ML050940025

Request for Approval of a Common Quality Assurance Program for Constellation Generation Group, LLC

ML050940025
Person / Time
Site:	Calvert Cliffs, Nine Mile Point, Ginna
Issue date:	03/30/2005
From:	Heffley J Constellation Generation Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML050940025 (88)

Text

J. M. Heffley Constellation Generation Group Chief Nuclear Officer 1997 Annapolis Exchange Parkway Suite 310 Annapolis, MD 21401 410-897-5020 Constellation Energy Generation Group, LLC March 30, 2005 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos. I & 2; Docket Nos. 50-317 & 50-318 Nine Mile Point Nuclear Station Unit Nos. I & 2; Docket Nos. 50-220 & 50-410 R.E. Ginna Nuclear Power Plant Docket No. 50-244 Request for Approval of a Common Quality Assurance Program for Constellation Generation Group, LLC Constellation Generation Group, LLC (CGG) is submitting a common Quality Assurance Topical Report (QATR) for your review and approval in accordance with the provisions of 10 CFR 50.54(a). This revision is proposed to provide consistency in quality assurance (QA) program implementation within CGG.

Upon your approval and site-specific implementation, the CGG QATR (Attachment 1) will replace the current site-specific QA programs for the Calvert Cliffs Nuclear Power Plant (CCNPP), Nine Mile Point Nuclear Station (NMPNS), and R.E. Ginna Nuclear Power Plant (Ginna).

1.0 DISCUSSION The current three site-specific QA programs were initially developed during the design and construction phases for the units at each site and differ in a number of respects due to the various time frames in which they were developed.

Constellation Generation Group, LLC's QATR was written to standardize the nuclear fleet to a common QA program based on American Society of Mechanical Engineers (ASME)

NQA-1-1994, Part 1, Quality Assurance Program Requirements for Nuclear Facilities, and American National Standards Institute (ANSI) N18.7-1976/American Nuclear Society (ANS)-3.2, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, using the methodology of Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) 17.3, Quality Assurance Program Description. In order to facilitate development and review of the CGG QATR, the transition to a common QA program from site-specific QA programs utilizes previously approved alternatives or exceptions in NRC Safety Evaluation Reports.

Differences between QA requirements at each site were duly considered and resolved to arrive at a common and consistent set of QA controls. The QATR content was extracted from existing site-specific QA programs. Nine Mile Point Nuclear Station is committed to Regulatory Guide 1.28, Revision 3,

Document Control Desk March 29, 2005 Page 2 Quality Assurance Program Requirements (Design and Construction), which endorsed ASME NQA-1-1983 and 1983 Addenda. Since ASME NQA-1-1983 and ASME NQA-1-1994, Part I are very similar, NMPNS's QA program was used primarily to address ASME NQA-1-1994, Part I commitments.

Calvert Cliffs Nuclear Power Plant is committed to ANSI N18.7-1976/ANS-3.2 and their QA program was used primarily to address this standard.

R.E. Ginna Nuclear Power Plant's QA program was previously converted to SRP 17.3 and provided a baseline to address the SRP 17.3 content requirements.

Compared to site-specific QA programs, the level of detail in the CGG QATR is reduced. Site-specific implementation details are removed and generic organization titles and descriptions replace the detailed descriptions.

As required by 10 CFR 50.54(a)(4)(ii), Attachment 2 provides for each plant a comparison of that plant's existing program with the new QATR, identifies any changes considered to be reductions in commitment, and provides a basis for concluding the program, as changed, continues to meet 10 CFR Part 50, Appendix B requirements. Because of the extensive formatting differences involved, CGG considers that the combination of the QATR and the site comparison matrices included with this letter fulfill the requirement of 10 CFR 50.54(a)(4)(ii) to submit "all pages affected by that change."

Constellation Generation Group, LLC plans to implement the common nuclear fleet QATR within three months of NRC approval.

2.0 SRP 17.3, QUALITY ASSURANCE PROGRAM DESCRIPTION Constellation Generation Group, LLC's QATR addresses the acceptance criteria described in SRP 17.3 Section II, including applicable commitments to the references in Section VI. Constellation Generation Group, LLC's exceptions and alternatives to SRP 17.3 Sections II and VI are defined in CGG QATR Table A.7.2, CGG's Conformance to Regulations, Regulatory Guides, and Industry Standards.

Exceptions and alternatives are based on existing exceptions and alternatives or those previously approved in NRC Safety Evaluation Reports. Where applicable, site-specific exceptions or alternatives were combined into CGG exceptions or alternatives.

In some cases, site-specific exceptions or alternatives were applied to the other sites. These cases are documented in Section 4.0 of this letter.

3.0 NRC SAFETY EVALUATION REPORTS As part of the CGG QATR development process, quality assurance alternatives or exceptions approved by an NRC safety evaluation for other licensees were incorporated. The basis and applicability of these approvals are as follows:

3.1 Consolidation of Quality Assurance (QA) Programs into one Quality Assurance Program Manual for all Entergy Sites - Arkansas Nuclear One, Grand Gulf Nuclear Station, River Bend Station, and Waterford Steam Electric Station (TAC No. M97893), dated November 6, 1998.

• In the Safety Evaluation Report (SER) to Entergy Operations, Inc., the NRC concluded that the Quality Assurance Program Manual (QAPM) submitted by Entergy is acceptable in that it satisfies the review guidance in SRP 17.3 and the regulatory requirements in 10 CFR Part 50, Appendix B. Similar to Entergy's QAPM, the CGG QATR is based on the guidance in SRP 17.3. The level of detail in CGG's QATR meets or exceeds the guidance in SRP 17.3 and the level of detail approved by the NRC for Entergy.

• Similar to Entergy, the fire protection program and Independent Safety Evaluation Group (ISEG) commitments will be relocated from NMPNS's Quality Assurance Program Topical Report to the Updated Safety Analysis Report (USAR). This action will be completed prior to

Document Control Desk March 29, 2005 Page 3 implementing the CGG QATR. A statement is included in QATR Addendum A.7.1, Review Functions of the PORC, Qualified Reviewers, NSRB, and ISEG, that the ISEG function, as applicable, will be described in the unit's safety analysis report.

• The NRC notes in the Entergy SER that "Very few of the changes in the proposed QAPM are considered to be reductions in commitments since most are consistent with existing statements in the current descriptions for at least one of the plants." "...and, if the staff review of the QAPM determined that a QA element was consistent with only one of the QA program description commitments, it was found acceptable in the QAPM for all sites." Constellation Generation Group, LLC does not propose any new reductions in commitments. However, CGG does apply a few site-specific exceptions to the other sites that may be considered a reduction in commitment for that site. These exceptions may not have been formally approved in a previous NRC SER. This allows sites to maintain their existing exceptions and minimizes the impact for sites upgrading their programs to the later standards.

These changes are discussed in Section 4.0 of this document.

• Similar to Entergy's QAPM, many site-specific QA element details included in the existing QA program descriptions do not appear in the consolidated QATR. The basic commitments to implement these QA elements still remain by appropriate referencing of regulatory guides and the endorsed standards in CGG QATR Table A.7.2. Site-specific aspects, such as organization titles and arrangements, have been described in more general terms, with sufficient detail to ensure 10 CFR Part 50, Appendix B requirements are met.

In addition, some general requirements repeated throughout the existing descriptions have been combined into common statements that apply throughout the QATR. The result is a less redundant and more generic type of format.

• The CGG QATR organization description mirrors the organization description approved for Entergy. A written description is provided that identifies the functional groups and executives to whom they report.

The described organization arrangement assures that adequate independence between the performing and verifying activities is maintained.

• Quality assurance elements described in the QATR are applied to safety-related structures, systems, components and activities to an extent commensurate with their importance to safety.

Each site is retaining their current commitments for grading of QA requirements. Table A.7.2, under ANSI/ANS 51.1-1983, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, describes the site commitments for grading of QA requirements.

• The SER for Entergy supports several exceptions that are taken at one or more sites and will be applied across the CGG nuclear fleet. The exceptions are:

> "The person who holds a senior reactor operator license for the affected unit and approves a temporary change to a procedure is not required to be in charge of the shift." (CGG QATR Table A.7.2 under ANSI Ni 8.7-1976/ANS-3.2.)

> "Prospective lead auditors shall demonstrate their ability to effectively implement the audit process and lead an audit team. They shall have participated in at least one audit within the year preceding the individual's effective date of qualification.

Upon successful demonstration of the ability to effectively lead audits, licensee management may designate a prospective lead auditor as a lead auditor." (Each site has already adopted this exception.

Located in CGG QATR Table A.7.2 under ASME NQA-1-1994, Part 1.)

> The staff found that alternatives to tagging equipment with calibration information are acceptable provided that the specific method implemented by the license meet the requirements of ANSI N45.2.4-1972/1EEE 336-1971, Installation, Inspection, and Testing

Document Control Desk March 29, 2005 Page 4 Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations, Section 6.2.1 as committed to in the QATR. They also found that the requirements of 10 CFR Part 50, Appendix B continued to be met. (CGG's alternative to tagging equipment with calibration information is described in CGG QATR Table A.7.2 under ANSI N45.2.4-1972/IEEE 336-1971. This is a re-wording of the existing exemptions at CCNPP and NMPNS.)

3.2 Browns Ferry, Watts Bar, and Sequoyah Nuclear Plants - Approval of Nuclear Quality Assurance Plan (TAC Nos. MA8574, MA8575, MA8576, MA8577, MA8578, MA8579), dated July 25,2000

• The NRC SER approved Tennessee Valley Authority's (TVA) request to standardize the requirements for plant reviews and the Plant Operations Review Committees (PORC). The principle changes approved for TVA were the elimination of plant reviews that were unique to individual TVA nuclear sites, a reduction in the threshold of items requiring PORC review, and a change of PORC composition requirements.

Constellation Generation Group, LLC proposes to standardize the requirements for plant reviews and PORC in the QA program. Similar to TVA, each CGG site has qualified reviewer programs for reviews of procedures, procedure changes, and proposed changes to safety-related structures, systems, and components.

In addition, each CGG site has an on-site review committee that provides an advisory function to the Plant General Manager. The CGG QATR conforms to the guidance of ANSI N18.7-1976/ANS-3.2, in the manner described in the NRC SER to TVA.

The CGG QATR description of the review function differs slightly from TVA's since it is based on, and designed to accommodate, the existing CGG QA programs:

> The PORC is composed of members with expertise in the organization areas specified in the TVA submittal and NMPNS's current QA program. In lieu of specifying the Operations Manager as PORC Chairperson, the Plant General Managers will appoint the Chairperson.

This is consistent with each site's current practice.

> CGG requires that one PORC member hold, or have held, an Senior Reactor Operator (SRO)

License for the respective nuclear power plant, or similar unit. This is a relaxation of CCNPP's requirement for an SRO License from CCNPP I and 2 only. This is more stringent for Ginna that allows an SRO License or SRO Certification. Nine Mile Point Nuclear Station's QA program does not specify SRO participation. This is more stringent than TVA since they did not specify SRO participation on PORC in their submittal.

> The PORC review requirements in the CGG QATR include the existing requirements for Ginna and NMPNS. The additional PORC reviews that were unique to Calvert Cliffs QA program, and are not in the CGG QATR, are discussed in Attachment 2.

3.3 Clarification of Revision 1 to the FirstEnergy Nuclear Operating Company Quality Assurance Program Manual - Beaver Valley Power Station, Unit Nos. 1 and 2, Davis-Besse Nuclear Power Station, Unit 1, Perry Nuclear Power Plant, Unit 1 (TAC Nos. MB3009 and MB3010), dated December 19,2001

• FirstEnergy Nuclear Operating Company (FENOC) developed its consolidated QAPM following the guidance of SRP 17.3 and the consolidated QA program description previously approved for Entergy Operations, Inc., dated November 6, 1998.

Constellation Generation

Document Control Desk March 29, 2005 Page 5 Group, LLC's consolidated QATR also mirrors the FENOC submittal approved by the NRC.

