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License Renewal RAIs and Audit Questions (Audit Question No.: 2.1.1-01
	ML052580256
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Site:	Monticello
Issue date:	08/29/2005
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Monticello Nuclear Generating Plant License Renewal RAIs and Audit Questions Total Items: 18 SSR Audit Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.1.1 According to Section 2.1.1 of the License Renewal Application, NEI 95-10 Revision 3 was utilized for scoping and screening guidance. Revision 4 of the guideline appears to be referenced in some project implementing procedures. What elements of Revision 4 were used in the scoping and screening process?

As described in the Monticello LRA, Revision 4 guidance was used for overall structure and formatting of the LRA whereas Revision 3 was used for scoping and screening guidance. A comparison of NEI 95-10 Revision 3 and Revision 4 was performed and the major changes, by document section, are summarized below. Relative to Scoping and Screening, Revision 4 did not change Revision 3 guidance but supplemented it by including a new appendix (Appendix F) to provide further guidance on Scoping Criterion

2. Final Scoping decisions were not based on Appendix F. Appendix F was only used to structure the Monticello presentation of Scoping Criterion 2 results and to supplement other sources of information used by Monticello such as NRC requirements, the CLB and industry experience.

Specific to Appendix F, Monticello used the additional guidance as follows:

- The same Scoping Criterion 2 subgroupings were used for presenting Monticello results in the LRA. Subgroups were established for (1) CLB topics, (2) non safety related (NSR) SSCs directly connected to Scoping Criterion 1 SSCs, and (3) NSR SSCs not directly connected but whose failure due to spatial proximity could prevent satisfactory completion of a Scoping Criterion 1 function (LRA Sections 2.1.4.2.2.a, b, and c).

- For CLB topics, Appendix F was only used to help ensure all applicable CLB topics were considered, i.e., missiles, heavy loads, flooding, and high energy lines (LRA Section 2.1.4.2.2.a).

- For NSR piping systems directly connected to Scoping Criterion 1 piping systems, Monticello used an approach of "equivalent anchors" for defining Scoping Criterion 2 boundaries similar to Appendix F. However, the definition of equivalent anchor is largely based on the Monticello CLB as discussed in response to RAI 2.1.4.2.2-02 and is not based on Appendix F (LRA Section 2.1.4.2.2.b).

- For NSR SSC spatial proximity, industry experience contained in Appendix F was used to help support Monticello's position for including NSR piping hangers in scope for Scoping Criterion 2 and for not including the associated NSR air lines and HVAC ducting. For fluid containing systems, both the piping and supports were included in scope if in the same general area as Scoping Criterion 1 equipment. Line pressure and exposure duration considerations of Appendix F were not used by Monticello (LRA Section 2.1.4.2.2.c).

Summary of NEI 95-10, Revision 4 Changes by Section:

- Included additional guidance on standard component types and functions,

- Added discussion on Interim Staff Guidance,

- Added reference to NUREG-1800, 1801, and NRC inspection procedures,

- Expanded discussion on LRA update process,

- Added a discussion on appeals process, and

- Added Scoping Criterion 2 guidance in a new appendix.

Question:

Draft Response:

Sufficient per NRC Status:

Dave Sexton MNGP Owner:

Mechanical Discipline:

SSR Audit Source:

Audit Question No.: 2.1.1-01 Author:

Galletti Potential Submittal on Date Received:

6/22/2005 Potential LRA Update Required Page 1 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number More specifically, by section, changes included:

Table of Contents, Title page, Headers//Changed to be consistent with NEI template.

1.0 Introduction//No major changes.

1.1 Background//No major changes.

1.2 Purpose and Scope

//No major changes.

1.3 Applicability//No major changes.

1.4 Utilization of NUREG 1800, NUREG 1801, Regulatory Guide 1.188 and NRC Interim Staff Guidance Documents//Information about NUREG 1800, NUREG 1801, Regulatory Guide 1.188 and Interim Staff Guidance (ISG) process was added.

1.5 Resolution of Current Safety Issues//Made minor edits to bring into line with NUREG-1800 Appendix A.3.

1.6 Organization of the Guideline//Added reference to Section 7 and Appendix D.

2.0 Overview of Part 54//No major changes.

3.0 Identify The SSCs Within The Scope Of License Renewal and Their Intended Functions//Pointers to NUREG 1800 were added.

3.1 Systems, Structures, and Components Within the Scope of License Renewal//Pointers to NUREG 1800, ISGs and new Appendix F were added.

3.2 Intended Functions of SSCs Within the Scope of License Renewal//Proposed changes to NUREG-1800 for sub components and consumables were added.

3.3 Documenting The Scoping Process//Pointer to NRC Inspection Procedure 71002, License Renewal Inspection, was added. Also added discussion of boundary drawings to accompany LRA.

4.0 Integrated Plant Assessment//Discussion of aging effects identification and AMP identification was added as a brief introduction to follow-on subsections.

4.1 Identification of Structures and Components Subject to an Aging Management Review and Intended Functions//Standardized discussions of SSCs, SCs and intended functions. Evaluation boundary discussion was removed. Complex Assemblies discussion was added. Component intended function table was updated.

4.2 Identification of Aging Effects Requiring Aging Management//Significant revision to incorporate NUREG 1801. Lessons learned were incorporated. Discussion of consistency with NUREG 1801 was added. Use of SLRA Standard Notes was added.

4.3 Demonstrate That the Effects of Aging Are Managed//Significant revision to incorporate NUREG 1801. AMP elements table was added. Discussion of use of NUREG 1801 AMPs, non-NUREG 1801 AMPs and NRC approved non-NUREG 1801 AMPs were added.

