

9/18/02 9:54PM

Subject:

Re: 2003 Meeting Schedule I don't mind moving the September meeting to the second week. Please remember that it should be 9/11-13 (Thursday - Saturday). If it stays in the first week, it should be 9/4-6.

George At 01:20 PM 9/18/2002 -0500, Steve Rosen wrote:

>1 would also prefer September 10, 11 and 12.

>- Original Message

>From: <mailto:GMLeitch@aol.com>GMLeitch@aol.com

>To: <mailto:mvbonaca@snet.net.>mvbonaca@snet.net. ;

><mailto:JDSIEBER@aol.com>JDSIEBER@aol.com ;

><mailtodapower@sandia.goY>dapower@sandia.goY ;

><mailto:graham.b.wallis@dartmouth.edu>graham.b.wallis@dartmouth.edu :

><mailto*.wjshack@anl.gov>wjshack@anl.goY ;

><mai!to: historyart@computron.net>historyart@computron.net ;

><mailto:TSKress@aol.com>TSKress@aol.com ;

><mailto:FPCTFord@aol.com>FPCTFord@aol.com ;

><mailto:apostola@mit.edu>apostola@mit.edu ;

><mailtoransom@ecn.purdue.edu>ransom@ecn.purdue.edu

>Cc: <mailtoMVJV\\/@nrc,goY>MWW@nrc.goY ; <mailto:JNS@nrc.goY>JNS@nrc.goY ;

><mailtoAXS3@nrc.goY>AXS3@nrc.goY ; <mailto: RPS 1@nrc.goY>RPS1 @nrc.goY ;

><mailto:MTM@nrc.goY>MTM@nrc.goY ; <mailto:HJL@nrc.goY>HJL@nrc.goY ;

><mailto:JTL@nrc.goY>JTL@nrc.goY : <mailto:MME@nrc.goY>MME@nrc.goY ;

><mailto:SXD1@nrc.gov>SXD1@nrc.goY ; <mailto:NFD@nrc.goY>NFD@nrc.goY ;

><mailto:AWC@nrc.goY>AWC@nrc.goY; <mailto:PAB2@nrc.goY>PAB2@nrc.gov ;

><mailto:SX8@nrc.gov>SXB@nrc.goY

>Sent Wednesday, September 18, 2002 10:36 AM

>Subject. 2003 Meeting Schedule

>1 notice that there is an ACRS Meeting scheduled for Sept. 3,4, and 5

>2003. Assuming that we would not have a meeting Labor Day week, I

>scheduled a foreign trip that conflicts with these dates. I could do Sept.

>10, 11, 12, and 13 (if Saturday is necessary). Would any of the members

>have a problem moving back a week? We can discuss this at our October 2002

>meeting at which time we plan to finalize the schedule. Thanks Graham L.

cc:

<MWW@nrc.goY>, <JNS@nrc.goy>. <AXS3@nrc.goY>, <RPS1@nrc.goY>,

<MTM@nrc,goY>, <HJL@nrc.goY>, <JTL@nrc.goY>, <MME@nrc.goY>, <SXD1@nrc.goY>,

<NFD@nrc,goY>, <AWC@nrc.goY>, <PAB2@nrc.goY>, <SXB@nrc.goY>

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UNITEO,STATES NUCLEAR REGULATORY COMMISSION AOVISORY COMMITTEE ON REACTOR SAFEGUAROS WASHINGTON, D.C. 2D555*0001 October 18, 2001 MEMORANDUM TO:

FROM:

SUBJECT:

LETTER FROM DR. KENNETH D. BERGERON REGARDING TENNESSEE VALLEY AUTHORITY LICENSE AMENDMENT REQUEST TO ALLOW TRITIUM PRODUCTION AT THE WATTS BAR NUCLEAR POWER PLANT The purpocse of this memorandum is to forward information received from Dr. Kenneth D.

Bergeron, formerly associated with Sandia National Laboratories, and now a member of the public, concerning the Tennessee Valley Authority license amendment request to allow tritium production at the Watts Bar Nuclear Power Plant. Dr. Bergeron has raised issues concerning tritium production in commercial nuclear power plants and deterministic versus risk-informed approach in NRC review process. He encourages use of probabilistic methods to supplement traditional analysis in evaluating the Watts Bar license amendment request.

I understEnd that the staff is reviewing the issues in Dr. Bergeron's letter. The ACRS would like to be kept informed of the staff's disposition of this matter.

At:achment:

Letter dated September 13, 2001, from Kenneth D. Bergeron to Dana A. Powers, Advisory Committee on Reactor Safeguards cc w/atts:

A. Vietti-Cook, SECY J. Craig, EDO I. Schoenfeld, EDO S. Collins, NRR B. Sheron, NRR A. Correia, NRR M. Padovano NRR A. Thadani, RES cc wlo attach: K. Bergeron

A*eIJ.D~ D. Berl."On, PbD 17 Tierra ltfonu 1',;£

.AlbuquUfju:, "~f 67122 e-mail: !eIJberG.i.ol.Sh.D~

September 13, 2001 Dr. Dana Powers Advisory Committee en Reactor Safeguards U.S. Nuclear Regulatory Commission MaiIstop 1*2 £26 W..~hl.jg(on. DC 20~SS Dear Dana.

Enclosed is I lener Jjust sent to Brian Sheron at }I.~. Its message is intended as much for the ACRS as for him, since J believe that the staff'reviewini the Wins Bar License Amendment Request \\!rill neee! guidance from the most senior levels ofthe NRC to understand tb,..'t probabilistic methods should playa central role in their review.

lfthe ACRS or one ofits S'lJbcommittees includes the Wins Bar LAR in I future meetin,g 2Benda., please let me know (e-mail is fine),as 1w01Jld like to have some input into the meeting.

Sincerely,

\\~

Ken Bergeron to(

II A-rnJ'1f'lh D. Bergeron, PhD J7 Tj~ It/onu1\\'E

.AlblJf{lJ~rque, !\\"M 87122 e-mail: l*enberg,@nllSb.n~

September 13,2001 Dr. Brian Vt'. Sheren NRR/ADPT US Nuclear Regulatory Commission Mailstop O-S E7 Washing1on, DC :ZOSSS

Dear Dr. Sheron,

Jam -u.nting to you about the onJoing stafT'review ofTVA's License Amendment Request that would allow production of tritium at the Watts Bar plant. J have a specific suggt:stion in that regard, but before getting jnto it, I'd like to re-introduce myself to you.

In the late 19805, J worked for you as a manager of one of the sroups at Sandia doing*

research on severe accidents. My group's principal foc-us was the COl\\'TAlN' code and performing studies with it for ""Re. J remember a number ofvery stimulating meetings with you after you took over the severe accident program for.RES. Around 1989. not 10Tlg after you moved into RES, J got out ofl\\"RC work in order to manage Sandja's suppon to DOE's New Production Reactor, which was intended to replace Savannah River's K reactor lS the source oftritium for the US nuclear arsenal. For a containment specialist like me, this 'WIS a very exciting time. because the government and its industry paT1ners on the Heavy \\\\'ater design 'Were committed to building the most severe accident-proof c.ontainment in history. J had the job of coordinating severe-accident related work at Sandia, Argonne. Brookhaven aTld SavanTlah Ri.... er. and it was very satisfying to be able to apply some of the lessons from TM] to the design ora reactor that

'Wes actually going to be built (or so \\\\'e thought).