The NRC approved additional alternatives and exceptions for FENOC that CGG is incorporating into its QATR.

• A 24-month minimum frequency for audits was approved for FENOC in this SER.

This exception was previously incorporated into each CGG nuclear sites' QA programs without having to be considered a reduction in commitment. This exception is now a CGG QATR exception.

(Described in CGG QATR Table A.7.2 under ANSI N18.7-1976/ANS-3.2 and Regulatory Guide 1.28, Revision 3)

• In the SER to FENOC, the NRC approved relocating certain Regulatory Guide commitments from the site-specific QA programs to other sections of the Updated Final Safety Analysis Reports (UFSARs).

Constellation Generation Group, LLC sites are also relocating commitments to specific Regulatory Guides and standards from their QA programs to their safety analysis reports. These are discussed for each site in Attachment 2.

• As approved for FENOC, CGG also proposes to include in its QATR previously approved clarifications, alternatives, and exceptions to Regulatory Guides and standards that are currently site-specific. Some of these alternatives or exceptions may have been approved without a safety evaluation. The site-specific alternative or exceptions CGG would like to apply to the nuclear fleet are detailed in Section 4.0 of this document.

• Constellation Generation Group, LLC has proposed to standardize based on ANSI N18.7-1976/

ANS-3.2 as endorsed by Regulatory Guide 1.33, Revision 2. Since Ginna is committed to an earlier Regulatory Guide and standard, committing to a later Regulatory Guide and standard endorsed by the NRC is not a reduction in commitment. Nine Mile Point Nuclear Station is currently committed to ANSI/ANS-3.2-1982. Constellation Generation Group, LLC is using the NRC's approval of FENOC's request for a licensee to return to the earlier standard as a basis for not considering NMPNS's adoption of the earlier standard as a reduction in commitment. Six exceptions taken in NMPNS's Quality Assurance Program Topical Report were deleted by adopting ANSI N I8.7-1976/ANS-3.2.

• The NRC approved a clarification for qualification of nondestructive examination personnel.

(This clarification was already in use at Ginna, and will now be applied to the CGG nuclear fleet. The clarification is described in CGG QATR Table A.7.2, under ASME NQA-1-1994, Part 1.)

3.4 Approval of Proposed Revision 70 of Quality Assurance Topical Report EGC-1A, Revision 70, in accordance with 10 CFR 50.54(a) Requirements for Exelon/Amergen Plants (TAC Nos. MB4893, MB4894, MB4895, MB4896, MB4897, M14898, MB4899, MB4900, MB4901, M134902, MB4903, MB4904, MB4905, MB4906, MB4907, MB4908, MB4909, MB4910, MB4911, MB1412, and MB4913), dated December 24,2002

• Constellation Generation Group, LLC is adopting ASME NQA-1-1994, Part I based on the NRC's approval for Exelon Generation Company, LLC. Consistent with the SER, CGG's QATR is based on the regulatory positions of Regulatory Guide 1.28, Revision 3 and Regulatory Guide 1.33, Revision 2.

Regulatory Guide 1.28, Revision 3 endorses ASME NQA-1-1983 and 1983 addenda as a method acceptable to the NRC staff for complying with the provisions of Appendix B with regard to establishing and implementing the requisite QA program for the design and construction of nuclear power plants. The NRC staff examined Exelon's side-by-side comparison of ASMENQA-1-1983 and ASME NQA-1-1994, Part 1, and concurred with Exelon in finding ASME NQA-1-1994, Part 1 an acceptable basis for their

Document Control Desk March 29, 2005 Page 6 operational QA program. Constellation Generation Group, LLC's conformance to these guides and standards is documented in CGG QATR Table A.7.2.

3.5 Revision 15 of Operational Quality Assurance Manual for Grand Gulf Nuclear Station (TAC No. M99401), dated November 18, 1997

• In a letter dated January 30, 1996, the Nuclear Energy Institute proposed that the NRC modify its requirement concerning the annual evaluation of suppliers of components and services to nuclear power plant licensees. By letter dated October 24, 1996, the NRC accepted the Nuclear Energy Institute proposal with certain modifications. The acceptable alternative for conducting annual evaluations of suppliers is:

"A documented ongoing evaluation of the supplier should be performed. Where applicable, this evaluation should take into account (1) review of supplier-furnished documents such as certificates of conformance, nonconformance notices, and corrective actions, (2) results of previous source verifications, audits, and receiving inspections, (3) operating experience of identical or similar products furnished by the same supplier, and (4) results of audits from other sources (e.g., customer, ASME, or NRC audits). The results of the evaluations should be reviewed and appropriate corrective action should be taken. Adverse findings resulting from these evaluations should be periodically reviewed in order to determine if, as a whole, they result in a significant condition adverse to quality and to provide input to support supplier audit activities conducted by the licensee or a third party auditing entity."

Consistent with the SER for Grand Gulf, CGG will perform documented ongoing evaluations of suppliers in accordance with the NRC accepted alternative as stated above. Constellation Generation Group, LLC uses the following available sources of information in performing supplier evaluations:

> notifications of supplier deficiencies from the sites via the corrective action process;

> notifications of supplier-related industry events requiring action by the procurement function;

> Nuclear Procurement Issues Committee's program for immediate notification of significant findings;

> the on-line Nuclear Procurement Issues Committee's database; and

> Nuclear Netvork information.

4.0 ADOPTION OF SITE-SPECIFIC EXEMPTIONS AND PRACTICES Consistent with consolidated QA program submittals by Entergy Operations, Inc. and FENOC, CGG proposes to apply the following site-specific exceptions across the CGG nuclear fleet:

• Constellation Generation Group, LLC conforms to ANSI N18.7-1976/ANS-3.2 Section 4.5, Audit Program, with the exception that a 90-day grace period may be applied to the 24-month frequency for performing internal audits. A grace period of 90-days may be applied to the performance of annual evaluations of inspection, examination and testing personnel qualifications defined in ASME NQA-I-1994 Supplement 2S-1 Section 2.6. A grace period of 90-days may be applied to the performance of annual lead auditor recertifications describe in ASME NQA-1-1994 Supplement 2S-3 Sections 4.1 and 6.3. A grace period of 90-days may be applied to the performance of triennial supplier audits described in Regulatory Guide 1.28,

Document Control Desk March 29, 2005 Page 7 Revision 3, Section 3.2. (These exceptions are implemented at Ginna and NMPNS. These exceptions are approved in an NRC SER to Ginna dated July 22, 1998.)

• In lieu of the master surveillance schedule requirements of ANSI N18.7-1976/ANS-3.2 Section 5.2.8, independent master schedules may exist for different programs such as inservice inspection, pump and valve testing, and Technical Specification surveillance testing. (This interpretation is from NMPNS's QA program.)

• Implementation of ANSI N18.7-1976/ANS-3.2 Section 5.2.9, Plant Security and Visitor Control, is addressed in the respective site security plans. (This clarification is from NMPNS's QA program.)

• Interpretation of ASME NQA-1-1994 Supplement 2S-3: Personnel who perform audits that are outside the scope of 10 CFR Part 50, Appendix B are not required to be qualified in accordance with ASME NQA-1-1994 Supplement 2S-3.

(This interpretation is from NMPNS's QA program.)

• In lieu of the requalification requirements of ASMENQA-1-1994 Supplement 2S-3 Section 4.2, CGG requalifies Lead Auditors on the basis of the satisfactory performance of one audit, as observed by a qualified Lead Auditor. (This exception is from CCNPP's QA program.)

• ASME NQA-1-1994 Supplement 7S-1 Paragraph 8.2.4 states "...post installation test requirements and acceptance documentation shall be mutually established by the Purchaser and Supplier." In exercising ultimate responsibility for its QA program, CGG establishes post-installation test requirements, giving due consideration to supplier recommendations.

(This exception is from NMPNS's QA program.)

• Regulatory Guide 1.37, Revision 0, Paragraph C.3 requires the water quality for final flushes...shall be at least equivalent to the quality of the operating system water. Constellation Generation Group, LLC is committed to the stated requirement except for the oxygen content.

(This exception is from NMPNS's QA program.)

• ANSI N45.2.1-1973 Section 3.1.2 gives the requirements for Class B cleanness. Upgraded piping systems and components constructed of carbon steel materials will meet Class B cleanness requirements, except for final flushing/cleaning, which may exhibit rust staining in accordance with Class C cleanness requirements.

(This exception is from NMPNS's QA program.)

Constellation Generation Group, LLC proposes to apply an interpretation of ASME NQA-1-1994 Supplement 4S-I: Section 2.3 requires that suppliers have a documented quality assurance program in accordance with ASME NQA-1-1994, Part 1. Since many existing suppliers of basic components to the CGG nuclear fleet maintain quality assurance programs based on ANSI N45.2 or ASME NQA (Various Years), it is alternatively acceptable to require that suppliers have a quality assurance program in accordance with ANSI N45.2 or ASME NQA-1.

Constellation Generation Group, LLC proposes to implement the biennial fire protection program audit frequency used at Ginna and CCNPP. In lieu of the 12-month, 24-month, and 36-month fire protection and loss prevention audits, CGG will perform a biennial audit of the facility fire protection program and implementing procedures that combines the scope of the three audits into one. The biennial requirement includes an inspection and audit of the fire protection and loss prevention program by non-licensee personnel. The personnel may be representatives of a fire insurance brokerage firm or other qualified individuals. The NRC approved the biennial audit frequency for Ginna in a 10 CFR 50.54 Quality

Document Control Desk March 29, 2005 Page 8 Assurance Program Change Review dated March 22, 1995. The biennial audit requirement is located in CGG QATR Section C.2.

5.0 SITE-SPECIFIC OATR IMPACT provides a detailed comparison matrix of each site's existing QA program with the CGG QATR. Besides the overall changes already discussed, additional site-specific details are summarized as follows.

5.1 Calvert Cliffs Nuclear Power Plant Calvert Cliffs Nuclear Power Plant is committed to Regulatory Guide 1.33, Revision 2, which endorses ANSI N18.7-1976/ANS-3.2. Calvert Cliffs Nuclear Power Plant is also committed to various daughter standards of ANSI N45.2. The areas of CCNPP's current QA program that are impacted by the fleet QA program are:

• The CGG QA program is based on ASME NQA-1-1994, Part 1, which replaces ANSI N45.2 and the programmatic daughter standards.

• The Independent Spent Fuel Storage Installation QA program description for CCNPP is retained.

5.2 Nine Mile Point Nuclear Station Nine Mile Point Nuclear Station is committed to ANSI/ANS-3.2-1982, and Regulatory Guide 1.28, Revision 3, which endorses ASME NQA-1-1983 and 1983 Addenda. The areas of NMPNS's current QA program that are impacted by the fleet QA program are:

• The CGG QA program is based on ASME NQA-1-1994, Part 1, and ANSI N18.7-1976/

ANS-3.2.

These replace NMPNS's commitments to ASME NQA-1-1983 and ANSI/

ANS-3.2-1982.

• By committing to ANSI N18.7-1976/ANS-3.2 from ANSI/ANS-3.2-1982, six exceptions to ANSIIANS-3.2-1982 were deleted. With the exceptions to ANSIIANS-3.2-1982, NMPNS was essentially committed to ANSI N18.7-1976/ANS-3.2. As discussed in Section 3.3, the NRC previously approved adopting the earlier standard for FENOC.

• The NMPNS Quality Assurance Program Topical Report is described in the Unit I UFSAR and the Unit 2 USAR. Consistent with 10 CFR 50.54(a) and the other two CGG nuclear sites, the Unit I UFSAR and Unit 2 USAR will be revised to reference the CGG QATR upon implementing the CGG QATR.

• The technical review and control activities section is replaced by the qualified reviewers section, which is based on CCNPP QA program.

• The Nuclear Safety Review Board (NSRB) is based on CCNPP QA program:

> Composition and qualification requirements are not as prescriptive.