4.4 Operating Experience Review//New section. Reference to NUREG 1800 Appendix A.1 was added.

4.5 Documenting the Integrated Plant Assessment//Moved from Section 4.4.

5.0 Time-Limited Aging Analyses Including Exemptions//No major changes.

Page 2 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number 5.1 Time-Limited Aging Analyses//Examples from and links to NUREG 1800 were added. Potential TLAA table was improved based on LRAs.

5.2 Exemptions//No major changes.

5.3 Documenting the Evaluation of the Time-Limited Aging Analyses and Exemptions//No major changes.

6.0 License Renewal Application Format and Content//No major changes.

6.1 General Information//Pointers to NUREG 1800 and NUREG 1801 were added.

6.2 Application Format and Content Guidance//Changed to point to Appendix D. Table updated to current Standard License Renewal Application (SLRA) format, discussion of ISGs added, and more detailed guidance on application format and content included.

6.3 Identify CLB Changes//Moved to 7.1.

7.0 Post License Renewal Application Submittal Activities//New section, introduction added.

7.1 Update of the License Renewal Application for CLB Changes//New section to describe the annual update process.

7.2 License Renewal Application Appeals//New section.

7.3 Post License Renewal Newly Identified SSCs//New section with 10 CFR 54.37(b) process added.

APPENDICES A 10 CFR Part 54 The License Renewal Rule//No changes.

B Typical Structure, Component and Commodity Groupings And Active/Passive Determinations For The Integrated Plant Assessment//Clarified groupings for fuse holders

(#77 & 83)Clarified groupings for housings (# 58, 116, 124, & 125).

C References//References 5 through 15 added.

D Standard License renewal Application Format//New section to include SLRA in a separate appendix.

E Interim Staff Guidance Documents//New section with a summary discussion of each ISG.

F Guidance on Revised 54.4(a)(2) Scoping Criteria//New section to include industry guidance on Scoping Criterion 2.

Final Response:

As described in the Monticello LRA, Revision 4 guidance was used for overall structure and formatting of the LRA whereas Revision 3 was used for scoping and screening guidance. A comparison of NEI 95-10 Revision 3 and Revision 4 was performed and the major changes, by document section, are summarized below. Relative to Scoping and Screening, Revision 4 did not change Revision 3 guidance but supplemented it by including a new appendix (Appendix F) to provide further guidance on Scoping Criterion

2. Final Scoping decisions were not based on Appendix F. Appendix F was only used to structure the Monticello presentation of Scoping Criterion 2 results and to supplement other sources of information used by Monticello such as NRC requirements, the CLB and industry experience.

Specific to Appendix F, Monticello used the additional guidance as follows:

- The same Scoping Criterion 2 subgroupings were used for presenting Monticello results in the LRA. Subgroups were established for (1) CLB topics, (2) non safety related Page 3 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number (NSR) SSCs directly connected to Scoping Criterion 1 SSCs, and (3) NSR SSCs not directly connected but whose failure due to spatial proximity could prevent satisfactory completion of a Scoping Criterion 1 function (LRA Sections 2.1.4.2.2.a, b, and c).

- For CLB topics, Appendix F was only used to help ensure all applicable CLB topics were considered, i.e., missiles, heavy loads, flooding, and high energy lines (LRA Section 2.1.4.2.2.a).

- For NSR piping systems directly connected to Scoping Criterion 1 piping systems, Monticello used an approach of "equivalent anchors" for defining Scoping Criterion 2 boundaries similar to Appendix F. However, the definition of equivalent anchor is largely based on the Monticello CLB as discussed in response to RAI 2.1.4.2.2-02 and is not based on Appendix F (LRA Section 2.1.4.2.2.b).

- For NSR SSC spatial proximity, industry experience contained in Appendix F was used to help support Monticello's position for including NSR piping hangers in scope for Scoping Criterion 2 and for not including the associated NSR air lines and HVAC ducting. For fluid containing systems, both the piping and supports were included in scope if in the same general area as Scoping Criterion 1 equipment. Line pressure and exposure duration considerations of Appendix F were not used by Monticello (LRA Section 2.1.4.2.2.c).

Summary of NEI 95-10, Revision 4 Changes by Section:

- Included additional guidance on standard component types and functions,

- Added discussion on Interim Staff Guidance,

- Added reference to NUREG-1800, 1801, and NRC inspection procedures,

- Expanded discussion on LRA update process,

- Added a discussion on appeals process, and

- Added Scoping Criterion 2 guidance in a new appendix.

More specifically, by section, changes included:

Table of Contents, Title page, Headers//Changed to be consistent with NEI template.

1.0 Introduction//No major changes.

1.1 Background//No major changes.

1.2 Purpose and Scope

//No major changes.

1.3 Applicability//No major changes.

1.4 Utilization of NUREG 1800, NUREG 1801, Regulatory Guide 1.188 and NRC Interim Staff Guidance Documents//Information about NUREG 1800, NUREG 1801, Regulatory Guide 1.188 and Interim Staff Guidance (ISG) process was added.

1.5 Resolution of Current Safety Issues//Made minor edits to bring into line with NUREG-1800 Appendix A.3.

1.6 Organization of the Guideline//Added reference to Section 7 and Appendix D.

2.0 Overview of Part 54//No major changes.

3.0 Identify The SSCs Within The Scope Of License Renewal and Their Intended Functions//Pointers to NUREG 1800 were added.