AJJ that chanted in J992 when progress on nuclear arms reductions allowed President Bush to defer the tritium production program (the reason being that the tritium from decommissioned weapons could be used to replenish the weapons that remained in the arsenal). J then found other work 8J Sandia in international programs. but in ]994 Nestor Oniz ashd me to return to his program and manage all NRC work on se\\'ere accident compu1er codes. So I was responsible for not only CONTAm. but also MELCOR.

VlCTORJA, JFCI. RADTRAD {p,ctual1y an ~"RR project) and a number ofanalysis projects for RES and NRR. J continued in that roYe until Jretired in ]999 after 25 years at Sandia.

This little biography is rerevant to the \\VatlS Bar tAR because it shows that I'm pretty knowledgeable about tritium production and severe reactor accidents, paT1icularly from the perspective ofcontainment. It turns out. too, that) know quite a bi1 about TVA"s ice condenser plants, since they were a big focus for the CONTAm project during the

Dr. Brian II,

September 13,2001 Containment Perforrr~nce lmpro\\'ements program in the 'SOs, and since one of the lut projects I worked on a1 Sandia was the project to resolve DCH for lee Condensers. In that project I found myself in the unusual position of actually doing the COl\\"TAIN ealaJlations for the project leader, Many Pilch. This is because most orthe people who kMW how to run CO~" AIN had left the program Dr retired.

My pro~ess!o~l experience v.;th jet eondensers and tritium production lead me to have iTlve mIsgIvings about DOE's plans to obtain weapons tritium by having TVA produce it in the normal course of electricity production at their \\\\1at15 Bar and Sequoyah plants. J believe that the modifi:ations to the ruetor and the added mission for the nuclear management team at TVA \\ll-i)) Idd significantly to the already serious safety problems with these plants. J \\IoiJl, ofcourse, deail the reasons for my concerns in my comments to the licensing Project Manager, Mark Padovano \\\\'hat I want to ask you is on a higher level than such details. I want to encourege you to insist that the powerful new tools Dr Risk-Informed Regulation be brought to bur fully on this license amendment.

I v,-as alarmed to set the schedule Mr. Padovan distributed at the August 20 meeting It White Flint. He showed the h'RC review process being complete by early March 2002.

Such I compressed schedule is completely inconsistent with a thorough assessment even jfno element of Reg Guide 1.174 is brought to the review. As an aside, if the schedule is said to have actually begun in April2001 well I have to cry "foul," since in May I asked h"RC ~)' e-mail when the LAR was expected and was informed that it would not be until bte summer. I had asked to be kept informed about this and received no notification

\\mtil Padovan e-mailed me on August] 3 about the August 20 meeting.

In other recent public information, ~'RC has indicated they were planning for I yearlong review, so perhaps I should not focus too much on Padovan's handout. But what that dOC\\.iment suggests to me is that the starris !Ssuming that this license amendment wilJ be reviewed only via deterministic methods, with no additional insight brought in from risk methods.

For this LAR, J strongly encourage you to take fun advantage ofthe authority the Commission has given your starrto use probabilistic methods to supplement the incomplete picture that traditional anal)'sis provides. There &re many important reasons:

1. For most containment f)'pes, Design Basis analysis is not a bad sUTTogate for assessing the overall level ofprotection that the containment adds to the safety of the plant. For ice condensers, the DBA is almost irrelevant as a test for robustness. The ice does I great job with a DEGB LOeA, if)'ou ever were to have one, but it has the effect of increasing hydrogen concentrations in more risk-significant accidents, making the rea]

sarety problems worse. Put simply, it is impossible to gauge the effectiveness ofthe ice condenser containment s)'stem with traditional deterministic analysis.

2. It is also impossible to evaluate the true effect orthe core modifications on the safety of the plant via deterministic analysis. It is my guess that the principal effect will be on the complexity of fuel handling. and that new event pathways wilJ be impor1ant contributors to increa~ed risk. I also think that the likelihood oraccidents induced by

Dr. Brian W. Sheron

-3 Rbouge may be ina-e!Std because ofthe plant's new defense mission. Obviously, orJy ~el D PRA can address such effects.

3. A s1pUrant source of.dded risk is the burden that this new milit!T)' mission places on the overatl management orthe plant. There will be many new ways that ml.T'~1 commitment to I safety culture at the plant could be compromised. A 10r-ru:-i utility might be able to rise to such challenges and ensure that the.

~merrt 10 wety remains the highest priority, but TVA has sho\\\\'n itself not to be mthis cllSS. Moreover, TVA's motivation for cooperating with DOE in this pannership is troubling. Most knowledgeable observers believe that TVA is coopen:!ing only because by becoming effectively a pan orthe nuclear weapons complex the Igenc)' will be less wlnerable to those in Congress who for years hzve been trying to disband and privatize it. The c.onfiicted motivational situation at the highesl management level does not bode well for maintaining an adequate wet)'

c:uhLll"'C mthe plant. It may be difficult to assess the subtle effects of compromised mar.agement commitment, but we all know that such effects are re.aland can be lar~e.

It is inQJmbent upon the NRC to address the issue, and it is only throl.lgh risk methods that this can be done.

~. Normally, the starr might hesitate to apply risk methods when the lieensee doesn't voJumm such anal)'ses. btcause the l'-'RC has a responsibility to avoid imposing UTlnecesS!Jj' burdens on the licensee. The streamlining of many processes and rtgubtioTlS in recent yUIS has been motivated by this philosoph)' because orihe concc:m that oyer-regulation might threaten the viability orthe nuclear industry itself.

Sucll reasoning is inelevant in this case. The nuclear industry gets no benefit from these cnanEes (in fact, J believe it will be damaged by it in the long run because of public concerns about mixing militaT)' and civilian missions). The cost of the tAR and its re....iew is not coming nom ratepl)'ers but ITom the DOE, which is saving billions by not ha..ing to build a dedicated production facility.

S. Time is not ofthe essence. DOE's schedlJle for producing tritium by 2005 is.

ridiculous eX2Bgeration. 11 ignores the anns reductions dictated b)' 5TART-II, which has been ratified by both Russia and the US. The respected physicist Frank von Hippe} (rormer Assistant Director for National Security at OSTP) estimates that we won't rean~, need new tritium until 2029 or later.

6. This is an eX1raordinaril)' sensitive Federal interagency issue. Never berore have two giant 2gencies, each with compJex agendas quite difrerentrrom l'."RC's,joined forces 10 demand concunence from your licensing organization on an operating license change. An possible resources should be made available to )'our reviewers, and the overall process should come under the most intense scrutiny by senior management and the Commission itself. 1believe firml)' that this license amendment request.

ntisfies the criterion cited in RlS 2001-02, that the change "could create 'special circumstances' under which compliance with existing regulations may not produce the intended or expected level orsaret)' and plant operation rna)' pose an undue risk to public health and safety." Thererore use ofrisk-irliormed methods is appropriate. J would go fanher and sa)' that not to use the much*...aunted RG-1.174 methods in these extraordinary circumstances would be irresponsible in the highest degree. ]t would

{p 1

Dr. Brian September 13, :2001

~"a;nly streng-then the case ofcritks who see risk-informed regulation as nothing but a way for licensees to be relieved of any safet)" requirements they dislike.