> Nine Mile Point Nuclear Station references the Unit 2 USAR for qualification requirements.

The site-specific qualification requirements will be deleted.

> The NSRB will have at least seven members, an increase from NMPNS's current five member requirement.

> Use of subcommittees is addressed in the CGG QATR.

Document Control Desk March 29, 2005 Page 9

> Audit requirements are addressed in CGG QATR Section C Assessment.

5.3 R.E. Ginna Nuclear Power Plant R.E. Ginna Nuclear Power Plant is currently committed to Regulatory Guide 1.28, Revision 2, which endorses ANSI N45.2-1977, and its daughter standards; and Regulatory Guide 1.33, Revision 0, which endorses ANSI N 18.7-1972. The areas of Ginna's current QA program that are impacted by the fleet QA program are:

• The CGG QA program is based on ASME NQA-1-1994, Part 1, which replaces ANSI N45.2-1977 and the programmatic daughter standards. The program is also based on ANSI N18.7-1976/ANS-3.2, which replaces Ginna's commitment to ANSI N18.7-1972.

• The use of qualified procedure reviewers in lieu of PORC is more descriptive.

• The NSRB is based on CCNPP QA program. The CGG QATR is more descriptive than Ginna's QA program. The new qualification requirements are less restrictive.

• The supplementary glossary of definitions was deleted. Standard industry definitions will be used.

• Ginna currently has a five-year procedure review frequency.

American National Standards Institute N18.7-1976/ANS-3.2 requires a biennial procedure review frequency. Constellation Generation Group, LLC is adopting CCNPP's existing exception to ANSI N 18.7-1976/ANS-3.2 for biennial procedure reviews. Calvert Cliffs Nuclear Power Plant exception is based on NRC internal guidance provided by the Director of Reactor Inspection and Licensing Performance to the Regional Directors of Reactor Safety, dated December 21, 1992. An NRC SER to Millstone Nuclear Power Station, Units 2 and 3, RE: Change to Revision 21 of the NUQAP Topical Report (TAC Nos. MA8551B and MA8552B), dated May 8, 2000, approved the same exception. To mitigate the administrative burden for Ginna associated with the change from a five-year review frequency to the revised biennial requirement, Ginna plans on a two-year transition period. The two-year transition period is needed to mitigate the administrative burden for Ginna associated with changing the review frequencies and processes upon implementing the CGG QATR. This transition period will also allow equalizing the workload for the new review process.

6.0 CONCLUSION

Revision 0 of the QATR describes the quality program controls for the three CGG nuclear sites. In developing the QATR, plant specific elements have been suppressed in favor of a more generic approach.

Based on the NRC staff's review of other consolidated licensee QA programs, CGG concludes that the consolidated CGG QATR satisfies the requirements of 10 CFR Part 50, Appendix B and follows the guidance contained in SRP 17.3.

Document Control Desk March 29, 2005 Page 10 Should you have questions regarding the information in this submittal, please contact Mr. J. Traynor at (585) 771-3189.

Very truft yours, STATE OF MARYLAND

TO WIT:

COUNTY OF ANNE ARUNDEL:

1, John M. Heffley, begin duly sworn, state that I am Senior Vice President and Chief Nuclear Officer, Constellation Generation Group, LLC ("CGG"), and that I am duly authorized to execute and file this request on behalf of CGG. To the best of my knowledge and belief, the statements contained in this document are true and correct.

To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CGG employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of C64LVERT this 30 dayof MaRc IH 2005.

.-, anW.

HESS my Hand and Notarial Seal:

Wendy L. Hurtet NOTARY PUBLIC M C omiCsion Expires: Calvert County, Maryland My Commission Expires 01/01/06

-JMH/JT --

Ndtry Public Date Attachments:

(1) Quality Assurance Topical Report for Constellation Generation Group, LLC (2) QA Program Comparison Matrix cc:

Director, Project Directorate 1-1, NRC R. V. Guzman, NRC S. J. Collins, NRC D. M. Skay, NRC P. S. Tam, NRC Resident Inspector, NRC (Calvert Cliffs)

Resident Inspector, NRC (Ginna)

Resident Inspector, NRC (Nine Mile Point)

R. I. McLean, Maryland DNR J. P. Spath, NYSERDA

ATTACHMENT (1)

QUALITY ASSURANCE TOPICAL REPORT FOR CONSTELLATION GENERATION GROUP, LLC Constellation Generation Group March 29, 2005

Constellation Energy Generation Group, LLC QUALITY ASSURANCE TOPICAL REPORT Revision 0 for Calvert Cliffs Nuclear Power Plant Nine Mile Point Nuclear Station R.E. Ginna Nuclear Power Plant Docket Nos.

50-317, 50-318 50-220, 50-410 50-244 License Nos.

DPR-53, DPR-69 DPR-63, NPF-69 DPR-18 Vice President R.E. Ginna Nuclear Power Plant Date 3

'/5

Constellation Generation Group, LLC Corporate Statement of Quality Assurance Policy Constellation Generation Group, LLC is an advocate of quality performance in our daily activities. The quality assurance program described in procedures has been developed to assure that activities, as defined within the program scope, are being performed correctly and in conformance with applicable requirements. This program is designed to assure the safe operation of each Nuclear Station and to meet the requirements of Title 10, Code of Federal Regulations, Part 50 (IOCFR50), Appendix B, "Quality Assurance Criteria for Nuclear Power Plants."

The quality assurance program applies to all activities affecting the safety related functions of those structures, systems, and components that prevent or mitigate the consequences of or monitor the course of postulated accidents, events, or phenomena that could cause undue risk to the health and safety of the public. These activities include operating, maintaining, modifying, refueling, testing, and inspecting. In addition, this program applies to IOCFR50 concerns associated with:

• maintaining the high degree of integrity of primary and secondary barriers of systems or structures containing radioactive materials.

providing fire detection, suppression, and consequence mitigation items utilized both to protect the safety related structures, systems, and components and to assure safe operation in the event of postulated fire.

providing assurance that instrumentation and controls which monitor accidents, or provide a secondary role in accident monitoring, function correctly and accurately.

This program also applies to the shipping of licensed radioactive material under IOCFR71, except for design and fabrication of shipping casks. The quality assurance program has also established controls to ensure that the construction, operational, and decommissioning phases for the Independent Spent Fuel Storage Installation (ISFSI) are conducted at Calvert Cliffs Nuclear Power Plant (CCNPP) in compliance withlOCFR72.

The Manager, Quality and Performance Assessment is responsible for coordinating the formulation of the quality assurance program and for assuring the program's implementation. Nuclear organization personnel are responsible for implementing the quality assurance program in accordance with the requirements of their procedures.

WicaJI J. Wallace PresWnt, Constellation Generation Group, LLC Date:

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 TABLE OF CONTENTS Section Title Page A

MANAGEMENT A. I Methodology 4

A.2 Organization 5

A.3 Responsibility 7

A.4 Authority 7

A.5 Personnel Training and Qualification 8

A.6 Corrective Action 8

A.7 Regulatory Commitments 8

B PERFORMANCE VERIFICATION B.1 Methodology 9

B.2 Design Control 9

B.3 Design Verification 10 B.4 Procurement Control 12 B.5 Procurement Verification 14 B.6 Identification and Control of Items 15 B.7 Handling, Storage, and Shipping 15 B.8 Test Control 16 B.9 Measuring and Test Equipment Control 17 B.10 Inspection, Test, and Operating Status 17 B.1 I Special Process Control 18 B.12 Inspection 19 B.13 Corrective Action 20 B.14 Document Control 21 B.15 Records 22 C

ASSESSMENT C.1 Methodology 23 C.2 Assessment 23 ADDENDUM A.7.1 Review Functions of the PORC, Qualified Reviewers, NSRB, and ISEG.

25 ATTACHMENT TABLE CGG's Conformance to Regulations, Regulatory Guides, and Industry A.7.2 Standards.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 A.

MANAGEMENT A.1 METHODOLOGY The President of Constellation Generation Group, LLC (CGG), establishes the CGG quality assurance policy. This policy is set forth in the CGG Corporate Statement of Quality Assurance Policy and is binding on all organizations and individuals performing CGG quality affecting activities. The policy is implemented under the overall direction of the Senior Vice President and Chief Nuclear Officer, CGG.

The QA program applies to all activities associated with structures, systems, and components that are safety-related.

Quality affecting activities are controlled to an extent consistent with their safety significance. The applicability of the requirements of the QA program to other items and activities is determined on a case-by-case basis. (Additional details are provided under Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Powver Plants, in Table A.7.2)

In addition, the QA program applies to the activities associated with the packaging of licensed radioactive materials to be shipped in accordance with I OCFR7 1, excluding design and fabrication of shipping casks.

QA program elements that apply to radioactive waste handling activities include:

1.

Quality and Performance Assessment will provide oversight of the radioactive waste handling activities through audits and inspections.

2.

Quality Inspectors will be trained in Department of Transportation (DOT) and NRC radioactive waste handling requirements.

The QA program has also established controls to ensure that the construction, operational, and decommissioning phases for the Independent Spent Fuel Storage Installation (ISFSI) are conducted at Calvert Cliffs Nuclear Power Plant (CCNPP) in compliance with 10CFR72. ISFSI records pertaining to the design, fabrication, erection, testing, audits, maintenance, and use of structures, systems, and components important to safety are maintained for the duration of the ISFSI license. All other activities associated with the operational and decommissioning phase shall be controlled under the CGG I0CFR50 Appendix B QA program. Existing policies, programs, directives, and procedures stated as applicable for CCNPP are also applicable for the ISFSI.

Nuclear directives address actions identified in the QA program. Nuclear directives identify regulatory commitments, management requirements, and assign responsibilities for business activities (e.g., design, maintenance, operations) within the CGG nuclear program. Lower tier documents are developed to implement the requirements addressed in the nuclear directives, and are consistent with the QA program.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 A.2 ORGANIZATION The organization structure responsible for implementing the quality assurance program is described below. The specific organization titles for the quality assurance functions described are identified in procedures. The authority to accomplish the quality assurance functions described is delegated to the incumbent's staff as necessary to fulfill the identified responsibility.

1.

The President, CGG is responsible for overall corporate policy and provides executive direction and guidance for the corporation as well as promulgates corporate policy through the Company's senior management staff. Overall responsibility for the implementation of the quality assurance program is delegated to the Senior Vice President and Chief Nuclear Officer, CGG.

2.

The Senior Vice President and Chief Nuclear Officer, CGG (CNO) reports to the President, CGG and has overall responsibility for the safe and reliable operation of the Company's nuclear stations including management oversight and support of the day-to-day operations of the stations.

This is the senior executive responsible for setting and implementing policies, objectives, expectations, and priorities to ensure activities are performed in accordance with the quality assurance program and other requirements. The CNO oversees activities of the Nuclear Safety Review Board.

3.

The Vice President for each nuclear site reports to the CNO and is responsible for overall plant nuclear safety and implementation of the Company's quality assurance program. This position is responsible for the station's compliance with its NRC Operating License, governmental regulations, and ASME Code requirements.

This position provides day-to-day direction and management oversight of activities associated with the safe and reliable operations of a nuclear station.

4.

The management position responsible for Quality and Performance Assessment (Q&PA) activities reports to the CNO, is independent of production, and assures that an appropriate quality assurance program is established, maintained, and effectively executed throughout the nuclear organization.

This position provides overall direction for the implementation of the quality assurance program and for the effective implementation of quality assurance functions that verify activities affecting safety-related functions. A staff of supervisory, administrative, and technical personnel supports assessment and quality verification.

5.

The individuals fulfilling the following management functions report to the positions identified above.

These individuals may report through an additional layer of management but shall maintain sufficient authority and organizational freedom to implement the assigned responsibilities. These individuals may be responsible for a single unit/location or for multiple units/locations and may fulfill more than one function described below. Conversely, more than one individual may fulfill these responsibilities.

a.