3.1 Systems, Structures, and Components Within the Scope of License Renewal//Pointers to NUREG 1800, ISGs and new Appendix F were added.

3.2 Intended Functions of SSCs Within the Scope of License Renewal//Proposed changes to NUREG-1800 for sub components and consumables were added.

Page 4 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number 3.3 Documenting The Scoping Process//Pointer to NRC Inspection Procedure 71002, License Renewal Inspection, was added. Also added discussion of boundary drawings to accompany LRA.

4.0 Integrated Plant Assessment//Discussion of aging effects identification and AMP identification was added as a brief introduction to follow-on subsections.

4.1 Identification of Structures and Components Subject to an Aging Management Review and Intended Functions//Standardized discussions of SSCs, SCs and intended functions. Evaluation boundary discussion was removed. Complex Assemblies discussion was added. Component intended function table was updated.

4.2 Identification of Aging Effects Requiring Aging Management//Significant revision to incorporate NUREG 1801. Lessons learned were incorporated. Discussion of consistency with NUREG 1801 was added. Use of SLRA Standard Notes was added.

4.3 Demonstrate That the Effects of Aging Are Managed//Significant revision to incorporate NUREG 1801. AMP elements table was added. Discussion of use of NUREG 1801 AMPs, non-NUREG 1801 AMPs and NRC approved non-NUREG 1801 AMPs were added.

4.4 Operating Experience Review//New section. Reference to NUREG 1800 Appendix A.1 was added.

4.5 Documenting the Integrated Plant Assessment//Moved from Section 4.4.

5.0 Time-Limited Aging Analyses Including Exemptions//No major changes.

5.1 Time-Limited Aging Analyses//Examples from and links to NUREG 1800 were added. Potential TLAA table was improved based on LRAs.

5.2 Exemptions//No major changes.

5.3 Documenting the Evaluation of the Time-Limited Aging Analyses and Exemptions//No major changes.

6.0 License Renewal Application Format and Content//No major changes.

6.1 General Information//Pointers to NUREG 1800 and NUREG 1801 were added.

6.2 Application Format and Content Guidance//Changed to point to Appendix D. Table updated to current Standard License Renewal Application (SLRA) format, discussion of ISGs added, and more detailed guidance on application format and content included.

6.3 Identify CLB Changes//Moved to 7.1.

7.0 Post License Renewal Application Submittal Activities//New section, introduction added.

7.1 Update of the License Renewal Application for CLB Changes//New section to describe the annual update process.

7.2 License Renewal Application Appeals//New section.

7.3 Post License Renewal Newly Identified SSCs//New section with 10 CFR 54.37(b) process added.

APPENDICES A 10 CFR Part 54 The License Renewal Rule//No changes.

Page 5 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number B Typical Structure, Component and Commodity Groupings And Active/Passive Determinations For The Integrated Plant Assessment//Clarified groupings for fuse holders

(#77 & 83)Clarified groupings for housings (# 58, 116, 124, & 125).

C References//References 5 through 15 added.

D Standard License renewal Application Format//New section to include SLRA in a separate appendix.

E Interim Staff Guidance Documents//New section with a summary discussion of each ISG.

F Guidance on Revised 54.4(a)(2) Scoping Criteria//New section to include industry guidance on Scoping Criterion 2.

Assoc LRA Section - 2.1.4 Are the flood protection plates outside warehouse #5 included in the plant tornado analysis as a possible missile?

Per USAR Section 12.2.1.8, the plant is designed for two tornado missiles, a 35 foot long utility pole, and a 1 ton automobile. The weight of a 4 x 8 x 1/2 plate is approximately 650 lbs. Thus the steel plates weigh less than the 1 ton design basis tornado missile.

In addition, in 2003, a field walkdown of the site was performed for potential tornado missiles. Almost 7000 missiles were considered from the zone that contains the steel plates. These missiles ranged in size from small motors to automobile size. Although the flood plates were not explicitly called out, they are considered included in the potential tornado missile population.

Question:

Draft Response:

Sufficient per NRC Status:

Scott Tradup MNGP Owner:

Civil Discipline:

SSR Audit Source:

Audit Question No.: 2.1.4.2.2-01 Author:

Galletti Potential Submittal on Date Received:

6/22/2005 Final Response:

Per USAR Section 12.2.1.8, the plant is designed for two tornado missiles, a 35 foot long utility pole, and a 1 ton automobile. The weight of a 4 x 8 x 1/2 plate is approximately 650 lbs. Thus the steel plates weigh less than the 1 ton design basis tornado missile.

In addition, in 2003, a field walkdown of the site was performed for potential tornado missiles. Almost 7000 missiles were considered from the zone that contains the steel plates. These missiles ranged in size from small motors to automobile size. Although the flood plates were not explicitly called out, they are considered included in the potential tornado missile population.

Potential LRA Update Required Page 6 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.1.4 Define Equivalent Anchor.

Issue of grouted penetrations serving as anchor converted to formal RAI. Transmitted in NRC letter dated 7/20/2005.

MNGP primarily used existing piping analyses to define equivalent anchors. In cases where there was no piping analysis, i.e. some small bore piping, the following were considered equivalent anchors:

A. A large piece of plant equipment B. A combination of restraints/supports such that a minimum of two levels of support are provided in each of the three orthogonal directions.

C. Grouted wall penetrations These equivalent anchors are consistent with equivalent anchors used in the piping analyses.