[

JTp.ccgniu that the NRC is in I very uncomfortable position because ofthis License Amendment Request. But the recent, terrible events of this week show only too clearly that the price ofreEUlatory complacency can be incalculably high. J sugsest to you that the only rational WlY for you 10 proceed is cautiously, using the best scientific tools 1\\'Iilable.

I would be glad to discuss this matter \\1rith you or your starr further, if you so desire.

J have taken the liberty of sharing this letter with some ormy former colleagues who are members of the ACRS.

Sincerely,

~.~

Kenneth Bergeron Copies to:

'Q.Powm T.Kress G. ApoS101akis

September 10, 2002 MEMORANDUM TO: John T. Larkins, Executive Director Advisory Committee on Reactor Safeguards FROM:

Herbert N. Berkow, Project Director Project Directorate II IRA!

Division of Licensing Project Management

SUBJECT:

LEITER FROM DR. KENNETH D. BERGERON REGARDING TENNESSEE VALLEY AUTHORITY'S LICENSE AMENDMENT REQUEST TO PRODUCE TRITIUM AT THE WAITS BAR NUCLEAR POWER PLANT Your memorandum of October 18, 2001, to William D. Travers forwarded Dr. Kenneth D.

Bergeron's letter of September 13, 2001. Dr. Bergeron was concerned about the ongoing U.S. t\\'Jclear Regulatory Commission's review of Tennessee Valley Authority's (TVA's) license amendment request to irradiate tritium-producing burnable absorber rods (TPBARs) at the Watts Bar nuclear plant. You requested that the Advisory Committee on Reactor Safeguards be kept informed of the staff's disposition of this matter.

Attached is Brian W. She ron's letter of September 6,2002, to Dr. Bergeron responding to his letters of Se~tember 13, 2001, and January 16, 2002. We expect to issue license amendments to TVA this month in response to TVA's amendment requests of August 20, and September 21, 2001, to irradiate TPBARs in the Watts Bar and Sequoyah reactors.

Attachment:

Letter to Dr. Bergeron dated 9/6/02 cc wI attachment:

A. Vietti-Cook, SECY J. Craig, EDO I. Schoenfeld, EDO A. Thadani, RES CONTACT: L. Mark Padovan, NRR 415*1423

September 6, 2002 Dr. Kenneth D. Bergeron 17 Tierra Monte NE Albuquerque, NM 87122

SUBJECT:

NRC STAFF RESPONSE TO YOUR SUGGESTIONS TO RISK-INFORM THE REVIEW OF THE SEQUOYAH AND WAITS BAR TRITIUM PRODUCTION LICENSE AMENDMENT REQUESTS

Dear Dr. Bergeron:

r am responding to your letters of September 13, 2001, and January 16, 2002, requesting that we risk-inform our process for reviewing Tennessee Valley Authority's (TVA's) license amendment requests to produce tritium at Sequoyah and Watts Bar, and expressing other safety concems. We reviewed your written requests and evaluated your concerns expressed during the November 7,2001, meeting held at One White Flint North.

As you are aware, RIS-2001-002 "Guidance on Risk-Informed Decision Making in License Amendment Reviews," addresses our process for determining when requests for risk information are justified as part of our review of a license amendment request. We conducted a technical assessment of the issues you identified following the guidance in RIS-2001-002. We were not able to substantiate that there would be a significant increase in risk if the U.S.

Nuclear Regulatory Commission (NRC) approved TVA's amendment requests. However, we elevated your concerns to the risk informed licensing panel (RILP) even though our staff's assessment did not identify any issues that would raise questions about TVA's ability to maintain adequate protection of public health and safety. The RILP convened on July 11, 2002, and unanimously agreed that gathering additional risk information to evaluate TVA's amendment requests was not necessary. However, in our July 29,2002, letter to TVA, we did ask TVA to send us some risk-informed background information to confirm our decision. In your email of August 10,2002, to Mark Padovan of the NRC, you asked for a copy of TVA's response to the staff's request for information. TVA's August 9, 2002, response is enclosed.

Your letters noted numerous safety concerns. NRC staff considered each of your concerns against the guidance of RIS-2001-002, but grouped the concerns into the following broad categories:

• historical safety performance of Sequoyah and Watts Bar

• postulated increased risk from intemal events, extemal events, and security concems stemming from the dual-purpose civilian and military-related uses of the TVA reactors

• potential ice condenser plant design vulnerabilities to severe accident conditions, in particular, under station blackout (S80) scenarios

K. Bergeron

-2

• other issues, such as NRC's legal authority to issue the amendments, Advisory Committee on Reactor Safeguards (ACRS) participation in the amendment reviews, and more time for public comments The staff's assessment of your concerns is provided below.

Regarding your concerns about TVA's performance. the staff does not use overall plant performance as a criterion for approving amendment requests. The NRC's reactor oversight process (ROP) continuously monitors licensee performance to provide assurance that licensees are operating plants safely and in accordance with the regulations and licensing bases. The ROP allows for a graded, predictable agency response commensurate with licensee performance. This can result in agency actions up to and including ordering the plant to shut down should NRC determine performance to be unacceptable.

The ROP relies on objective performance indicators (Pis) along with risk-informed inspections using 39 inspection procedures to monitor and evaluate plant performance. As discussed in the most recent Annual Assessment Letters for Watts Bar and Sequoyah. the results of the Pis and inspections are in the "licensee response band" of the ROP Action Matrix. This means that both plants have acceptable performance that does not require additional oversight beyond the haseline level of inspection. Plant performance results are available for public view on the NRC's external website at http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/index.html.

You also expressed concern that Watts Bar may not be capable of operating with a tritium production core. The staff notes that TVA has successfully demonstrated its ability to irradiate and handle tritium producing burnable absorber rods (TPBARs). TVA successfully irradiated 32 lead test assemblies for one cycle as part of TPBAR efficacy testing at Watts Bar.

Therefore, the staff does not have any basis to question TVA's capability to manage such a change.

You postulated several new accident scenarios in your letter of January 16, 2002. In particular, you were concerned that a TPBAR ejection was not evaluated. Each TPBAR has a threaded end plug that is connected to a hold-down plate. The TPBAR is also secured in place via a crimping device as described in TVA's submittal of August 20. 2001. The TPBARs are inserted into fuel assemblies, similar to traditional burnable poison rod assemblies, and do not contain fissile material. Immediately above the fuel assemblies containing TPBARs is the upper core plate and reactor vessel upper internals package. Therefore, the staff does not agree that a realistic scenario exists for TPBARs to be ejected, or that there is a significant increase in initiating event transients.

You also noted that you believed it was not an appropriate neutronic practice to offset. by soluble boron poisoning, additional reactivity from higher fuel enrichment. Changes to the core design and core reactivity issues will be fully addressed in the staff's safety evaluation.

However, the staff did not identify in its deterministic design basis review any reactivity issues that would warrant probabilistic treatment of TVA's amendment requests.

You suggested that a potential TPBAR drop accident "during the TPBAR consolidation process" was not adequately addressed. You note that the TPBAR drop accident could occur with the

K. Bergeron

-3 plant at full power. The rod consolidation process is performed in the spent fuel pool and, as such, does not increase the likelihood of a reactor trip. From a dose perspective, TVA addressed dropping a TPBAR and NRC staff evaluated it for (1) fuel movement in the reactor cavity and (2) spent fuel pool operations. All of the fuel rods in an irradiated fuel assembly, and 24 TPBARs, are assumed to rupture, releasing the radionuclides within the fuel-clad gap to the fuel pool or reactor cavity water. TVA's analyses show the offsite consequences of this event are well within Title 10, Code of Federal Regulations (10 CFR), Part 100, dose guidelines. A complete radiological assessment of potentially dropped TPBARs will be addressed in the staff's safety evaluations.