The Plant General Manager assures the safe, reliable, and efficient operation of the plant within the constraints of applicable regulatory requirements and the operating license.

The on-site safety review committee reports to the Plant General Manager.

b.

The individual responsible for plant modifications provides direction, control, and overall supervision of the implementation of plant modifications and assigned maintenance.

c.

The individual responsible for training provides direction, control, and overall supervision of all training of personnel required by regulations.

d.

The individual responsible for records management provides direction, control, and overall supervision of the records management program and associated activities.

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QUALITY ASSURANCE TOPICAL REPORT I

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e.

The individual responsible for document control provides direction, control, and overall supervision of the document control program and associated activities.

f.

The individual responsible for the corrective action program provides direction, control, and overall supervision of the corrective action program and associated activities.

g.

The individual responsible for engineering is responsible for the development and maintenance of engineering programs, policies, and procedures and for providing engineering services.

h.

The individual responsible for materials, purchasing, and contracts is responsible for procurement, services, receipt, storage, and issue of materials, parts, and components.

i.

The individual responsible for quality control has the responsibility for establishing, controlling, and implementing the quality control inspection program. The individual responsible for quality control has the authority and responsibility to escalate matters as needed.

Review of Operations Procedures require that CGG's nuclear power plants shall be operated and maintained in accordance with their Technical Specifications and operating licenses.

The following organizations review plant operations to ensure that these procedures are followed:

1.

The Quality and Performance Assessment organization provides independent verification that the requirements contained in the plants' operating licenses, safety analysis reports, Technical Specifications, and procedures are met. This is accomplished through Quality and Performance Assessment audits and assessments.

2.

The Nuclear Safety Review Board (NSRB) provides independent verification by review that CGG nuclear plants are operated in accordance with established requirements.

NSRB membership and functions are governed by Addendum A.7.1 and written procedures.

Proceedings of all meetings are documented and sent to the CNO, Committee members, and others designated by the Committee Chairperson.

3.

The on-site Plant Operations Review Committee (PORC) reviews matters pertaining to nuclear plant safety. This Committee screens subjects of potential concern to the NSRB and performs preliminary investigations under the direction of the Plant General Manager. PORC membership and functions are governed by Addendum A.7.1 and written procedures. Results of all meetings are documented and sent to the Chairperson of the NSRB, and others designated by the PORC Chairperson.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 A.3 RESPONSIBILITY All employees of CGG involved in the operation of the fleet nuclear power plants and associated support activities have full personal and corporate responsibility to assure that the plant is operated, maintained, tested, inspected, and modified in a safe and reliable manner. This responsibility includes assuring that an effective quality assurance program is implemented. Although authority for development and execution of specified parts of the program may be delegated to others (e.g. suppliers), CGG retains overall responsibility.

The QA program status, scope, adequacy and compliance with I0CFR50 Appendix B are regularly reviewed by CGG management through reports, meetings, review of audit results, and documented assessments performed by management teams. The NSRB reviews the status and adequacy of the QA program at least once every two years to assure that it is meaningful and effectively complies with corporate policy and 10CFR50, Appendix B. This review consists of an audit, or a review equivalent to an audit, performed by company personnel or outside organizations.

CGG is responsible for ensuring that the applicable portions of the QA program are properly documented, approved, and implemented (people are trained and resources are available) before an activity with the scope of the QA program is undertaken by CGG or by others. Individual managers ensure that personnel working under their cognizance are provided the necessary training and resources to accomplish their assigned tasks. The QA program is implemented through procedures prepared and maintained by the responsible organization and approved for use by their designated manager. Quality affecting activities are performed in accordance with these procedures, utilizing sufficiently trained personnel and necessary resources to accomplish the work.

Adherence to procedures is vital to the safe and reliable operation of CGG's Nuclear Power Plants.

Personnel are responsible for adhering to established procedures, interpreting them conservatively in case of doubt, and recommending changes when necessary. Procedures with the potential to affect nuclear or personnel safety shall be strictly adhered to. When an activity controlled by such procedures cannot be accomplished as described or accomplishment of such activity would result in an undesirable situation, the work shiall be stopped and the plant placed in a safe condition. Work shall not resume until the procedure is changed to reflect correct work practices.

Procedures may be deviated from during emergencies to prevent or minimize injury to personnel or damage to plant equipment. Any such deviations should be thoroughly documented.

A.4 AUTHORITY Persons or organizations who are delegated responsibility for planning, establishing, or implementing any part of the CGG quality assurance program also have the authority to carry out those responsibilities.

Nuclear operations and support organization personnel are empowered to take stop work action on their own activities if they determine that continuing the activity would preclude identifying and correcting a condition adverse to quality or lead to an unsafe condition. Designated independent inspection and audit personnel have the authority to stop work within nuclear operations and support organizations, and at supplier locations. The Plant General Managers have stop work authority for all activities performed in operating their respective stations.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 A.5 PERSONNEL TRAINING AND QUALIFICATION To ensure that important activities are performed correctly, CGG conducts formal training programs for Company personnel with significant responsibilities. These programs include both initial and continuing training, and are conducted in accordance with written procedures or instructions.

Department Managers/Directors are responsible for ensuring that the training needs of personnel in their Departments are identified, formal training programs to satisfy those needs are developed, and the training programs are implemented in accordance with the requirements of the QA program documents. Training and qualification records are maintained for each employee when required.

A.6 CORRECTIVE ACTION CGG has established a corrective action process whereby all personnel are responsible for assuring that conditions adverse to quality are promptly identified, reported, controlled, and corrected. The process is focused on correcting the problem and its root cause rather than assigning blame or fault. Adverse trends in performance are identified, monitored, and reported to management.

Corrective action and nonconformance control processes are discussed in Section B.13.

The Employee Concerns Program provides CGG and contractor employees an opportunity to communicate their quality concerns regarding operation, maintenance or modification while keeping their identity confidential, if they desire, and to receive feedback regarding the results of investigations with respect to their concerns. Quality concerns determined to be valid are acted upon by the responsible organization, and the actions are verified prior to closeout.

A.7 REGULATORY COMMITMENTS The CGG QA program was developed to meet the requirements of Regulations and Regulatory Guides of the Nuclear Regulatory Commission (NRC), and Industry Standards listed in Table A.7.2. Exceptions taken to the guidance contained in these documents, and equivalent CGG alternatives, are stated in Table A.7.2.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.

PERFORMANCE VERIFICATION B.]

METHODOLOGY The QA program provides for safety-related activities to be performed under suitably controlled conditions, including the use of appropriate equipment, maintenance of proper environmental conditions, assignment of qualified personnel, and assurance that applicable prerequisites have been met.

Personnel performing work activities, such as design, engineering, procurement, manufacturing, construction, installation, startup, maintenance, modification, operation, and decommissioning are responsible for achieving acceptable quality. Employees are empowered to make decisions in their areas of responsibility and are held accountable for the quality of their own work.

Personnel performing verification activities are responsible for verifying the achievement of acceptable quality and are different personnel than those who performed the work.

Verification of work is performed by each organization to assure that quality objectives are achieved and established performance standards are met.

Nuclear Operations and supporting organizations involved in CGG quality related activities accomplish and verify their work using instructions and procedures. For quality affecting activities within their area of responsibility, each organization develops, reviews, approves, and implements such documents.

Persons preparing, reviewing, and approving instructions and procedures are responsible for assuring that they include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. In addition, they assure that these documents are kept current.

B.2 DESIGN CONTROL Station modifications are accomplished in accordance with approved designs and procedures.

The controls apply to preparation, review and revision of design documents, including the correct translation of applicable regulatory requirements and design bases into design, procurement and procedural documents.

Configuration changes, including temporary changes, are implemented utilizing design control measures at least commensurate with those applied to the original design. Changes to design output documents, including field changes, are controlled in a manner commensurate with that used for the original design.

Information on approved changes is transmitted to affected organizations.

Engineering has overall control of design documents. Design output documents, and revisions thereto, are controlled by the design office (architect-engineer, supplier, contractor, consultant or engineering) responsible for the design work. Each design organization controls design documents in accordance with approved procedures that provide for development, review, approval, distribution, document control and revision.

Design control measures are defined and implemented by trained and qualified personnel through approved procedures and instructions. These procedures and instructions assure that:

1.

Design inputs are appropriately specified on a timely basis and correctly translated into design documents.

2.

Design interfaces are identified and controlled.

3.

The design is suitable for its intended application.

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4.

Personnel other than those who performed the design verify design adequacy.

5.

Design changes, including field changes, are governed by control measures.

6.

Deviations and nonconformances are controlled.

7.

Design records are identified, controlled, and retrievable.

Design inputs (such as design bases, performance requirements, regulatory requirements, codes, and standards) and changes to design inputs are identified, documented, reviewed and approved, and controlled.

Design documents, including drawings and specifications, are prepared and technically reviewed by qualified personnel. The technical reviewer ensures that the design document is in accordance with the design concept, incorporates appropriate design inputs, and conforms to approved procedures and instructions. Appropriate management or supervision approves design change packages prior to release for implementation.

Design interfaces, both internal and external, are formally identified, and design activities are coordinated among the participating organizations to ensure that design inputs and outputs are properly developed, reviewed, approved, and distributed. Multi-discipline changes are reviewed to ensure integration of design outputs.

Computer programs used in safety-related design and Station operation are appropriately documented, verified, certified for use and controlled in accordance with procedures.

Design databases, documents, and procedures are revised to reflect changes installed in the plant. Design records are identified, indexed, and controlled to allow for retrievability of design basis information and to provide evidence of appropriate design controls.

B.3 DESIGN VERIFICATION Design verification is the process of reviewing, confirming, or substantiating the design to assure:

1.

the acceptability of the design inputs

2.

adherence to the design process

3.

that design inputs are reflected in the design outputs

4.

that design changes are implemented under controls commensurate with those applied to the initial design The extent of and methods used for design verification are documented. Methods for design verification include evaluation of the applicability of standardized or previously proven designs, alternate calculations, qualification testing and design reviews.

These methods may be used singly or in combination, depending on the needs for the design under consideration.

When design verification is done by evaluating standardized or previously proven designs, the applicability of such designs is confirmed. Any differences from the proven design are documented and evaluated for the intended application.

Design reviews are performed by individuals, or by interdisciplinary or multi-organizational groups, as appropriate. Unless otherwise stated, the verification of design addresses the information conveyed by Page 10 of 30

QUALITY ASSURANCE TOPICAL REPORT Revision 0 the design document. When the verification is limited to certain areas or features, the scope or extent and any limitations on the verification are documented.

Qualification testing of prototypes, components, or features is used when the ability of an item to perform an essential safety function cannot otherwise be adequately substantiated. This testing is performed before Station equipment installation where possible, but always before reliance upon the item to perform a safety-related function. Qualification testing is performed, under conditions that simulate the most adverse design conditions as determined by analysis, considering relevant operating modes.

Test requirements, procedures and results are documented.

Results are evaluated to assure that test requirements have been satisfied. Modifications are made if shown to be necessary through testing.

Following modification, any necessary retesting or other verification is performed. Scaling laws are established and verified when applicable. Test configurations are documented.

Persons representing applicable technical disciplines are assigned to perform design verifications. These persons are qualified by appropriate education or experience and are not directly responsible for the design being verified. The originator's supervisor may perform this verification, provided that:

1.

The supervisor did not specify a singular design approach or rule out certain design considerations, and did not establish the design inputs used in the design; or

2.

The supervisor is the only individual in the organization competent to perform the verification, and receives written approval by the appropriate engineering manager.

When designs must be released for use before they have been completed or before they have been verified, the incomplete or unverified parts of the design and the hold point to which work may proceed are identified, and design output documents based on unverified data are identified and controlled. This hold point occurs before the work becomes irreversible or before the item is relied on to perform a safety-related function. Justification for such early release is documented.