Question:

Draft Response:

Sufficient per NRC Status:

Scott Tradup MNGP Owner:

Civil Discipline:

SSR Audit Source:

Audit Question No.: 2.1.4.2.2-02 Author:

Galletti Potential Submittal on Date Received:

6/22/2005 Final Response:

This was a VERBAL question on 6-22-05.

MNGP primarily used existing piping analyses to define equivalent anchors. In cases where there was no piping analysis, i.e. some small bore piping, the following were considered equivalent anchors:

A. A large piece of plant equipment B. A combination of restraints/supports such that a minimum of two levels of support are provided in each of the three orthogonal directions.

C. Grouted wall penetrations These equivalent anchors are consistent with equivalent anchors used in the piping analyses.

Potential LRA Update Required Page 7 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.1.4 What is the basis for a wall/floor penetration to be an anchor?

The MNGP piping analysis specification defines an anchor as "engineered components designed to limit translation and rotation in three orthogonal directions". As long as the wall/floor penetration is grouted solid, it meets the criteria for an anchor. The grout used to fill the space between the pipe and the surrounding concrete is as strong or stronger than the concrete and provides a means to transfer the forces and moments to the surrounding concrete. In addition, the walls/floors are designed for piping reaction loads.

Wall/floor penetrations that are open or filled with fireproofing or other relatively soft materials are not used as anchors. The materials to fill these penetrations are designed to accommodate pipe movements, they are not designed to provide any restraint.

Question:

Draft Response:

Sufficient per NRC Status:

Scott Tradup MNGP Owner:

Civil Discipline:

SSR Audit Source:

Audit Question No.: 2.1.4.2.2-03 Author:

Galletti Potential Submittal on Date Received:

6/22/2005 Final Response:

This was a VERBAL question on 6-22-05.

The MNGP piping analysis specification defines an anchor as "engineered components designed to limit translation and rotation in three orthogonal directions". As long as the wall/floor penetration is grouted solid, it meets the criteria for an anchor. The grout used to fill the space between the pipe and the surrounding concrete is as strong or stronger than the concrete and provides a means to transfer the forces and moments to the surrounding concrete. In addition, the walls/floors are designed for piping reaction loads.

Wall/floor penetrations that are open or filled with fireproofing or other relatively soft materials are not used as anchors. The materials to fill these penetrations are designed to accommodate pipe movements, they are not designed to provide any restraint.

See response to RAI 2.1-02 for basis of wall/floor grouted penetrations as anchors.

Potential LRA Update Required Page 8 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.1.1 What is the definition of "General Area" if not as listed in LRA Section 2.1.4.2.2 and TR-011?

The LRA defines "General Area" as follows:

"General area is defined as being on the same floor of a building with no barrier walls between the fluid or steam source and the Scoping Criterion 1 component."

MNGP conservatively included in the above definition areas that "communicate" with the general area of interest. An example would be the main turbine deck at 951' and the condenser room at 911'. Since there are open hatchways between these two areas, NSR pressurized components in one area were deemed to affect SR equipment in the other area.

Question:

Draft Response:

Sufficient per NRC Status:

Joe Pairitz MNGP Owner:

Civil Discipline:

SSR Audit Source:

Audit Question No.: 2.1.4.2.2-04 Author:

Galletti Potential Submittal on Date Received:

6/22/2005 Final Response:

The LRA defines "General Area" as follows:

"General area is defined as being on the same floor of a building with no barrier walls between the fluid or steam source and the Scoping Criterion 1 component."

MNGP conservatively included in the above definition areas that "communicate" with the general area of interest. An example would be the main turbine deck at 951' and the condenser room at 911'. Since there are open hatchways between these two areas, NSR pressurized components in one area were deemed to affect SR equipment in the other area.

Potential LRA Update Required Page 9 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.2 Please provide a list of systems that were placed in License Renewal scope based solely on criterion (a)(2).

The following systems were placed in License Renewal scope based solely on criterion (a)(2):

Chemistry Sampling (CHM) LRA Section 2.3.3.2 Circulating Water (CWT) LRA Section 2.3.3.3 Fuel Pool Cooling and Cleanup (FPC) LRA Section 2.3.3.10 Turbine Generator (TGS) LRA Section 2.3.4.5 Question:

Draft Response:

Sufficient per NRC Status:

Joe Pairitz MNGP Owner:

Mechanical Discipline:

SSR Audit Source:

Audit Question No.: 2.2-01 Author:

Galletti Potential Submittal on Date Received:

6/22/2005 Final Response:

The following systems were placed in License Renewal scope based solely on criterion (a)(2):

Chemistry Sampling (CHM) LRA Section 2.3.3.2 Circulating Water (CWT) LRA Section 2.3.3.3 Fuel Pool Cooling and Cleanup (FPC) LRA Section 2.3.3.10 Turbine Generator (TGS) LRA Section 2.3.4.5 Potential LRA Update Required Assoc LRA Section - 2.1 The process for the treatment of consumables is contained in LRPP 2-1 and TR-007. However, there are no (or very few) dispositions of consumables in ALEX. Where are the results of the evaluations of consumables documented?

Section 5.0 of TR-007 states that the basis for the determinations of consumables will be presented in the respective AMR reports. Every mechanical and civil MNGP AMR report contains a section entitled "Short-Lived Components and Consumables." This is the location of the basis for the consumable determinations as directed by TR-007.