Previous performance issues with the ice condenser system were also noted in your January 16, 2002, letter. For example, you noted problems with lower inlet door binding for both plants. These issues have been corrected, and the staff is not aware of any recent door failures due to floor upheavaVdoor binding within the past few years. The lower inlet doors continue to be tested in accordance with each plant's Technical Specifications and are monitored under several licensee programs, including the regulatory-required 10 CFR 50.65 maintenance rule program. More important, there is no direct nexus between a change in the core design and any effect on the reliability or availability of the ice condenser system.

Therefore, overall, given no demonstrated significant increase from the baseline core damage frequency. and no demonstrated significant change in containment systems performance, the staff could not substantiate that there would be a significant increase in the baseline severe accident large early release frequency because of tritium production.

In your letters, and during the November 7,2001, meeting, you noted concerns that safeguards measures at Sequoyah and Watts Bar may be inadequate once tritium production begins at these stations, especially in view of the events of September 11, 2001. The NRC and its licensees have taken a number of actions following the terrorist attack of September 11, 2001, to increase security at NRC-licensed facilities, including a heightened security stance pursuant to safeguards advisories. On February 25, 2002, the NRC issued Orders to all commercial nuclear power plants to implement interim compensatory measures for the current threat environment. Some of the requirements made mandatory by the Orders formalize the security measures that NRC licensees had taken in response to NRC's advisory letters. The specific actions are sensitive, but generally include requirements as follows:

increased patrols augmented security forces and capabilities additional security posts installation of additional physical barriers checks at greater stand-off distances enhanced coordination with law enforcement and military authorities restrictive site access for all personnel

K. Bergeron

-4

• additional security measures pertaining to waterways and the owner-controlled land outside the plants' protected areas During our meeting of November 7,2001, you raised a specific terrorist scenario against Watts Bar. Further, you alluded to this postulated vulnerability in your letter of January 16, 2002. Although the exact scenario you described is not evaluated in the plant's Updated Final Safety Analysis Report (UFSAR), the effects of the scenario had been analyzed for design basis considerations and are documented in the UFSAR. Under such a scenario, the specific plant structures and systems of interest to your concem are protected from such a phenomenon. The analysis used bounding design-basis assumptions and conditions beyond the nominal conditions that would be present from the scenario that you postulated. This phenomenon was also evaluated in the licensee's individual plant examination (IPE) of external events submittal. The staff concludes that the outcome of the scenario you postulated during our meeting and in your letter is not credible.

On the matter of the NUREG/CR-6427, "Assessment of the DCH [Direct Containment Heating]

Issue for Plants with Ice Condenser Containments," the staff is in the process of resolving Generic Safety Issue (GSI)-189, "Susceptibility of Ice Condenser Plants and Mark III Containments to Early Failure from Hydrogen Combustion During a Severe Accident."

Although NUREG/CR*6427 notes a 0.97 conditional containment failure probability (CCFP) for Sequoyah under SSO conditions, this CCFP value results from assumptions that were appropriate for resolving the Direct Containment Heating issue and must be used in the correct context. The NUREG gives no credit for offsite power recovery, and provides no evaluation of recovery of one of several simultaneously failed emergency diesel generators. The NUREG also does not reflect plant improvements since the licensee's original IPE submittal that reduces the frequency 01 SBO and reduces the likelihood of core damage during SSO conditions such as the following:

ma;ntaining high emergency diesel generator reliability

• a maintenance rule program

• high-temperature reactor coolant pump seals modifications to the turbine driven auxiliary feedwater operation procedures improved emergency operating procedures More realistic treatment of SSO scenarios would probably reduce the core damage frequency, containment failure frequency, and CCFP. Also, the tritium amendment requests would not result in an increase in core damage frequency or large early release frequency above the current values. The CCFP value. as it stands today, is appropriate for its intended purpose of resolving the direct containment heating issue and use as a screening value for GSI-189 regulatory backfit analysis.

In summary, the staff evaluated your suggestions and concerns against the special circumstances criteria noted in RIS*2001-002 and against standards defined in NRC Regulatory Guide 1.174. The staff was not able to substantiate that there would be a significant increase in

K. Bergeron

-5 risk of intemal or external events because of tritium production. The staff concluded this primarily because a tritium production core in itself does not:

• increase the likelihood of an initiating event

• affect the probabilistic risk assessment (PRA) success criteria

• affect the functionality, availability, or reliability of equipment and structures necessary to prevent core damage (Level I PRA) or mitigate core damage effects (Level II PRA)

The staff determined that the only salient issue relevant to Sequoyah and Watts Bar is GSI-189, which is unaffected by TVA's amendment requests. The Office of Nuclear Regulatory Research is completing GSI-189 regulatory analysis, and will forward it to Office of Nuclear Reactor Regulation for final resolution. However, the staff does not believe that approving the amendment requests depends on resolving GSI-189 for reasons previously noted.

You commented on the NRC's legal authority to issue the amendments in light of 42 USC 7272.

This very issue was analyzed by the Atomic Safety and Licensing Board in the recent consolidated tritium license amendment proceedings. In a decision issued on July 2, 2002 (LBP-02-14), the Board concluded that Public Law 106-65, section 3134(a), which provides that the Secretary of Energy shall produce tritium at Watts Bar or Sequoyah, and its legislative history "clearly show that Congress intended for the NRC to entertain" TVA's tritium license amendment applications, notwithstanding 42 USC 7272. Thus, there should be no doubt that the NRC has the legal authority to issue the amendments.

The ACRS determines what involvement it will have reviewing licensing actions. It received your letter of October 18, 2001, on the subject of allowing tritium production at Watts Bar. The ACRS has not asked to participate in the review of TVA's amendment requests, but wanted to be informed of our response to you. Accordingly, we are forwarding a copy of this letter to the ACRS.

You also suggested that the NRC should allow more than 30 days for public comment on the staff's proposed no significant hazards consideration determinations. On January 15, 2002, Mr. David Lochbaum of the Union of Concerned Scientists sent us a letter requesting a 60-day extension of the public comment period. The letter of January 17, 2002, from the Secretary of the Commission, denied that request. However, the Secretary's letter said that the NRC staff would consider additional comments as it received them while reviewing other comments.

Likewise, we continued to assess the information in your letters of September 13, 2001, and January 16. 2002, and are now responding to your concerns.

71

K. Bergeron

-6 We appreciate your comments and suggestions regarding the amendment requests for tritium production and we hope that our response addresses your concerns. Please feel free to contact L. Mark Padovan at (301) 415*1423 or me should you have any questions.

Sincerely,.

IRA!

Brian W. Sheron, Associate Director for Project Licensing and Technical Analysis Office of Nuclear Reactor Regulation

Enclosure:

TVA letter to NRC dated 8/9/02 cc: Donald J. Moniak Community Organizer and SRS Project Coordinator Blue Ridge Environmental Defense League PO Box 3487 Aiken, South Carolina 29802 Ms. Ann P. Harris 341 Swing Loop Road Rockwood, Tennessee 37854 Dr. Gary Drinkard 340 DrinKard Dr.