Procedures define acceptable verification methods and controls, design parameters subject to verification, acceptance criteria, and verification documentation and records requirements.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.4 PROCUREMENT CONTROL Procurement documents define the characteristics of items or services to be procured, identify applicable regulatory and industry codes or standards requirements, and specify supplier QA program requirements to the extent necessary to assure adequate quality.

CGG qualifies suppliers (including selected suppliers of commercial grade items) by performing a documented evaluation of their capability to provide items or services specified by procurement documents. Other means of evaluating supplier qualification include objective evidence of supplier's current quality performance and audits or surveys performed by consultants, other nuclear utilities, or other organizations such as Nuclear Procurement Issues Committee (NUPIC).

Alternately, items or services may be procured from suppliers and accepted based on appropriate inspection or verification activity.

Procurement of safety-related items and services from suppliers is permitted when CGG has performed a documented evaluation of their capability to provide the items or services specified by procurement documents. However, such evaluation for items or services specified may not necessarily result in the supplier being placed on the Approved Vendors List (AVL). In instances where an identified supplier is the only practical source, procurement may be authorized subject to satisfactory surveillance of the processes and characteristics identified in items (L.) and (2.) below:

1.

The supplier is capable of meeting specific procurement document requirements by virtue of their ability to control critical manufacturing and functional processes and characteristics identified by engineering; and

2.

methods have been identified and documented which will verify conformance to these requirements.

When required by operational considerations, an order may be placed with a supplier prior to completion of the evaluation and approval process only after obtaining approval from the nuclear site individual responsible for the procurement function.

CGG's acceptance of basic component items or services provided by an unapproved supplier is contingent on the subsequent Q&PA evaluation and approval of the supplier as stated above.

CGG qualified suppliers are evaluated on an ongoing basis to ensure that they continue to provide acceptable products and services. Qualified suppliers involved in active procurement are audited every 3 years (with a tolerance of 90-days) to maintain their qualified status except as follows. Supplier 3-year audits are not necessary to maintain qualification when the items or services supplied are determined and documented by procurement engineering and Q&PA to satisfy each of the following conditions:

1.

Relatively simple and standard in design, manufacture and test.

2.

Adaptable to standard or automated inspections or tests of the end product to verify quality characteristics after delivery.

3.

Receiving inspection does not require operations that could adversely affect the integrity, function, or cleanness of the item.

The National Institute of Standards and Technology (NIST), by virtue of its being the nationally recognized standard, is an acceptable provider of calibration masters, standards or services. Utilities holding an NRC Construction Permit or Operating License are acceptable suppliers of items and services.

Neither of the above are required to be listed on the Approved Vendors List (AVL).

Procurement planning by procuring organizations consists of determining the supplier of choice, methods to be used for acceptance of the item or service, and provisions for ensuring that qualified suppliers continue to provide acceptable products and services. Source inspection (surveillance), certificate of Page 12 of 30

QUALITY ASSURANCE TOPICAL REPORT Revision 0 conformance, receipt inspection, and pre-or post-installation testing are methods that are considered for item acceptance. The extent of the acceptance methods and associated verification activities will vary depending upon the relative importance and complexity of the purchased item or service and the supplier's past performance.

The contents of procurement documents vary according to the item and/or service being purchased and its function in the plant. Procurement documents include the following, as applicable:

I.

Material description and/or scope of work to be performed.

2.

Technical requirements with reference to applicable drawings, specifications, codes and standards identified by title, document number, revision and date. Any required procedures, such as special process instructions, are identified in such a way as to indicate source and need.

3.

Regulatory, administrative and reporting requirements. This includes IOCFR21 requirements, specifications, codes, standards, tests, inspections, and special processes. (The QA programmatic requirements of ASME NQA-1 or ANSI N45.2 may be used, where appropriate.)

4.

A requirement for a documented QA program.

5.

A requirement for the supplier to invoke applicable quality requirements on subtier suppliers.

6.

Provisions for access to supplier and subtier suppliers' facilities and records for inspections, surveillances and audits.

7.

Identification of documentation to be provided by the supplier.

8.

Provisions for documentation and dispositioning of nonconformances.

Documentation supporting the conformance of material and equipment with the procurement documents is to be available at the site prior to the installation whenever possible.

In those instances when equipment and material are received without the required documentation, installation may be accomplished if:

1.

The installation is controlled in accordance with the requirements of Section B.13, Corrective Action, of this QATR;

2.

The installed items are readily removable or more readily protected by installation than by segregated storage; and

3.

The supporting documentation is reasonably expected to arrive prior to the use of the equipment or material.

In no case shall material or equipment be relied upon for its safety function without receipt of proper documentation.

Spare and replacement parts are procured in accordance with the following provisions to assure that their performance and quality are at least equivalent to those of the parts that will be replaced:

1.

Specifications and codes referenced in procurement documents for spare or replacement items are the same or equivalent to those for the original items or to reviewed and approved revisions;

2.

Where quality requirements for the original items cannot be determined, requirements and controls are established by an engineering evaluation; and Page 13 of30

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3.

Any additional or modified design criteria imposed after previous procurement of the item(s), are identified and incorporated.

Appropriate controls are imposed for the selection, determination of suitability for intended use (critical characteristics), evaluation, receipt, and quality evaluation of commercial grade items or services to ensure that they will perform satisfactorily in service.

B.5 PROCUREMENT VERIFICATION Activities that implement approved procurement requests for material, equipment, and services are controlled to assure conformance with procurement document requirements. Controls include a system of supplier evaluation and selection, source inspection, examination and acceptance of items and documents upon delivery, and periodic assessment of supplier performance.

The degree of supplier surveillance (including review, inspection, or audit) required during design, fabrication, inspection, testing, and shipping shall be determined and documented. The objectives of supplier surveillance are to provide a sampling review of the supplier's quality assurance program implementation and/or verify product conformance with respect to the purchase order requirements. The extent of supplier surveillance will be consistent with the safety significance, complexity, quantity, and frequency of procurement of the item or service. As necessary, this may require verification of the activities of suppliers below the first tier. When suppliers perform work under their own QA programs, those programs are reviewed for compliance with the applicable requirements of IOCFR50 Appendix B and the contract.

Since CGG accepts the responsibility of verifying the conformance of commercial grade items and/or service, they may be procured from suppliers with no formal quality assurance program. In this instance, CGG dedicates the commercial grade item and/or service for safety-related use. A survey may be performed of commercial suppliers to assess what, if any documented controls are implemented in the manufacture of items or performance of services for CGG. Supplier controls evaluated to be satisfactory may be invoked as requirements within the purchase order and may be used as part of the basis for acceptance of the item. The depth of supplier evaluation varies according to the complexity and function of the item involved and to the role of the supplier in acceptance of the item.

When a supplier's certificates of conformance are used as part of the acceptance of an item or service, the validity of these documents is periodically evaluated and documented by the above-mentioned evaluation processes. Suppliers' certificates of conformance are reviewed for completeness and accuracy and are supplemented by receipt inspection to verify conformance to purchase requirements.

Receipt inspections are performed to verify that items are undamaged and properly identified and that required supplier-furnished documentation is available and correct. To be acceptable, the items and services must conform to procurement documents and have satisfied required inspection and test requirements. In addition, depending upon the nature, complexity and importance of the item and amount of source inspection or surveillance, selected characteristics may be examined or tested on a sampling basis to verify conformance to procurement requirements.

Items inspected are identified as to their acceptance status prior to storage or release for installation or use.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.6 IDENTIFICATION AND CONTROL OF ITEMS The identification and control of materials, parts, and components (including consumable materials and items with limited shelf life) is accomplished in accordance with written requirements and applies in any stage of fabrication, storage, or installation. Identification and control requirements are established either by an existing procedure or by requirement documents prepared during the planning stages of a project.

The identification and control requirements cover:

1. Traceability to associated documents (such as drawings, specifications, purchase orders, manufacturing test data and inspection documents, and physical and chemical mill test reports) to ensure that only correct and acceptable items are used

2. Specification of the degree of identification to preclude a degradation of the item's functional capability or quality

3. Proper identification of materials, parts, and components prior to release for manufacturing, shipping, construction, and installation B.7 HANDLING. STORAGE. AND SHIPPING Activities with the potential for causing contamination or deterioration that could adversely affect the ability of an item to perform its intended function and activities necessary to prevent damage or loss are identified and controlled. Controls are achieved through the use of appropriate procedures.

Procedures are used to control the cleaning, handling, storage, packaging, preserving and shipping of materials, components and systems in accordance with design and procurement requirements.

These procedures include the following functions:

I.

Cleaning, to assure that required cleanliness levels are achieved and maintained.

2.

Packaging and preservation, to protect against damage or deterioration. When necessary, these procedures provide for special environments such as inert gas atmospheres, specific moisture content levels and temperature levels.

3.

Handling, to preclude damage or safety hazards. Routine inspection of handling equipment is included.

4.

Storing, to minimize the possibility of loss, damage or deterioration of items in storage, including consumables such as chemicals, reagents and lubricants.

Storage procedures also provide methods to assure that items having limited shelf life are not used after their expiration date.

5.

Marking and labeling of items for packaging, shipment and storage is to be adequate to identify the shipment and to indicate the need for special environments and special control.

Controls have been established for the safe storage of hazardous materials Page 15 of 30

QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.8 TEST CONTROL Testing is performed to demonstrate that items will perform satisfactorily in service. The testing is performed in accordance with written procedures, at a frequency specified by the test program, that incorporate specified requirements and acceptance criteria.

The test program includes qualification, acceptance, preoperational, startup, surveillance and post-maintenance tests.

Requirements and acceptance criteria, including any prerequisites, instrumentation requirements and environmental conditions, are specified and met.

Test results are documented, evaluated, and their acceptability determined by a qualified individual or group.

Tests are performed in accordance with procedures and criteria that designate when tests are required and how they are to be performed. Such testing includes the following:

I.

Qualification tests, as applicable, to verify design adequacy.

2.

Tests of equipment and components to assure their proper operation prior to delivery or preoperational tests.

3.

Special tests to assure proper and safe operation of systems and equipment prior to startup tests or operations.

4.

Startup tests, including precritical, criticality, low-power and power ascension tests performed after fuel loading to assure proper and safe operation of systems and equipment.

5.

Surveillance tests to assure continuing proper and safe operation of systems and equipment.

6.

Post-maintenance tests.

Test procedures and instructions include provisions for the following, as applicable:

1.

The requirements and acceptance limits contained in applicable design and procurement documents.

2.

Test prerequisites such as calibrated instrumentation, required test equipment, degree of completeness of the item to be tested, suitable and controlled environmental conditions and provisions for data collection and storage.

3.

Verifying that test prerequisites have been met.

4.

Instructions for performing the test.

5.

Any hold points.

6.

Acceptance criteria.

7.

Documenting or recording test data and results.

8.

Verification of completion.

Unacceptable test results and test anomalies are evaluated in accordance with established procedures to determine the cause of the problem and the need for retest or for increasing test frequency.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.9 MEASURING AND TEST EQUIPMENT CONTROL Measuring and test equipment is identified, controlled, calibrated and adjusted at specified intervals to maintain accuracy within prescribed limits. The types of equipment covered by the program (such as instruments, tools, gages, standards, and nondestructive examination equipment) are defined in procedures.

Measuring and test equipment is uniquely identified and is traceable to its calibration test data. Labels or tags are attached to measuring and test equipment to display the next calibration due date. Where labels or tags cannot be attached, a control system is used that identifies to potential users any equipment beyond the calibration due date.

Measuring and test equipment is calibrated at specified intervals. These intervals are based on the amount of use, stability, characteristics and other conditions that could adversely affect the required measurement accuracy. Calibration standards are traceable to nationally recognized standards where they exist. Where national standards do not exist, provisions are established to document the basis for calibration.

Where practical, standards that have at least four times the required accuracy of the item being calibrated are used to calibrate measuring and test equipment. When this accuracy is not possible, these standards shall have an accuracy that assures that the equipment being calibrated will be within required tolerance.

In such cases, the basis of acceptance is documented and is authorized by responsible management personnel.