Question:

Draft Response:

Sufficient per NRC Status:

Joe Pairitz MNGP Owner:

Mechanical Discipline:

SSR Audit Source:

Audit Question No.: 2.1.5.3-01 Author:

Kavanagh Potential Submittal on Date Received:

6/23/2005 Final Response:

Section 5.0 of TR-007 states that the basis for the determinations of consumables will be presented in the respective AMR reports. Every mechanical and civil MNGP AMR report contains a section entitled "Short-Lived Components and Consumables." This is the location of the basis for the consumable determinations as directed by TR-007.

Potential LRA Update Required Page 10 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.0 Confirm if we used NEI-95-10 rev 4 and any of the exception the NRC has taken to this documents. Be prepared to discuss.

NEI 95-10, Rev 4, was used as supplementary guidance for the format of the LRA. NEI 95-10, Rev 3, was the version used, and referenced, for all other guidance with respect to preparation of the LRA.

Question:

Draft Response:

Sufficient per NRC Status:

Dave Musolf MNGP Owner:

Licensing Discipline:

SSR Audit Source:

Audit Question No.: 2.0-01 Author:

Tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

NEI 95-10, Rev 4, was used as supplementary guidance for the format of the LRA. NEI 95-10, Rev 3, was the version used, and referenced, for all other guidance with respect to preparation of the LRA.

See also response to item 2.1.1-01.

Potential LRA Update Required Assoc LRA Section - 2.1.3 Reviewers would like a list of the CHAMPS special concern codes and code descriptions that were used as a basis for input to determine components in-scope.

Codes provided to audit team.

Question:

Draft Response:

Sufficient per NRC Status:

Marv Engen MNGP Owner:

Mechanical Discipline:

SSR Audit Source:

Audit Question No.: 2.1.3-01 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

Codes provided to audit team.

Potential LRA Update Required Page 11 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.1.4.2.1 MNGP was alerted to be prepared to discuss the following items further during the scoping and screening audit:

The definition of Safety Related has used in the MNGP equipment database (CHAMPS) for coding equipment and as used in the LR rule, are these the same if they are not the same how did you disposition any differences. It would be desirable to have available a technical report or any other engineering basis concerning what are the MNGP design basis events per the MNGP CLB and how was this information used in LR scoping.

Dave Sexton has developed brief position papers. See final response section.

During the trip we did show the reviewers information in TR-005 (IPA) methodology report) listing the DBEs and discussion in the LRA concerning internal flooding, external flooding, HELB, etc. However the reviewers will also will need further clarifying information on how other CLB events outside of USAR chapter 14 were addressed for scoping (anticipated operational occurrences). NRC reviewers indicated Farley prepared a document similar to what they are looking for.

Question:

Draft Response:

Sufficient per NRC Status:

Dave Musolf MNGP Owner:

Licensing Discipline:

SSR Audit Source:

Audit Question No.: 2.1.4.2-01 Author:

Tingen Potential Submittal on Date Received:

Final Response:

Scoping Criterion 1 Safety Related SSCs:

The first of the three criteria in 10 CFR Part 54 was used by the Monticello Nuclear Generating Plant (MNGP) to determine if Systems, Structures, and Components (SSCs) fall within the scope of the rule:

Safety-related systems, structures, and components are relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49 (b)(1)) to ensure the following functions-(i) The integrity of the reactor coolant pressure boundary (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (iii) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the guidelines in §50.34(a)(1),

§50.67(b)(2), or § 100.11 of this chapter, as applicable.

This is also the NRC definition of safety-related in 10 CFR 50.2 and 10 CFR 50.49. In recent years this, or very similar wording, has been included in several places in NRC regulations and guidance documents.

MNGP Current License Basis (CLB) Definition of Safety-Related:

The design, construction, and licensing of the MNGP predates the above definition of safety-related.

The MNGP Q-List and Q-List Extension were used to code items as safety-related in the MNGP CHAMPS database. The MNGP CHAMPS database, in turn, served as one of the information sources used to identify SSCs meeting Scoping Criterion 1 Safety Related Systems and Structures.

SSC functions were identified using a number of information sources including the CLB. These functions were compared to Scoping Criterion 1 to identify those that should be considered in-scope for license renewal for MNGP Design Basis Events (DBEs), regardless of their current classification in CHAMPS or supporting Q-List information sources (i.e., CHAMPS was not the only information source used to make Criterion 1 determinations).

In addition to CHAMPS, the Monticello Color Coded P&IDs and other controlled drawings were used to identify components required to support system-level and structure-Potential LRA Update Required Page 12 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number level functions. These components were included in-scope for license renewal and generally matched information contained in CHAMPS. Where differences were noted they were documented and resolved. Some of these differences resulted in the issuance of an Action Request for further evaluation within the site Corrective Action Program. In addition to the License Renewal Database (LRDB), results of this effort are documented on License Renewal Boundary Drawings for mechanical components.

Scoping Criterion 1 Evaluation Method A systematic (holistic) approach was implemented to identify and confirm Scoping Criterion 1 systems, structures, and components (SSCs) were properly identified and captured in scope for License Renewal.

Preparation:

The following approach was used to ensure complete identification of SSCs satisfying Scoping Criterion 1:

Identify system and structure function(s). Major information sources included:

o SAR (with emphasis on Chapter 12 for Class I items, Chapter 14 for accident and transient event analyses, and individual system and structure description and design bases sections),

o Technical Specifications, o Color Coded (Q-List) P&IDs, o Classification information in the plant equipment database (CHAMPS),

o Design Bases Documents (unofficial source of functions and safety classification),

o Plant drawings, o Operations Manuals, o Maintenance Rule Basis Documents, and o Docketed correspondence.

Classify system and structure functions in accordance with 10CFR54.4 definitions. Information sources include those noted previously.