Spring City, TN 37381 Ms. Vickie G. Davis TDEC-DOE Oversight Division 761 Emory Valley Road Oak Ridge. TN 37830*7072

Tennessee Valley Authority. 1101 Markel Street. Chattanooga. Tennessee 37402*2801 August 9, 2002 10 CFR 50.90 u.s. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C. 20555-0001 GeI"tlemen:

In the Matter of

)

Docket Nos.50-327, 328, 390 Tennessee Valley Authority

)

SEQUOYAH (SQN) AND WATTS BAR (WEN) NUCLEAR PLANTS -

REQUEST FOR RIS~-INFORMED INFO~~TION RE:

TRITIUM PRODUCTION PROGRAM (TAC NO. MB1884)

The purpose of this letter is to respond to NRC questions provided in a letter dated July 29, 2002.

This information is being provided to support the ongoing NRC review of WBN and SQN License Amendment Requests submitted by TVA on August 20, 2001, and September 21, 2001, respectively.

TVA has separated the responses into two enclosures. provides the SQN responses. provides the WBN responses.

a....,...**.. ***..**.... l~..,

U.s. Nuclear Regulatory Commission Page 2 August 9, 2002 There are no regulatory commitments made by this letter.

The delay in submitting this information was coordinated via telecon with the NRC staff on August 7, 2002.

If you have any questions, please contact me at (423) 751-2508.

Sincerely, J:!ra:0J~u~k Manager Nuclear Licensing Subscribed and sworn to ~r.e; me en this ct%

day of u.J.-t"

{I..J.LCL... l\\,--~~§,---__

~ary PUbl!c 1-1;.' Commission expires Enclosures cc:

See page 3

u.s. Nuclear Regulatory Commission Page 3 August 9, 2002 cc (Enclosures):

NRC Resident Inspector Sequoyah Bar Nuclear Plant 2600 Igou Ferry Road Soddy Daisy, Tennessee 37379-3624 NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. Ronald W. Hernan, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 Mr. L. Mark Padovan, Senior Project Manager U.S. Nuclear Regulatory Commission MS 08G9 One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739 u.s. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303-8931

ENCLOSURE 1 SEQOOYAH NUCLEAR PLANT (SQN)

RESPONSES

1. Please provide the SQN maintenance rule program (a) (2) performance criteria for the following systems:

A. Emergency Diesel Generators (EDGs)

B. Turbine Driven Auxiliary Feedwater Pump C. Emergency 125 VDC Supply D. Emergency 120 VAC Supply E. Hydrogen Igniters F. Containment Air Return Pans G. Emergency Raw Cooling Water (ERCW)

H. Ice Condenser TVA RESPONSE The maintenance rule program (a) (2) performance criter~a for the systems listed above is as follows:

A. Emergency Diesel Generators - Please note that the term Valid Failure is equivalent to Functional Failure (FF) and Valid Test is equivalent to valid Demand.

Unavailability - No more than 2.5% for each DG average over a rolling 24 months (438 hrs/24 months).

Function Level Unreliability - The DG target reliability of 97.5% is met provided the following trigger values are not reached:

3 combined functional failures (FFs) (start demand and/or load run demand) out of 20 combined demands (all DGs combined) 4 combined FFs out of 50 combined demands (all DGs combined) 5 combined FFs out of 100 combined demands (all DGs combined) 4 FFs out of 25 demands (for each DG)

Component Level Unreliability - No more than 2 Component (Pump) Failures (CFs) per Fuel Oil Transfer Pump per rolling 24 months.

El-l

r----------------------------------------------

B.

Turbine Driven Auxiliary Feedwater Pump Unavailability -

No more than 2.5% per train or 219 hrs/year, based on a 24 month rolling average when risk significant.

Unreliability -

No more than 1 FF per 24 months per train.

C.

Emergency 125 VDC Supply Unavailability -

No more than 0.194% or 17 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />s/year, based on a 12 month rolling average (all modes and all Outage Risk Assessment Management (ORAM) states).

Unreliability -

No more than one FF of a vital battery or vital battery board per 24 months.

D.

Emergency 120 VAC Supply Unavailability - No more than 16.4% or 60 days/year, based on a 12 month rolling average (all modes and all ORAM states).

Unreliability -

No more than four FFs of a 120 VAC vital instrument power board per 24 months.

E.

Hydrogen Igniters Unavailability - No more than 0.95% average unavailability per unit during a rolling 24 month interval when risk significant (Modes 1&2).

The function is unavailable whenever there are no functional igniters in one or more of the 34 zones.

Unreliability - No more than 1 FF per.unit during a rolling 24-month interval.

A FF in Modes 1 & 2 is 1) a loss of two igniters in the same zone, or 2) a loss of any combination of three or more igniters in any combination of zones.

When in State 11 or 12, a FF is the loss of either Train A or Train B.

£1-2 71

F. Containment Air Return Fans Unavailability - No more than 0.28% per train every 24 months when risk significant (Modes 1&2 and ORAM States 1&2).

Unreliability - No more than one FF per train every 24 months.

A FF is defined as a failure of the train to start or operate as required.

G. Emergency Raw Cooling Water (ERCW)

Unavailability - Train Level - No more than 2.7%

per train per 24-month rolling average.

Unreliability Train/Functional Level - No more than two FFs per train per 24 months.

Component (ERCW Pump) Level - No more than one failure per pump per 24 months.

H. Ice Condenser Unavailability - In Mode 1, no actual unplanned capability loss events attributable to the ice condenser system are permitted in a rolling 24 month interval.

In Modes 1 and 2 or ORAM states 1 and 2, no unavailability that if it had occurred at 100% power, it would have caused a greater than 20% power loss.

Unreliability - No failure of a required flow path is permitted in a rolling 24 month interval.

Condition No more than one failure to maintain the ice bed temperature at or below 27°F during Modes 1 and 2, ORAM States 1 and 2, and States 11 and 12 when required is permitted in a rolling 24 month period.

El-3

No failure to maintain the design basis ice mass is permitted in a rolling 24 month interval when required.

No failure to maintain the m~n~mum sodium tetraborate concentration and proper range of pH as defined in LCO 3.6.S.1.a is permitted in a rolling 24 month interval.

2. Are any of the above systems currently in maintenance rule program (a) (1) status and if so why?

TVA RESPONSE None of the systems listed in Item 1 are currently in maintenance rule program (a) (1) status.

3. How many EDG failures (failure-to-start and failure-to run) have occurred in the previous 100 starts for each of the EDGs?

TVA RESPONSE As of June 30, 2002, the number of BOG valid failures which have been recorded for the last 100 starts are as follows:

Generator BOG lA EOG lB EOG 2A EOG 2B Number of Failures 1

0 6*

0

• This data is consistent with the response to Question 2.

As indicated in the response to Question 1, the trigger criteria for each individual Sequoyah EOG is 4 FF out of 25 demands.

The maximum number of valid failures per 25 demands EOG 2A has reached in the past is 2.

4. Are any of the above EDG failures a common-mode failure of the SQN EDGs (i.e. were the other EDGe actually El-4

unavailable because the root cause of the failed EOG also actually affected the other EOGs)?

TVA RESPONSE None of the EDG failures listed in Item 3 resulted from a common-mode failure.

s. Do all reactor coolant pumps (RCPs) at SQN have the newer style high-temperature O-ring seals?