Standards normally have a greater accuracy than the equipment or installed plant instrumentation being calibrated. Standards with the same accuracy may be used when shown to be adequate for specific calibration requirements. The basis for this acceptance is documented and is approved by responsible management.

When measuring and testing equipment used for inspection and test is found to be out of calibration, evaluations are conducted to determine the validity of the results obtained since the most recent calibration. The results of evaluations are documented. Retests or reinspections are performed on suspect items as necessary.

B.10 INSPECTION. TEST, AND OPERATING STATUS Systems are established which ensure that the inspection, test and operational status of items is known and controlled. Non-operational status of systems and components for inspection, maintenance and tests is indicated by tagging, marking, logging or other specified means under procedural controls to prevent inadvertent use.

For modification activities (including item fabrication, construction, installation and test), procurement documents, service contracts and procedures specify the degree of control required for inspection and test status of items. Application and removal of status indicators such as tags, markings, labels, etc., are controlled by procedures.

The sequence of inspections, tests and other operations, and changes thereto, are controlled by procedures.

Changes in the approved sequence are controlled in accordance with the applicable procedures.

The status of inoperable systems, components, and supporting structures is known and controlled from the control room in accordance with procedures that are kept up-to-date to preclude unintentional operations.

The bypassing of required inspections, tests, and other critical operations is controlled to ensure that bypassed inspections or tests are properly documented and that the effect of bypassing the inspection or test is evaluated by the organization responsible for specifying the inspection or test.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.I I SPECIAL PROCESS CONTROL Special processes are controlled and are accomplished by qualified personnel using qualified procedures and equipment in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

Processes subject to special process controls are those for which the results are highly dependent on the control of the process or the skill of the operators, or both, and in which the specified quality cannot be readily determined by inspection or test of the product. Such processes include welding, heat treating, chemical cleaning, application of protective coatings, and nondestructive examination (NDE).

Requirements for control of special processes involve one or more of the following methods, each of which is approved by qualified personnel:

1.

Written instructions on the drawing or specification for the piece or assembly.

2.

Written procedure(s) including the specific application involved.

3.

Reference to a recognized code or standard published by a national society or institute.

4.

Combinations of above with addenda, exceptions, or alternates clearly indicated and in terminology familiar to personnel involved in planning, performing the process and evaluating the results.

When special process qualification requirements are not included in existing codes and standards, they are described in procedures that give details of the special process, personnel qualification requirements, equipment necessary, and special process qualification requirements.

Personnel with expertise in the discipline involved prepare special process procedures. The procedures are reviewed for technical adequacy by other personnel with the necessary technical competence, and are qualified by testing, as necessary.

Special process personnel qualification is determined by individuals authorized to administer the pertinent examinations.

Certification of qualification is based in part on examination results.

Personnel certification is kept current by performance of the special process(es) and/or reexamination at time intervals specified by applicable codes, specifications and standards. Unsatisfactory performance or, where applicable, failure to perform within the designated time intervals requires requalification and recertification.

For special processes that require qualified equipment, such equipment is qualified in accordance with applicable codes, standards and specifications.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.12 INSPECTION Inspections required to verify conformance of an activity to specified requirements or continued acceptability of items in service shall be planned and executed. Characteristics subject to inspection and inspection methods shall be specified. Inspection results shall be documented. Qualified persons other than those who performed or directly supervised the work being inspected shall perform inspection for acceptance.

Inspections are applied to procurement, maintenance, modification, testing, fuel handling, operation and in-service inspection to verify that items and activities conform to specified requirements. Inspection planning documents are prepared in accordance with established procedures.

Documentation of inspection planning identifies the following as applicable:

1.

Characteristics and activities to be inspected.

2.

Inspection organization.

3.

Accept/reject criteria.

4.

Hold points.

5.

Description of the method of inspection.

6.

Provisions for recording objective evidence of inspection results.

7.

Specific measuring and test equipment of the necessary accuracy for performing inspection.

Inspection is performed on selected operations where it is deemed necessary to verify conformance with specified requirements.

Process monitoring is used where direct inspection alone is impractical or inadequate. Both inspection and process monitoring are performed when control is inadequate without both.

Hold points are designated as mandatory inspection points when confirmation is needed that the work at that point is acceptable before additional work can proceed. Hold point inspections are performed, and designated inspection personnel release work for further processing or use. Only designated personnel may waive hold points.

Inspections are performed and documented in accordance with written procedures.

The results are evaluated and documented by qualified personnel in order to verify the acceptability of the item or work.

When items are reworked, the rework is re-inspected to the original or equivalent requirements.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.13 CORRECTIVE ACTION Controls have been established to ensure that conditions adverse to quality are identified promptly, documented, reviewed, and corrected as soon as practical. These controls are applied to deficiencies associated with the programmatic content, process, and implementation of the QA program as well as nonconformances.

For significant conditions adverse to quality, corrective action includes determining the cause and extent of the condition and taking appropriate action to preclude recurrence. These identified conditions, their causes, and corrective actions taken are documented and reported to appropriate levels of management for review and assessment.

Corrective action verification is performed for significant conditions adverse to quality prior to the closeout of the corrective action document. Unacceptable corrective action is reported to supervisory or management personnel directly responsible for resolving the issue, and to progressively higher levels of management until the issue is resolved.

Issues are periodically analyzed for the identification of adverse quality trends. The existence of an adverse quality trend is resolved in accordance with this section. A trend report is issued to management at intervals specified in approved procedures.

Material, parts, components or services, as applicable (including computer codes), that do not conform to requirements are controlled in order to prevent their inadvertent use. Nonconforming items are identified, documented in accordance with procedures, segregated when practical, and dispositioned.

Affected organizations are notified of nonconformances. Nonconforming items are reviewed for reportability.

Personnel authorized to disposition, conditionally release and close out nonconformances are designated.

Nonconformances identified at a supplier's facility and reported to CGG, for which the supplier has recommended a disposition of use-as-is or repair, are normally reviewed and the disposition approved by the procuring organization.

Any individual identifying an actual or suspected nonconforming item is responsible for documenting and reporting such nonconforming item promptly to supervisory or Q&PA personnel. Nonconforming items are identified by marking, tagging, or segregating, or by administrative controls that do not adversely affect the end use of the item.

Nonconformance control documents are submitted to responsible departments for resolution.

Documentation describes the nonconformance, the disposition of the nonconformance and the inspection requirements. It also includes signature approval of the disposition.

Nonconformance control documents are submitted to responsible departments for resolution.

Nonconformance control documents are not closed until corrective actions have been completed.

Nonconforming items are dispositioned as rework, repair, reject, or use-as-is.

Nonconformances to design requirements dispositioned use-as-is or repair are subject to design control measures commensurate with those applied to the original design. Reworked, repaired, and replacement items are inspected and/or tested in accordance with the original inspection and/or test requirements or acceptable alternatives to ensure that critical characteristics possibly affected by the nonconformance remain acceptable.

Nonconforming items may be conditionally released for installation, test, energization, pressurization, or use if the conditional release will not adversely affect nor preclude identification and correction of the nonconformance. Dispositions of conditionally released items are resolved before the items are relied upon to perform their safety-related functions.

Conditional release evaluations are documented, reviewed, and approved prior to implementation.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 B.14 DOCUMENT CONTROL Documents are controlled, issued and changed according to established procedures. Documents such as instructions, procedures and drawings, including changes thereto, are reviewed for adequacy, approved for release by authorized personnel, and are distributed and used at the location where a prescribed activity is performed. Controlled documents, including changes, are issued and distributed so that the documents are available as required at the work location prior to commencing work, and obsolete or superseded documents are removed from work areas and replaced by applicable revisions in a timely manner.

Changes to controlled documents, other than those identified as minor changes, are reviewed and approved by the same organizations that performed the original review and approval, or by the organizations designated in accordance with the procedures governing these documents. The criteria for establishing and defining minor changes are provided in approved implementing documents such as directives, procedures, and instructions.

Procedures are established for review, approval, issue, change and use of documents in the following categories:

1.

Design documents (e.g., calculations, drawings, specifications, analyses) including documents related to computer codes.

2.

As-built drawings, procedures and related documents.

3.

Procurement documents.

4.

Instructions and procedures for activities such as fabrication, construction, modification, installation, inspection, test, calibration, and station maintenance and operation.

5.

Procedures that implement the QA program.

6.

Updated Safety Analysis Report (USAR) or Updated Final Safety Analysis Report (UFSAR), as applicable.

7.

Reports of nonconformances and corrective action.

8.

Unit Technical Specifications.

9.

ISFSI updated Safety Analysis Report and Materials License (CCNPP).

10.

Vendor Manuals.

The Plant General Manager may designate specific procedures or classes of procedures in writing to be reviewed by Qualified Reviewers in lieu of review by the PORC. Review by Qualified Reviewers shall be in accordance with Addendum A.7.I.

Procedures required by Technical Specifications shall be approved by the Plant General Manager or by cognizant Managers/Directors, General Supervisors, and other supervisory personnel that report directly to a Manager/Director, prior to implementation as specified by administrative requirements.

The approval authority for specific procedures or classes of procedures shall be designated in writing by the Plant General Manager and shall be a different individual from the Qualified Reviewer.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 Temporary changes to procedures required by Technical Specifications may be made provided:

1.

The intent of the original procedures is not altered.

2.

The change is approved by two members of the plant management staff knowledgeable in the areas affected by the procedures, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

3.

The change is documented, reviewed by the PORC or by a Qualified Reviewer, and approved by the designated approval authority within 14 days of implementation.

B.15 RECORDS Organizations performing quality-affecting activities are responsible for forwarding the records they initiate to records management.

Each organization generating records is responsible for preparation, review, approval, and implementation of specific quality assurance record procedures for their area of responsibility.

Records to be controlled are delineated in Regulatory Guide 1.28 Revision 3. Sufficient records of items and activities are generated and maintained to document completed work. Items and activities requiring records include design, engineering, procurement, manufacturing, construction, inspection and test (e.g.,

manufacturer's proof and receipt), installation, operations, maintenance, modification and audits.

Requirements and responsibilities for preparation, inspection, identification, indexing, review, storage, retrieval, maintenance, safekeeping, retention, and disposition of quality assurance records are in accordance with applicable records procedures, codes, standards, and procurement documents.

Records of radioactive shipments shall be retained for 3 years beyond the date the licensee last engaged in these activities. These records must include the instructions, procedures, and drawings required by 10 CFR 71.111 to prescribe quality assurance activities and must include closely related specifications such as qualifications of personnel, procedures, and equipment. The records must include the instructions or procedures that establish the records retention program, which is consistent with applicable regulations and designates factors such as duration, location, and assigned responsibility.

Superseded procedures/instructions shall be retained for 3 years after they are superseded.

At CCNPP, Independent Spent Fuel Storage Installation (ISFSI) records pertaining to the design, fabrication, erection, testing, audits, maintenance, and use of structures, systems, and components important to safety are maintained for the duration of the ISFSI license.

Controls have been provided to ensure that records are protected from possible destruction.

Within established time-intervals, completed lifetime records are transmitted to the records management organization for incorporation into the records storage and retrieval system.

Records may be original documents, legible copies, or in various microfilm formats. Records may also be stored as electronic images using technology that does not allow deletion or modification of record images.

Authorized personnel may issue corrections or supplements to records. Procedures address acceptable methods of making corrections to records.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 C.

ASSESSMENT C.1 METHODOLOGY Personnel responsible for the self-assessment function, including onsite and offsite nuclear safety review committee activities, audits, and other independent assessments are cognizant of day-to-day activities so that they can act in a management advisory function. Self-assessment activities are technically and performance oriented, with the primary focus on the quality of the end product and a secondary focus on procedures and processes. Self-assessments are accomplished using instructions, procedures, or other appropriate means that are of a detail commensurate with the activity's complexity and importance to safety.

Ongoing and periodic self-assessments are used to identify safety concerns and improve performance.