Identify component functions using standard function types. This effort further validated a complete list of system and structure functions was prepared (e.g., for containment isolation components, an associated function of containment isolation was also included at the system level). Information sources include those previously noted.

Document system, structure, and component level results (with references) in the License Renewal Data Base (LRDB) and System/Structure Scoping and Screening Output Reports.

Review:

In order to further ensure all Scoping Criterion 1 SSCs were properly identified and classified, an independent review by a technical expert was performed. This review used the same information sources available to the scoping report preparers and performed the following additional actions:

Compared scoping results across all systems and structures (over 100) for consistency,

Performed numerous key word searches on both the USAR (review of over 5,600 hits) and electronic docket (review of over 75 individual documents) to confirm all systems and structures had been appropriately identified and evaluated,

Reviewed select modifications, calculations, and obtained input from site subject matter experts,

Assisted in the development of some standard conventions for conveying scoping results onto License Renewal Boundary Drawings and reviewed all drawings, and

Documented results of this review in the form of comments, by signing as reviewer on Revision 0 System/Structure Scoping and Screening Output Reports and license renewal boundary drawings, and by issuing a review report.

The conclusion to this review noted no major issues were identified. A few additional system/structure level functions were added and, in most cases, the associated components were already in License Renewal scope to support other functions. The most significant changes were to include some components for Scoping Criterion 2, not Criterion 1, functions.

Page 13 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number MNGP Design Basis Events:

As noted in Section 2.1.3.2 of the MNGP License Renewal Application (LRA), DBEs include both design basis accidents and bounding transients evaluated in Chapter 14 of the USAR. Chapter 12 of the USAR includes an evaluation of natural phenomena and external events applicable to MNGP. Further, Chapter 12 identifies those systems and structures as Class I whose failure could cause a significant release of radioactivity or which are vital to safe shutdown of the plant under normal or accident conditions and the removal of decay and sensible heat from the reactor. Class I SSCs are included in scope for License Renewal per 10 CFR 54.4(a)(1).

DBEs are not defined in the MNGP USAR in the same way they are defined in 10 CFR 50.49(b)(ii). For the purposes of equipment qualification in accordance with 10 CFR 50.49, DBEs included those that could create a harsh environment for which the functions of 10 CFR 50.49 were required (i.e., loss of coolant accidents or high energy line breaks). For the purposes of License Renewal, MNGP used the broader definition of the term Design Basis Event from 10 CFR 50.49(b)(ii).

Therefore, safety related SSCs relied upon to perform the functions of 10 CFR 54.4(a)(1) for Design Basis Events are included in License Renewal scope. For MNGP this includes:

Normal operation (e.g., startup, shutdown, power operation, refueling),

Anticipated Operational Occurrences (limiting plant transients per USAR Section 14.4):

o Generator Load Rejection Without Bypass, o Loss of Feedwater Heating, o Feedwater Controller Failure (maximum demand),

o Rod Withdrawal Error, and o Turbine Trip without Bypass.

Design Basis Accidents (USAR Section 14.7 and Appendix I):

o Control Rod Drop, o Loss of Coolant Accident (LOCA),

o Main Steam Line Break (MSLB),

o Fuel Assembly Loading Accidents, o Recirculation Pump Seizure, o Refueling Accident, and o High Energy Line Breaks outside containment (in addition to the MSLB).

External Events, Natural Phenomena, and Internal Events (USAR Section 12.2.1):

o Live Loads (e.g., roof snow loads),

o Wind Loads, o Flooding (External and Internal),

o Tornadoes, and o Seismic Events.

Non-Limiting Plant Transients Contained in the CLB Page 14 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - None The definition of Safety Related has used in the MNGP equipment database (CHAMPS) for coding equipment and as used in the LR rule, are these the same if they are not the same how did you disposition any It would be desirable to have available a technical report or any other engineering basis concerning what are the MNGP design basis events per the MNGP CLB and how was this information used in LR scoping.

During the trip we did show the reviewers information in TR-005 (IPA methodology report) listing the DBEs and discussion in the LRA concerning internal flooding, external flooding, HELB, etc. However the reviewers will also will need further clarifying information on how other CLB events outside of USAR chapter 14 were addressed for scoping (anticipated operational occurrences). NRC reviewers indicated Farley prepared a document similar to what they are looking for.

See response to Audit Item 2.1.4.2-01.

Question:

Draft Response:

Sufficient per NRC Status:

Dave Musolf MNGP Owner:

Licensing Discipline:

SSR Audit Source:

Audit Question No.: 2.1.4-01 Author:

Tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

Addressed to satisfaction of NRC during Scoping and Screening Audit.

See response to Audit Item 2.1.4.2-01.

Potential LRA Update Required Assoc LRA Section - 2.1 When looking at section 5 (System Changes) Equipment moved to other systems for the SSR report for CRD, the NRC reviewer noted that some components were moved to "Miscellaneous Components not used". NRC reviewer questioned what this group of components was.

I explained that this was a place holder to move components out of the system for components evaluated in a separate area or that were duplicates entries, etc. The CRD component moved to "Miscellaneous Components not used" was a Control Panel (C-216), which we explained the control panels were evaluated by Civil via a commodity.

Need to be prepared to discuss with the audit team how the place holder systems (Miscellaneous Components not used, JUNK, etc. were used when we provide the audit team an overview of ALEX.