Xf not, how many do not and on which unit?

For those Reps that do not have the new O-ring design, what is the schedule to replace tham?

TVA RESPONSE All Sequoyah Reps currently have the high temperature 0 ring seals installed.

6. Does SQN conduct Severe Accident Management Guidelines (SAMG) drills and how often?

TVA RESPONSE Severe Accident Management Guidelines (SAMG) training for SQN emergency preparedness teams normally consists of classroom instruction and a table top drill and are conducted annually with the teams being trained based on a four year rotation.

7. How many failures of the ice condenser lower inlet doors have occurred during the previous two operating cycles (i.e. did not meet technical specifications surveillance requirements)?

Are any of these failures attributed to floor upheaval/buckling causing door binding?

Does Tennessee Valley Authority continue to monitor ice condenser floor growth from cycle-to-cycle?

TVA RESPONSE Surveillance instructions performed during the past two refueling outages, Cycle 10 and Cycle 11, for both Sequoyah Unit 1 and Unit 2 were evaluated for failures.

Based on the data packages reviewed, all lower inlet doors met the Technical Specification surveillance requirements.

El-S

No lower inlet door surveillance requirement failures during the specified time period were due to floor upheaval/buckling.

SQN continues to monitor ice condenser floor movement during operation under Procedure No. 0-PI-SXX-061-001.0

-Ice Condenser Lower Plenum Floor Monitoring".

This Instruction provides detailed steps for monitoring vertical movement of the ice condenser lower plenum floor to ensure lower inlet door operability.

8. Please provide the following information for SQN based on the current PRA model:

A. total core damage frequency (COF) from internal events B. total CDF from external events (if ~deled)

C. percentage of COF due to station blackout

o. loss of offsite power frequency and basi.

E. probabilities of non-recovery of offsite AC power fqr various times in the model and basis for numbers used.

F. probability of EDG/emergency AC bus recovery (if modeled) and the basis for the number(s)

TVA RESPONSE Based on Revision 01 of the Sequoyah Probabilistic Safety Analysis (PSA) model, the requested information has been established as follows.

A. The total CDF from internal events is 3.77E-05/yr.

B. The total CDF from external events has not been quantified.

In the IPEEE (Individual Plant Examination for External Events) no vulnerabilities from external events were identified.

C. The percentage of CDF due to station blackout is 10.5%.

D. The loss of offsite power frequency is 0.04BS/yr based on a Baysian update of generic industry data using site specific experience.

E. The probability of non-recovery from a loss of offsite AC power at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is 0.255.

At 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> the steam generator secondary side inventory is depleted when no makeup is available.

The probability of non-recovery El-6

at 1.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> is 0.604.

At 1.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> core damage occurs when no secondary side makeup is available. The probability of non-recovery at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is 0.275.

At 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the station batteries are depleted.

These non recovery probabilities are based on the information in NUREG/CR-5032.

F. The probability of EDG/emergency AC bus recovery within 1.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of 1/1 EOG is 0.39 and of 1/2 EOGs is 0.536.

The probability of recovery within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of 1/1 EOG is 0.60 and of 1/2 EOGs is 0.80.

The basis for these probabilities is a site specific EOG recovery model.

This model is described in detail in Section 3.3.3.4.3.2 of the individual plant evaluation.

9. What are the normal and emergency power supplies for the ERCW (intake structure> sump pumps?

TVA RESPONSE All of the sump pumps at the ERCW pumping station are.

powered from the various ERCW Motor Centrol Center (MCC>

boards.

The building basement sump pumps (not safety related) are powered from the MCC in their respective bays, the deck sump pump lA is powered from the lA ERCW 480v MCC, the deck sump pump lB is powered from the 1B ERCW 480v MCC.

All of the ERCW 480v MCC receive power from the 6.9 Kv Shutdown Boards, and are therefore Diesel backed.

The deck sump pumps are safety related and remain loaded to the Diesel after blackout, the building basement sump pumps are non-safety related and are therefore load-stripped upon blackout.

El-7

ENCLOSURE 2 WATTS BAR Nt7CLEAR PLANT (WBN)

RESPONSES

1. Please provide the WBN maintenance rule program Ca) (2) perfor.mance criteria for the following systems:

A. Emergency Diesel Generators (EDGs)

B. Turbine Driven Auxiliary Feedwater Pump C. Emergency 125 VDC Supply D. Emergency 120 VAC Supply E. Hydrogen Igniters F. Containment Air Return Fans G. Emergency Raw Cooling Water CERCW)

H. Ice Condenser TVA RESPONSE The maintenance rule program Ca) (2) performance criteria for the systems listed above is as follows:

A. Emergency Diesel Generators Unavailability No more than 2% for each DG averaged over a rolling 24 months (approximately 350 hour0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />s/24 months).

No more than 0.1% for the fuel oil transport support function for each EDG set averaged over a rolling 24 months (approximately 17 hour1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />s/24 months).

Unreliability No more than 1 failure of any of the fuel oil transfer pumps within a 24-month period.

Unreliability performance criteria for the EDG function is based on trigger values established as a result of 10CFRSO.63.

Nuclear Engineering established a target reliability of 97.5'.

These trigger values are used as unreliability performance criteria for the Maintenance Rule as follows:

E2-1 13

3 combined functional failures (FFs)

(start demand and/or load run demand) out of 20 combined demands (all DGs combined) 4 combined FFs out of 50 combined demands (all DGs combined) 5 combined FFs out of 100 combined demands (all DGs combined) 4 FFs out of 25 demands (for each DG)

B. Turbine Driven Auxiliary Feedwater Pump Unavailability - No more than 2% per train or 350 hour0.00405 days <br />0.0972 hours <br />5.787037e-4 weeks <br />1.33175e-4 months <br />s/24 months based on a 24 month rolling average.

Unreliability - No more than two FFs per train in a 24-month interval.

C. Emergency 125 VDC Supply Unavailability (Battery Board) - No unavailability of the boards are allowed during power operation lO hours).

Additionally, no unavailability is planned at othex times.

This does not include swapping the battery with the spare battery, which includes a momentary loss of backup power.

Unreliability - No more than one FF of a vital battery or vital battery board per 24 month period.

~.

Emergency 120 VAC Supply Unavailability - No more than 0.274% or 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />s/inverter/24 months interval.

The inverters are not required available during certain pre analyzed conditions during outages.

Unreliability - No more than one FF per channel per 24-month interval.

E2-2

E. Hydrogen Igniters Unavailability - No more than 7 days (168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />) during a 24 month period (Modes 1 & 2).

The system will be considered unavailable during periods in which there are no functional igniters in one or more of the 34 zones.

Unreliability - No more than one FF within a 24 month interval.

Functional failure is defined as any failure or combination thereof that results in the loss of ignition capability in any of the 34 zones.

Supplemental component level performance criteria is no more than three igniter failures in a 24 month interval.

F. Containment Air Return Fans Unavailability - No more than 1% per train per.24 months (approximately 175 hrs/train/24-months) reporting period.

Unreliability - No more than one FF per train per 24-month interval.

Functional failure is definp.d as a failure of the fans to start or operate as required.

G. Emergency Raw Cooling Water (ERCW)

Unavailability - The train unavailability performance criteria for modes 5 and 6 is 1.4%

(approximately 245 hour0.00284 days <br />0.0681 hours <br />4.050926e-4 weeks <br />9.32225e-5 months <br />s/24-months). Risk considerations preclude the elective removal of either ERCW train from service during power operation.