Self-assessments compare actual performance to management expectations, performance of other high-performing organizations, industry standards of excellence, and regulatory requirements.

Skilled, knowledgeable internal and external personnel perform self-assessments. Improvement needs identified by self-assessments are assigned for action and tracked through completion.

Benchmarking is used to identify options for solving problems, improving performance, and emulating best practices. Managers and coworkers frequently observe work and training activities to recognize strong performance and identify needed improvements. Performance measures are used to identify areas of strong performance, areas needing improvement, and precursors to significant problems.

The organization supports and learns from participation in self-assessments and evaluations at other facilities.

Results of self-assessments, observations, corrective actions, and independent oversight assessments are reviewed for underlying problems that need resolution. Self-assessment and corrective action program effectiveness is periodically assessed and the programs are adjusted.

C.2 QUALITY AND PERFORMANCE ASSESSMENT A program of planned and periodic assessments is established and implemented to confirm that activities affecting quality comply with the QA program and that the QA program has been implemented effectively. Assessments provide comprehensive independent evaluation of activities and procedures.

Monitoring and assessment activities are conducted in sufficient depth to identify potentially significant nuclear safety problems. Planning activities identify the characteristics and activities to be assessed and the acceptance criteria. Assessments are conducted using the predetermined acceptance criteria. Use of relevant industry and in-house operating experience information is reviewed during periodic assessments.

Scheduling and resource allocation are based on the status and safety importance of the activity or process being assessed. Scheduling is dynamic to allow for additional assessments in areas where QA program effectiveness is in doubt. Activities of groups performing independent monitoring and assessment are coordinated to encompass all matters relevant to nuclear safety and reliability.

Audits of selected aspects of operational phase activities are performed with a frequency commensurate with their strength of performance and safety significance and in such a manner as to assure that an audit of all safety-related functions is completed within a period of two years. Audits specified in regulations are performed at the frequencies noted in their respective NRC-approved plans. Audits and assessment activities may be conducted continuously.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 In addition to the audit subjects specified in Regulatory Guide 1.33 Revision 2 and ANSI N18.7-1976/ANS-3.2, audits performed within a period of two years shall also encompass:

1.

the facility fire protection program and implementing procedures;

2.

inspection and audit of the fire protection and loss prevention program performed by non-licensee personnel. The personnel may be representatives of a fire insurance brokerage firm or other qualified individuals;

3.

the radiological environmental monitoring program and the results thereof;

4.

the Offsite Dose Calculation Manual and implementing procedures;

5.

the Process Control Program and implementing procedures for processing and packaging of radioactive wastes;

6.

the performance of activities required by the QA program for effluent and environmental monitoring; and

7.

the performance of activities required by the QA program to meet the criteria of 10CFR50, Appendix B.

A grace period of 90-days is applied to audits performed per Regulatory Guide 1.33 Revision 2, ANSI N18.7-1976/ANS-3.2, and those listed above.

The grace period does not allow the "clock" for a particular period to be reset forward. For example, if a biennial audit is due on June 15 of a particular year, but is not performed until August 13, the next due date for that audit will be June 15 of the second year following. However, the clock for an activity is reset backwards by performing the activity early.

The 90-day grace period does not apply to audits specified in regulations (i.e. Emergency Preparedness, Security, and Fitness for Duty).

Assessment results are documented and reviewed by the assessor's management and by management having responsibility in the area assessed. Actions to address issues identified through independent monitoring and assessment activities are tracked and completed in a timely manner. Follow-up action, including a re-look at deficient areas, is initiated as necessary.

When work carried out under the requirements of the QA program is delegated to others, implementation of that work is assessed by CGG.

If a difference of opinion arises between Q&PA personnel and those of other Sections or Departments, the dispute is resolved as follows: The site management position responsible for Q&PA first tries to resolve the matter with the organization responsible for conducting the activity. If a resolution cannot be obtained, the matter is referred up through the following management personnel until it is resolved:

1.

The site management position responsible for Q&PA and the site management position responsible for performing the activity. NOTE: If the dispute is internal to Q&PA, the site management position responsible for Q&PA will settle the issue.

2.

The fleet management position responsible for Q&PA and the appropriate Site Vice President

3.

The Senior Vice President and Chief Nuclear Officer Individuals assigned to perform independent monitoring and assessments have the necessary experience, training, and authority to conduct the reviews, audits, or analyses.

Individuals assigned to perform independent monitoring and assessments do not have line responsibility for the area being assessed.

The effectiveness of independent monitoring and assessments are periodically evaluated. Results are reported to senior management, and corrective actions are implemented as needed.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 ADDENDUM A.7.1 REVIEW FUNCTIONS OF THE PORC, QUALIFIED REVIEWERS, NSRB, AND ISEG 1.0 PLANT OPERATIONS REVIEW COMMITTEE (PORC) 1.1 FUNCTION The Plant Operations Review Committee shall function to advise the Plant General Manager on all matters related to nuclear safety.

1.2 COMPOSITION The Plant Operations Review Committee shall be composed of at least five members, including the Chairperson. Members shall collectively have experience in the following areas:

Nuclear Operations Maintenance and Plant Modifications Engineering Chemistry Radiation Protection The Plant General Manager shall appoint members in writing. Members shall have a minimum of eight years power plant experience of which a minimum of three years shall be nuclear power experience. At least one member holds, or has held, an SRO License for the respective nuclear power plant or similar unit.

1.3 CHAIRPERSON The Plant General Manager shall appoint the Chairperson and alternate Chairpersons of the Plant Operations Review Committee in writing. Chairpersons shall have a minimum of 10 years power plant experience of which a minimum of three years shall be nuclear power experience.

1.4 ALTERNATES The Plant General Manager shall appoint all alternate members in writing. Alternate members shall have a minimum of eight years power plant experience of which a minimum of three years shall be nuclear power experience.

1.5 MEETING FREQUENCY The Plant Operations Review Committee shall meet at least once per calendar month and as convened by the Plant Operations Review Committee Chairperson or one of the designated alternate Chairpersons.

1.6 QUORUM A quorum of the Plant Operations Review Committee shall include the Chairperson or one of the designated alternate Chairpersons and shall consist of a minimum of five individuals, including alternates. No more than two alternates shall participate as voting members in Plant Operations Review Committee activities at any one time. Individuals making up the quorum shall possess sufficient technical expertise to conduct an adequate review.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 1.7 RESPONSIBILITIES The Plant Operations Review Committee shall be responsible for the following except for those items designated for review by Qualified Reviewers in accordance with Addendum section 2:

a.

Review of all procedures required by Technical Specifications and changes thereto.

Cross-disciplinary reviews of these procedures are conducted in accordance with administrative procedures in addition to the reviews conducted by the Plant Operations Review Committee or Qualified Reviewer.

b.

Review of all proposed tests and experiments that affect nuclear safety.

c.

Review of all proposed changes to Technical Specifications or the Operating License prior to their submittal to the NRC.

d.

Review of changes to the Offsite Dose Calculation Manual.

e.

Review of all evaluations required by I OCFR50.59 (and I OCFR72.48 for CCNPP only).

f.

Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Plant General Manager, the Site Vice President, and to the Chairperson of the Nuclear Safety Review Board.

g.

Review of all Reportable Events.

h.

Review of unit operations to detect potential hazards to nuclear safety.

i.

Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant General Manager or the Chairperson of the Nuclear Safety Review Board.

1.8 AUTHORITY The Plant Operations Review Committee shall:

a.

Recommend to the approval authority approval or disapproval of procedures considered under 1.7.a above.

b.

Recommend to the Plant General Manager approval or disapproval of items considered under 1.7.b through e above.

c.

Render determinations in writing with regard to whether or not each item considered under I.7.a through e requires prior NRC approval.

d.

Evaluate root causes and recommended actions to prevent recurrence for items considered under I.7.f through h above.

e.

Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Site Vice President and the Chairperson of the Nuclear Safety Review Board of disagreement between the Plant Operations Review Committee and the responsible approval authority in the case of item 1.7.a above or between the Plant Operations Review Committee and the Plant General Manager; however, the Plant General Manager shall have responsibility for resolution of such disagreements pursuant to Technical Specifications.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 1.9 RECORDS The Plant Operations Review Committee shall maintain written minutes of each meeting and copies shall be provided to the Site Vice President, Chairperson of the Nuclear Safety Review Board, and the Plant General Manager. Open items shall be assigned, tracked and resolved.

2.0 QUALIFIED REVIEWERS 2.1 FUNCTION The Plant General Manager may designate specific procedures or classes of procedures described in 1.7.a above to be reviewed by Qualified Reviewers in lieu of review by the Plant Operations Review Committee.

2.2 AUTHORITY Qualified Reviewers shall:

a.

Recommend to the approval authority approval or disapproval of designated procedures and changes considered under 1.7.a above and

b.

Render determination in writing with regard to whether or not each procedure under 1.7.a above requires prior NRC approval.

c.

Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Plant General Manager, Site Vice President, and the Chairperson of the Nuclear Safety Review Board of disagreements between the Qualified Reviewer and the approval authority. The Plant General Manager shall have responsibility for resolution of such disagreements pursuant to Technical Specifications.

2.3 CERTIFICATION Qualified Reviewers shall be nominated, trained, and certified in accordance with administrative procedures. Certification shall be by a department manager/director.

2.4 CERTIFICATION REQUIREMENTS Certification requirements of personnel designated as Qualified Reviewers shall be in accordance with administrative procedures.

Qualified Reviewers shall have:

a.

A Bachelor's degree in engineering, related science, or technical discipline, and two years of nuclear power plant experience; OR

b.

Six years nuclear power plant experience; OR

c.

Equivalent combination of education and experience as approved by a department manager/director.

2.5 RECORDS Review of procedures by Qualified Reviewers shall be documented in accordance with administrative procedures.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 3.0 NUCLEAR SAFETY REVIEW BOARD (NSRB) 3.1 FUNCTION The Nuclear Safety Review Board shall function to provide independent review and audit of designated activities in the areas of:

a.

nuclear power plant operations

b.

nuclear engineering

c.

chemistry and radiochemistry

d.

metallurgy and non-destructive examination

e.

instrumentation and control

f.

radiological safety

g.

mechanical and electrical engineering

h.

quality assurance practices 3.2 COMPOSITION The Nuclear Safety Review Board shall be composed of at least seven members, including the Chairperson. Members of the Nuclear Safety Review Board may be from CGG or organizations external to CGG and shall collectively have expertise in all of the areas of 3.1 above.

3.3 QUALIFICATIONS The Senior Vice President and Chief Nuclear Officer (CNO) shall appoint the Chairperson of the Nuclear Safety Review Board in writing. The Chairperson shall appoint members and alternates in writing. The Chairperson and all members (primary and alternate) shall have an academic degree in engineering or a physical science, or the equivalent, and in addition shall have a minimum of five years technical experience in one or more areas given in 3.1 above. No more than two alternates shall participate as voting members in Nuclear Safety Review Board activities at any one time.

3.4 CONSULTANTS Consultants shall be utilized as determined by the Nuclear Safety Review Board Chairperson to provide expert advice to the Nuclear Safety Review Board.

3.5 MEETING FREQUENCY The Nuclear Safety Review Board shall meet at least once per six months.

3.6 QUORUM The quorum of the Nuclear Safety Review Board necessary for the performance of the Nuclear Safety Review Board review and audit functions shall consist of more than half the Nuclear Safety Review Board membership or at least four members, whichever is greater. This quorum shall include the Chairperson or his appointed alternate and the Nuclear Safety Review Board members, including appointed alternates, meeting the requirements of 3.3 above. No more than a minority of the quorum shall have line responsibility for operation of the plant.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0 3.7 SUBCOMMITTEES The Chairperson may establish subcommittees to perform reviews of selected items enumerated in 3.8 and 3.9 below. Each subcommittee shall be chartered in writing, have at least three members/alternates, and provide reports to the full committee on the results of its reviews with any appropriate recommendations.