Question:

Draft Response:

Sufficient per NRC Status:

Marv Engen MNGP Owner:

Mechanical Discipline:

SSR Audit Source:

Audit Question No.: 2.1-03 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

I explained that this was a place holder to move components out of the system for components evaluated in a separate area or that were duplicates entries, etc. The CRD component moved to "Miscellaneous Components not used" was a Control Panel (C-216), which we explained the control panels were evaluated by Civil via a commodity.

Need to be prepared to discuss with the audit team how the place holder systems (Miscellaneous Components not used, JUNK, etc. were used when we provide the audit team an overview of ALEX.

Potential LRA Update Required Page 15 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.3.3.4 During comparison of CRD SSR to LRA, it was note that LR-36042 is listed in the SSR as an applicable boundary drawing for CRD, but this information is not included in LRA section 2.3.3.4 (CRD system, list of LR drawings).

Drawing will be added in first Annual USAR Supplement.

Question:

Draft Response:

Sufficient per NRC Status:

Dave Musolf MNGP Owner:

Mechanical Discipline:

SSR Audit Source:

Audit Question No.: 2.3.3.4-01 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

Drawing will be added in first Annual USAR Supplement.

Potential LRA Update Required Assoc LRA Section - 2.3 Reviewers questioned how Heat Tracing was treated with respect to license renewal. I could not immediately recall the location of this information in our TR reports. Need to be prepared to discuss this issue further. Also be prepared to discuss insulation further concerning scope. Reviewers also questioned treatment of vessel/drywell system insulation.

NRC audit team satisfied with this area following discussions with MNGP LR Mechanical Group members.

Question:

Draft Response:

Sufficient per NRC Status:

Joe Pairitz MNGP Owner:

Discipline:

SSR Audit Source:

Audit Question No.: 2.3-01 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

NRC audit team satisfied with this area following discussions with MNGP LR Mechanical Group members.

Potential LRA Update Required Page 16 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.4 Staged equipment to mitigate CLB events such as external flooding. Need to confirm and be prepared to discuss if such equipment is in scope. Reviewers indicated the equipment credited to mitigate CLB events should be in scope consistent with CLB commitments.

Converted to formal RAI. See NRC letter dated 7/20/2005.

See response to RAI 2.1-01 on staged equipment.

Question:

Draft Response:

Sufficient per NRC Status:

Madalin O'Brien MNGP Owner:

Civil Discipline:

SSR Audit Source:

Audit Question No.: 2.4.1-02 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

See response to RAI 2.1-01 on staged equipment.

Potential LRA Update Required Page 17 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.4.15 During overview of ALEX for reviewers it was noted that for system Reactor Building (RXB), component CSSTEEL-RXB-EXT, the safety field (Y or N) was not filled in ALEX.

While the component is identified as In LR Boundary in ALEX, should confirm that there are not other inappropriate null values in ALEX.

A review of ALEX was performed to determine if there was a null value for the safety field for other components in the database. The following components had a null safety field:

Various subcomponents for the CFW, DGN, HTV, HPC, CDR and RAD systems T-EXPANSION-LIQ - DGN System V-FU HTV System CSSTEEL-RXB-EXT - RXB System CRD_GTBASE - RIT System CSP_TSLEEVE - RIT System Except for the E-25 Subcomponents under the RAD system, all the above components are in scope even though the safety field has a null value.

Question:

Draft Response:

Sufficient per NRC Status:

Scott Tradup MNGP Owner:

Civil Discipline:

SSR Audit Source:

Audit Question No.: 2.4.15-01 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Final Response:

A review of ALEX was performed to determine if there was a null value for the safety field for other components in the database. The following components had a null safety field:

Various subcomponents for the CFW, DGN, HTV, HPC, CDR and RAD systems T-EXPANSION-LIQ - DGN System V-FU HTV System Potential LRA Update Required Page 18 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number Assoc LRA Section - 2.5.1 Reviewers asked if all of the Reg Guide 1.97 instrumentation was in scope. We indicated that that we recalled that this instrumentation would be in-scope but that instrumentation typically screens out as active, but would have to review further to confirm treatment. Will need to be prepared to discuss this issue further.

Regulatory Guide 1.97 Components Position Statement 10 CFR 54 defines the requirements of systems, structures, and components within the scope of License Renewal. This includes: 1) Safety-related systems, structures, and components; 2) Non safety-related systems structures and components affecting safety-related components; and 3) Systems, structures, and components relied on in safety analyses or plant evaluations associated with the identified regulated events. See Attachment 1.

10 CFR 54.4 defines safety-related using the criteria provided by 10 CFR 50.49(b)(1). Safety-related is defined by 10 CFR 50.49(b)(1)(i) subparagraphs (A), (B) and (C). See. These subparagraphs do not include Regulatory Guide 1.97 components. Regulatory Guide 1.97 (RG 1.97) components are covered in 10 CFR 50.49(b)(3) which are included under the important to safety paragraph but are not, by definition, safety-related and therefore not within the scope of License Renewal, as a result of being RG 1.97 components.

NSP Commitment Number M87028A required the Q-list extension to include items related to Regulatory Guide 1.97. These components are identified in the site database with a special concerns code. Regulatory Guide 1.97 Category 1 and 2 components were identified with a "Y" in the Safety Related column in the License Renewal database (ALEX). The Safety Related column in ALEX designates components that are Safety Related or have some other special concerns category. Most of these components were included within the scope of License Renewal for reasons other than RG 1.97.