However, routine pump surveillance testing involves cross-tying of the trains for brief periods.

The test instructions have been reviewed against the requirements for operator recovery from planned maintenance.

It *was determined that cross-tying of trains for performance of the pump test does not require maintenance rule unavailability.

£2-3

Unreliability Train Level - No FFs per train within a 24 month interval.

Component level - No more than three component failures within a 24-month interval (ERCW pumps, strainers, and traveling water screens).

H. Ice Condenser Unavailability - No unplanned capability loss attributable to the ice condenser is permitted in a rolling 24-month interval.

Unreliability No FF due to loss of t.he mJ.nJ.mum required flew path through the ice bed within an operating cycle.

No FFs within an operating cycle where the minimum total ice mass is found to be less than that specified by the Technical Specification, and No instances within an operating cycle in which the average boron concentration or pH of the sample is found to be less than that specified by the Technical Specification Condition Not more than one failure to maintain the mean ice bed temperature below 27°F is permitted within a 24 month interval.

2. Are any of the above systems currently in maintenance rule progr~ (a) (1) status and if so why?

TVA RESPONSE The Auxiliary Feedwater (AFW) system is in (a) (1) status.

However, this is due to a start logic issue on the motor driven AFW pumps which has since been resolved.

At this E2-4

time, this equipment is being monitored for removal from (a) (1) status which is projected for 4th quarter FY03.

The Turbine Driven AFW Pump is not in (a) (1) status.

3. How many EDG failures (failure-to-start and failure-to run) have occurred in the previous 100 starts for each of the EDGs?

TVA RESPONSE As of June 30, 2002, the number of EDG valid failures which have been recorded for the last 100 starts are as follows:

Generator Number of Failures 1

0 1

0 DG lA-A:

DG lB-B:

DG 2A-A:

DG 2B-B:

4. Are any of the above EnG failures a common-mode failure of the WBN EDGs (i.e., were the other EDGs actually unavailable because the root cause of the failed EDG also actually affected the other EDGs)?

TVA RESPONSE None of the EDG failures listed in Item 3 resulted from a common-mode failure.

5. Do all reactor coolant pumps (RCPs) at WEN have the newer style high-temperature O-ring seals? If not, how many do not and on which unit? For those RCPs that do not have the new O-ring design, what is the schedule to replace them?

TVA RESPONSE All Watts Bar Reps currently have the high temperature 0 ring seals installed.

E2-5

6. Does WBN conduct Severe Accident Management Guidelines drills and if so how often?

TVA RESPONSE Severe Accident Management Guidelines (SAMG) training for WBN emergency preparedness teams normally consists of classroom instruction and a table top drill and are conducted annually with the teams being trained based on a four year rotation.

7. How many failures of the ice condenser lower inlet doors have occurred during the previous two operating cycle.

(i.e. did not meet technical specifications surveillance requirements)?

Are any of these failures attributed to floor upheaval/buckling causing door binding?

Does Tennessee Valley Authority continue to ~nitor ice condenser floor growth from cycle-to-cycle?

TVA RESPONSE Surveillance ins~ructions performed during the 3rd and 4~h refueling outages were reviewed.

Both of these performances were successfully completed with n~ doors failing their Technical specifications requirements.

No le-wer inlet door surveillance recrJirement faillu"es during the specified time period were due to floor upheaval/buckling.

WBN performs a maintenance instruction l-STRU-661-S000, "Ice Condenser Wear Slab Floor Inspection,H each refueling outage which monitors floor growth to ensure that any floor movement does not impair the opening of the lower inlet doors and prevent them from fulfilling their accident function.

8. Please provide the following information for WBN based on the current PRA model for each plants A. total core damage frequency (CDF) from internal events B. total COF from external events (if modeled)

C. percentage of CDF due to station blackout D. loss of offsite power frequency and basis E. probabilities of non-recovery of offsite AC power for various times in the model and basis for numbers used.

E2-6

F. probability of EDG/emergency AC bus recovery (if modeled) and the basis for the number(s)

TVA RESPONSE Based on Revision 2A of the Watts Bar Probabilistic Safety Analysis (PSA) model, the requested information has been established as follows:

A. The total CDF from internal events of 4.48E-5/yr.

B. The CDF from external events is not currently modeled in the WBN-PSA.

In the IPEEE (Individual Plant Examination for External Events) no vulnerabilities from external events were identified.

C. WBN has not calculated the percentage of CDF due to Station Blackout, we do calculate the percentage of CDF due to Loss of Offsite Power (LOOP) which is 14% of the CDF.

D. The loss of offsite power frequency is 0.0259/yr based on a Baysian update of a generic industry data using site specific experience. Specifically, WEN has experienced no LOOP.

E. The non-recovery of the 161-kv Grid for WBN has a mean value of 0.255.

Offsite power can also be restored to the WEN systems through the Unit 2 500-KV grid.

The non-recovery of the Unit 2 500-KV grid has a mean value of 0.205.

A Monte Carlo simulation is used in the electric power recovery analysis at WBN.

This recovery analysis for the WEN PSA model is an integrated, time dependent model that looks at several parameters and conditions.

These parameters include the recovery of offsite power, the recovery of one or two diesels, and the availability of auxiliary feedwater for heat removal.

The result of the recovery analysis is a recovery factor that is the ratio of two conditional frequencies, given a LOOP initiating event:

the conditional frequency of the loss of onsite power in a mission time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the failure to restore onsite or offsite power before core damage occurs, and the conditional frequency of onsite power failure in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period without recovery. Factors that influence the time available to restore AC power include the availability of 125V DC power (i.e., battery lifetime) and the length of time to core damage due to pump seal E2-7

leakage or power-operated relief valve (PORV) discharge following a loss of all onsite AC power.

The time to recover off-site power at nuclear power plants has been documented in NUREG/CR-5032,

~Modeling Time to Recovery of Loss of Off-site Power at Nuclear Power Plants.-

The Model for Group 12 in this NUREG was chosen as best representing WEN and was used in the WEN recovery analysis.

F. As described above, a Monte Carlo simulation (PLG STADIC program is used in the electric power recovery (offsite and OG) analysis at WBN.

Some of the assumptions used in this time dependent model are:

• The diesels generators are assumed to be unrecoverable after the depletion of the DC batteries

• The turbine-driven AFW pump is also assumed to be unav~ilable after DC control power is lost Examples of the non-recovery factors used for various conditions is provided in the following table:

C** I IN:blr Of Ioeit 1 Di...

c;.nlratorl p.valla.l:lla Po Recovery Dnit 1 Ilgi x.z,own To Initially

• • %D IK&in t.nlUlC4 AweiHar')

' **l:5".ter Av.Uat>ll Op.r.tor.

Cooll5OWD bl5 Depr.ll=i **

aes

"'-bIZ' Of 1I.covlr&t>11 Onit 1 PlUI t7n1t 2 I)ie..l Geelratorl Protlat>Uity Of ~.it.

Power.ailurl be! Offait..

)lonr.co.....ry I)il.ll Geelrat.or t!f:l......ilat>iUty Slqu.nc.