Regulatory Guide 1.97 Category 1 components were included within the scope of License Renewal with the exception of retired components, spare components, penetrations handled as commodities, CGCS components and associated power supplies currently being removed from the plant, and the following five components:

Equipment Type Equipment ID System Equipment Name Location RG 1.97 Category INSTR PLR-7251A PCT DW PRESS-TOR LVL-DW RAD-ACCD/RNG ADMIN 1 INSTR PR-2994 PCT DRYWELL AND TORUS PRESSURE ADMIN 1 INSTR PY-7251B PCT PRIMARY CONTAINMENT WIDE RANGE PI ISOLATOR EFT 1 INSTR TI-4072A PCT DIV 1 TORUS TEMP ADMIN 1 INSTR TY-4072A PCT DIV 1 TORUS TEMP ADMIN 1 The above listed components are designated Regulatory Guide 1.97 Category 1. These components are color-coded "Orange" on the Q-list extension P&ID, which indicates a special concern (i.e. RG 1.97). These components are all located in a mild environment. None of these components are designated as Safety Related by the site Q-list.

None of these components are credited as SBO, EQ, Fire Protection, or ATWS. Therefore, these components are not required to be included within the scope of License Renewal. Additionally, these components are all instruments, which screen out per NEI 95 10 Appendix B criteria.

With the exception of two computer points, all Regulatory Guide Category 2 components were included within the scope of License Renewal. The Computer System was evaluated to not be within the scope of License Renewal.

Regulatory Guide Category 3 components are typically commercial grade components powered by normal power supplies. Category 3 components are not within the scope of License Renewal. However, some Category 3 components were included within the scope of License Renewal due to other regulated event criteria.

Question:

Draft Response:

Sufficient per NRC Status:

Ron Siepel MNGP Owner:

Electrical Discipline:

SSR Audit Source:

Audit Question No.: 2.5.1-1 Author:

tingen Potential Submittal on Date Received:

5/18/2005 Potential LRA Update Required Page 19 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number

§ 54.4 Scope.

(a) Plant systems, structures, and components within the scope of this part are--

(1) Safety-related systems, structures, and components which are those relied upon to remain functional during and following design-basis events (as defined in 10 CFR 50.49 (b)(1)) to ensure the following functions--

(i) The integrity of the reactor coolant pressure boundary; (ii) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in § 50.34(a)(1), § 50.67(b)(2), or § 100.11 of this chapter, as applicable.

(2) All nonsafety-related systems, structures, and components whose failure could prevent satisfactory accomplishment of any of the functions identified in paragraphs (a)(1)(i), (ii), or (iii) of this section.

(3) All systems, structures, and components relied on in safety analyses or plant evaluations to perform a function that demonstrates compliance with the Commission's regulations for fire protection (10 CFR 50.48), environmental qualification (10 CFR 50.49), pressurized thermal shock (10 CFR 50.61), anticipated transients without scram (10 CFR 50.62), and station blackout (10 CFR 50.63).

§ 50.49 Environmental qualification of electric equipment important to safety for nuclear power plants.

(a) Each holder of or an applicant for a license for a nuclear power plant, other than a nuclear power plant for which the certifications required under § 50.82(a)(1) have been submitted, shall establish a program for qualifying the electric equipment defined in paragraph (b) of this section.

(b) Electric equipment important to safety covered by this section is:

(1) Safety-related electric equipment. Note 3 (i) This equipment is that relied upon to remain functional during and following design basis events to ensure--

(A) The integrity of the reactor coolant pressure boundary; (B) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (C) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in § 50.34(a)(1), § 50.67(b)(2), or § 100.11 of this chapter, as applicable.

(ii) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions (b)(1)(i) (A) through (C) of this section.

(2) Nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs (b)(1) (i) (A) through (C) of paragraph (b)(1) of this section by the safety-related equipment.

(3) Certain post-accident monitoring equipment. Note 4 Note 3: Safety-related electric equipment is referred to as "Class 1E" equipment in IEEE 323-1974. Copies of this standard may be obtained from the Institute of Electrical and Electronics.

Note 4: Specific guidance concerning the types of variables to be monitored is provided in Revision 2 of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."

Final Response:

Regulatory Guide 1.97 Components Position Statement 10 CFR 54 defines the requirements of systems, structures, and components within the scope of License Renewal. This includes: 1) Safety-related systems, structures, and components; 2) Non safety-related systems structures and components affecting safety-related components; and 3) Systems, structures, and components relied on in safety analyses or plant evaluations associated with the identified regulated events. See Attachment 1.

10 CFR 54.4 defines safety-related using the criteria provided by 10 CFR 50.49(b)(1). Safety-related is defined by 10 CFR 50.49(b)(1)(i) subparagraphs (A), (B) and (C). See. These subparagraphs do not include Regulatory Guide 1.97 components. Regulatory Guide 1.97 (RG 1.97) components are covered in 10 CFR 50.49(b)(3) which are included under the important to safety paragraph but are not, by definition, safety-related and therefore not within the scope of License Renewal, as a result of being RG 1.97 components.

Page 20 of 23 Current as of 8/29/2005 4:05:45 PM

Monticello Nuclear Generating Plant License Renewal Audit Questions Sorted by Status, NRC Reviewer/Auditor, and RAI/Question Number NSP Commitment Number M87028A required the Q-list extension to include items related to Regulatory Guide 1.97. These components are identified in the site database with a special concerns code. Regulatory Guide 1.97 Category 1 and 2 components were identified with a "Y" in the Safety Related column in the License Renewal database (ALEX). The Safety Related column in ALEX designates components that are Safety Related or have some other special concerns category. Most of these components were included within the scope of License Renewal for reasons other than RG 1.97.