Rlcov.ry

'actor 1

2 UnJo:lOYll Y.a No 2

1.17525,4 8.30762-3 5.5072-2 2

2 UnknOYll Yel Yel a

I. 7H2'* 5 1,307'2-3 4,OU82-2 3

2 Unknown Ho

'NIl.

a

',21547-4 1.30762*3 0.2U257 4

1 Unlmown Yel No 1

'.141'4-4 6.48421-2 5.484 7-2 5

1 Unknown Yel Yel 1

6.75282-4 6.48421-2 4.0U25-2 1

UnknOYll Ho NIl.

1 4.'5071*3 6.41421,2 0,2'67' 7

0 UnluloYll Yea No 0

3.7518-2 1.0 0.14712 8

0 UIlltnOYll Y.I Y**

• 0 2.Un*2 1.0 0.10587J 0

UIlltnOYll He NIl.

0 0.15J522 1.0 0,625577 E2-B C/o

9. What are the normal and emergency power supplies for the ERCW (intake structure) sump pumps?

TVA RESPONSE There are two sump pumps per ERCW Strainer Room.

The normal power supplies for the pumps in ERCW Strainer Room A are from the safety-related Control & Auxiliary Building Vent Board lAl-A and 2Al-A (respectively).

The normal power supplies for the pumps in ERCW Strainer Room B are from the safety-related Control & Auxiliary Building Vent Board lBl-B and 2Bl-B (respectively).

These boards receive diesel power; however, these pumps are load shed from their respective board in the event of Loss of Offsite Power.

E2-9

Sam Duraiswamy - Re: INRA Member Page INTERNAL USE ONLY From:

"Mario V. Sonaca" <mvbonaca@snet.net>

To:

"Sherry Meador" <SAM@nrc.gov>, <TSKress@aol.com>, <apostola@mit.edu>

Date:

10/2/029:18AM

Subject:

Re: INRA Member I am available, Mario

--- Original Message From: "Sherry Meador" <SAM@nrc.gov>

To: <TSKress@aol.com>; <apostola@mit.edu>; <mvbonaca@snet.net>

Cc: "Jenny Gallo" <JMG@nrc.gov>; "John Larkins" <JTL@nrc.gov>

Sent. Tuesday, October 01, 2002 11 :49 AM

Subject:

INRA Member

> Greetings Gentlemen,

> Dr. Larkins would like for you to know that he has been in touch with Laurence Williams thru the Nil Chief Inspector's Technical Support Staff.

Mr Williams will be in the U.S. in December.

> Mr. Williams will be in the Washington, DC area on December 5.

He will be invited to meet with some of the ACRS member during this visit to discuss Advanced Reactors (e.g., AP1 000) - the UK has plans for two advanced plants which are under serious consideration.

If there are no objections, Dr.

Larkins will attempt to schedule an hour with Mr. Williams on December 5, 2002, at or around lunchtime. Mr. Williams' schedule has him leaving on a flight back to the UK at approximately 6:30 pm.

> Please let me know as soon as possible of your decision.

> Thank you

> Sherry cc:

"Jenny Gallo" <JMG@nrc.gov>, "John Larkins" <JTL@nrc.gov>

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, Sam Duraiswamy - Davis-Besse Page 1 INTERNAL PSE ONLY From:

<TSKress@aol.com>

To:

<wjshack@anl.gov>, <apostola@mit.edu>, <FPCTFord@aol.com>,

<GMLeitch@aol.com>, <JDSIEBER@aol.com>. <mvbonaca@snet.net>, <dapower@sandia.gov>,

<ransom@.ecn.purdue.edu>, <historyart@computron.net>, <graham.b.wallis@dartmouth.edu>,

<JTL@nrc.gov>, <sxd1@nrc.gov>, <rps1@nrc.gov>, <sxb@nrc.gov>

Date:

1012102 1:40PM

Subject:

Davis-Besse Gentlemen: r have been giving some thought to the Davis-Besse (D-B) incident Of course, there is a safety culture issue with D-B but I think the incident raises broader concerns than that. I think there is a "boric acid" issue. There have been numerous events associated with boric acid.

Consider, for example, the following issues:

1. Boric acid corrosion of the pressure vessel

2. Insufficient boric acid in eccs water (there have been several examples of this)

3. Wrong concentrations of boric acid

4. Too much boric acid causing precipitation. Freezing of valves from solidified crystals.

5. Corrosion enhancement in general

6. Core melt accidents.....If the water is lost, much of the neutron absorber is lost I Boron has the potential to creat more volatile forms out of Csl.

Why is the boron there? Simply for convenience and flexibility in controlling the reactivity as burnup proceeds during a fuel cycle. It is not really needed. BWRs do very nicely, thank you, without boric acid using burnable poisons (gadolinium) and PWRs could do just as well. We need boric acid like an armadillo needs a road to cross in Texas (apologies to Steve). Why don't we just get rid of it?

We need to urge NRC to take a much broader look at the implications of boric acid than just what happened at D-B.

Sam Duraiswamy - Re: Davis-Besse Page 1 INTERNAL USE ONLY From:

<JDSIEBER@aol.com>

To:

<TSKress@aol.com>, <wjshack@anl.gov>, <apostola@mit.edu>,

<FPCTFord@aol.com>, <GMLeitch@aol.com>, <mvbonaca@snet.net>, <dapower@sandia.gov>,

<ransom@.ecn.purdue.edu>. <historyart@computron.net>, <graham. b.wallis@dartmouth.edu>,

<JTL@nrc.gov>, <sxd1 @nrc.gov>, <rps1@nrc.gov>, <sxb@nrc.gov>

Date:

10/3/0211:14AM

Subject:

Re: Davis-Besse In a message dated 10/2/02 1:41 :08 PM Eastern Daylight Time, TSKress writes:

> "We need boric acid like an armadillo needs a road to cross in Texas

> (apologies to Steve). Why don't we just get rid of it?"

I don't want to be in the business of arguing all the time but I offer some thoughts about eliminating the use of boric acid in PWR coolant systems.

I agree with all of the problems that you listed in your message, except one.

During a core melt accident when the coolant is lost, I agree that the boric acid (a significant neutron absorber is also lost. However, the moderator is also lost, so the fission reaction stops for lack of moderator. As the core rearranges itself, Keff declines, sometimes to as low as 0.7, as in the case ofTM12.

On the other hand, if boric acid in the moderator was eliminated and gadolinium or other lumped burnable poisons were used instead, several interesting effects (mostly economic) would occur. First, I doubt that current control rods in PWRs are black enough to bring the reactor to hot shutdown under all circumstances. Therefore both the materials and configuration of control rods would have to change to accommodate a boron free RCS. Secondly, operation under a boron free RCS regime would require the operator to operate the plant with control rods partially inserted. This is the manner in which Navy plants operate. However, operation with a deep bite in the core causes the power density to concentrate in the active part of the core In Navy plants, this makes no difference because the power

.densities are low and there is plenty of margin. In commercial plants, the increased power density in the active part of the core may cause the core to exceed Appendix K limits. Otherwise, the plant power capability would have to be de-rated.

On the other hand, if the fuel designer placed the right amount of burnable poisons in the core to exactly track core burnup with all rods out, the plant would still have difficulty changing power levels due to Xenon buildup and burnout.

Admittedly, I have not thought too much about this idea, since it is a radical concept. But, if we consider the merits and demerits of such an idea, hopefully, my thoughts would be considered. I would bet that the commercial fuel designers would comment on this concept most vociferously.

Cheers, Jack Sieber