3.8 REVIEW The Nuclear Safety Review Board shall review:

a.

IOCFR50.59 (and IOCFR72.48 for CCNPP) evaluations for changes to the facility or procedures and conducting tests or experiments completed under the provisions of 10 CFR 50.59 and 10 CFR 72.48 to verify that such actions did not require prior NRC approval.

b.

Proposed changes in Technical Specifications or the Operating License.

c.

Violation of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

d.

Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.

e.

All Reportable Events.

f.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

g.

Reports and meeting minutes of the Plant Operations Review Committee.

3.9 AUDITS Audits of facility activities shall be performed under the cognizance of the Nuclear Safety Review Board. These internal audits are discussed in Section C.2 of the QA Topical Report.

3.10 AUTHORITY The Nuclear Safety Review Board reports to the CNO. This includes direct access to the CNO for any nuclear safety issues. The CNO provides oversight of Nuclear Safety Review Board activities. The CNO will receive reports and meeting minutes, and can provide direction to the Nuclear Safety Review Board Chairperson regarding specific areas for review. Additionally, the Nuclear Safety Review Board Chairperson will brief the Site Vice-President and the CNO regarding NSRB activities and issues.

3.11 RECORDS Records of Nuclear Safety Review Board activities shall be prepared, approved and distributed as indicated below:

a.

Minutes of each Nuclear Safety Review Board meeting shall be prepared, approved and forwarded to the CNO within 14 days following each meeting.

b.

Reports of reviews encompassed by 3.8 above, shall be prepared, approved and forwarded to the CNO within 14 days following completion of the review.

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QUALITY ASSURANCE TOPICAL REPORT Revision 0

c.

Audit reports encompassed by 3.9 above, shall be forwarded to the CNO and to the management positions responsible for the areas audited within 30 days after completion of the audit.

4.0 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Independent safety review is performed to meet the individual unit's commitment to NUREG-0737, Section l.B.l.2, "Independent Safety Engineering Group," as described in the unit's safety analysis report.

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TABLE A.7.2 CGG'S CONFORMANCE TO REGULATIONS, REGULATORY GUIDES, AND INDUSTRY STANDARDS Revision of Industry Standards Applicable to the CGG Quality Assurance Program Some of the industry standards listed in this table and the QATR identify other standards that are required, and some Regulatory Guides define the revisions of those standards that are acceptable to the NRC. CGG's QA Program was developed to respond to the specific revision of the documents listed in this table and is not necessarily responsive to other documents listed in the referenced industry standards.

Ninety-Day Grace Period A grace period of ninety-days is applied to several activities specifically defined in this table. The grace period will not allow the "clock" for a particular activity to be reset forward. For example, if an annual activity is due on June 15th of a particular year, but is not performed until August 13th, the next due date for that activity will be June 15th of the following year. However, the clock for an activity is reset backwards by performing the activity early.

10CFR21 Reporting of Defects and Noncompliance CGG Conforms 10CFR50 Appendix A, Criterion 1 Quality Standards and Records COG Conforms IOCFR50 Appendix B Quality Assurance Criteriafor Nuclear Power Plants and Fuel Reprocessing Plants CGG Conforms IOCFR50.55(e)

Conditions of Construction Permits (e)

CGG Not Applicable - Applies to plants in the construction phase 10CFR50.55a Codes and Standards CGG IOCFR50.55a specifies ASME Section Xl code dates. The CCNPP, Ginna, and NMPNS programs conform to IOCFR50.55a with the specific editions and addenda of Section Xl specified in their Inservice Inspection Procedures.

10CFR50.59 Changes, Tests and Experiments CGG Conforms QATR Revision 0 Page I of 17

10CFR55 Operators'Licenses COG Conforms IOCFR72 Subpart G Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste - Quality Assurance CGG CCNPP Conforms - Quality Assurance (ISFSI). This is not applicable to Ginna or NMPNS.

BTP C51EB 9.5-1 Reg. Pos. 2 and 4 per SRP 9.5.1 CCNPP In lieu of the Programmatic QA Guidance, the CCNPP Quality Assurance Program is based on I OCFR Part 50, Appendix B; the quality assurance guidance in Appendix A to BTP APCSB, 9.5-1, "Guidelines for Fire Protection For Nuclear Power Plants Docketed Prior to July 1, 1976; and the NRC's guidance document "Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls, and Quality Assurance."

GINNA In lieu of the Programmatic QA Guidance, fire protection controls are in accordance with APCSB 9.5-I, regulatory position IV b.6 and IV b.7.

NMPNS In lieu of the Programmatic QA Guidance, NMPNS is committed to BTP APCSB 9.5-1, Appendix A -

1976, Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1, 1976 QATR Revision 0 Page 2of 17

Regulatory Guide 1.143 Rev. 1 Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed In Light Water-Cooled Nuclear Power Plants CGG Alternative - The SARs for CCNPP, Ginna, and Nine Mile Point Unit I describe the design and quality assurance provisions applied to existing radioactive waste management systems, structures, and components. New systems, structures, components and configuration changes to existing items meet the design and quality assurance provisions described in the respective SAR or the intent of those specified by this Regulatory Guide.

CCNPP The radioactive waste processing (NUKEM) skid is designed to meet the requirements of the Regulatory Guide with certain clarifications:

1. The ultra-filtration membranes (which could be perceived as piping components) are manufactured from fiberglass or polyvinylidene fluoride and meet the manufacturer standards for intended service.

2. Where practical, quick disconnects (CAMLOCK) and limited NPT connections, flanges, and victaulic are used.

3. Line sizes less than 3/4" are used.

NMP2 Alternative -

A. Liquid Waste System:

The fiberglass tanks purchased for the LWS have been designed in accordance with the National Institute of Standards and Technology (NIST) Product Standard (PS) PS 15-69, Custom Contact-Molded Reinforced-Polyester Chemical-Resistant Process Equipment, as identified in NMP2 PSAR, Table C-lOb. NBS PS 15-69 provides the necessary design and fabrication requirements to ensure the integrity of the tanks without the additional cost of burst testing.

The RWCU phase separator tanks (2LWS-TK6A & 6B), which had been purchased as code-stamped ASME VIll vessels, had their code stamps removed because they were not re-hydro tested after a nozzle was added to the top head of each of the two vessels in the field. The vessels still satisfy the intent of the requirements of this regulatory guide in that they are designed and fabricated to the requirements of ASME VIII (including the added nozzles) using materials which meet ASME VIII requirements, and the shop hydrotest established the integrity of the vessels before the nozzles were added. The nozzles were added near the top of the vessels' heads and the vessels see only atmospheric operating conditions (although designed for a nominal 15-psi design pressure). The new nozzles were added to allow improved operation of the vessels' level transmitters and are identical to the original level transmitter nozzles which were blind flanged and abandoned. Therefore, the added nozzles do not affect the proven integrity of the vessels in this application.

The Thermex unit (2LWS-FLT102), which is leased equipment, uses CPVC and PVC components.

The Thermex unit is connected to plant components using reinforced, non-collapsible hoses. The equipment vendor has evaluated these components as being suitable for processing low-level radioactive waste. All thermoplastic materials (CPVC and PVC components) comply with ASTM material and dimensional standards. Assembled liquid piping or hoses are hydrostatically tested at 150 percent of the maximum design operating pressure for the limiting assembly. Chemical transfer hoses (nonmetallic) satisfy the requirements of ANSI B3 1.1-1992 Appendix III and are inspected/tested/replaced on a routine interval to ensure equipment reliability.

B. Offgas System:

The charcoal adsorbers of the offgas system are not designed to the seismic requirements of this regulatory guide. Offsite dose calculations in accordance with Chapter 15.7.1 of the NMP2 USAR show that release of gaseous activity due to failure of the charcoal adsorbers results in offsite doses less than 0.5 Rem to the whole body. In accordance with Regulatory Guide 1.29, this permits classification as nonseismic. At the time of design and procurement of the offgas system (July 1974), Regulatory Guide 1.29 Revision I established the seismic requirements for the radioactive waste processing systems.

C. Waste Solidification System:

The waste solidification system complies with the requirements of NRC Branch Technical Position QATR Revision 0 Page 3 of 17

ETSB 11.1 Revision I as outlined in Werner and Pfleiderer Corporation (WPC) Topical Report No.

WPC-VRS-001 Revision 1, dated May 1978, with exceptions as discussed in Section 11.4.3. The waste sludge tank is designed, fabricated, examined, and tested (hydrotest at 1.5 times design) in accordance with the requirements of ASME Code Section Vill, Division 1, with no codes stamp. The WPC topical report lists API 620 or 650 for this tank. The WPC Topical Report has been accepted by the NRC as satisfying the requirements of ETSB 11.1 Revision 1, which are consistent with the requirements of RG 1.143. For the waste sludge tank, the requirements of ASME VIll are more stringent than the requirements of API 620 or 650.

Regulatory Guide 1.152 Rev. 0 Criteria for Programmable Digital Computer System Software in Safety-Related Systems of Nuclear Power Plants CGG In lieu of the regulatory guidance, CGG conforms to Generic Letter 95-02, and its endorsement of NUMARC/EPRI Report TR-I 02348 "Guidelines on Licensing Digital Upgrades".

Regulatory Guide 1.29 Rev. 3 Seismic Design Classification CGG Alternative - For CCNPP and Ginna, seismic design requirements for existing structures, systems, and components performing functions listed in positions C.l and C.3 of the Regulatory Guide are specified in the UFSAR. New structures, systems, and components, and configuration changes meet the seismic design requirements of this regulatory guide or the UFSAR. The pertinent quality assurance requirements of I OCFR50, Appendix B are applied as required by positions C.l and C.4 of this Regulatory Guide, irrespective of an item's seismic design.

Portions of existing structures, systems, and components with failure consequences described in position C.2 of this guide are designed and constructed to seismic requirements specified in the UFSAR. New structures, systems, and components, and configuration changes meet the design and construction seismic requirements of the UFSAR or this Regulatory Guide. A quality assurance program similar to IOCFR50, Appendix B is applied to the SSE failure prevention function of these items. These items are not considered basic components pursuant to IOCFR21.

NMPI Alternative - Those structures and components whose failure could cause significant release of radioactivity or which are vital to safe shutdown and isolation of the reactor are designated as seismic Class 1. Specific details regarding seismic design requirements are contained in the NMPI UFSAR.

For Class I structures, systems, and components, no quantitative basis is used to determine the limit for "significant release of radioactivity." The qualitative basis is that Class I systems are those whose failure could result in a continuous, uncontrolled release of radioactivity that could not be readily terminated.

There are two subclasses of Class I piping systems: (I) Class I systems which were designed using static seismic analysis, and (2) Class I systems which were designed using dynamic seismic analysis.

Class I systems were designed using static seismic analysis if they were low pressure and/or low temperature piping systems.

The codes used in the design of Class I system piping and containment isolation valves at the time of construction included ASME Section 1-1962 or ANSI B31.1-1955 and ANSI B 16.5-1955, with requirements of ASME Section 111-1965 for nondestructive testing (NDT). For subsequent modifications, Regulatory Guide recommendations are followed.

NMP2 Conforms QATR Revision 0 Page 4of 17

Regulatory Guide 1.36 Rev. 0 Nonmetallic Thermal Insulationfor Austenitic Stainless Steel CCNPP Conforms GINNA This Regulatory Guide is adopted for the testing of insulating materials installed on or near safety related stainless steel piping. Insulating materials are not considered basic components pursuant to I OCFR21 and thus the supplier is not required to have a quality assurance program to cover the testing, lot control, and contamination control provisions of this Regulatory Guide. A quality assurance program similar to I OCFR50, Appendix B is applied to insulating materials on or near Ginna Station safety related stainless steel piping and components.

NMPI Corrosion and chemical analysis testing of thermal insulation is conducted in accordance with the requirements of this regulatory guide. Appropriate provisions and requirements for material receiving, handling, storage, and installation are included in engineering specifications.

NMP2 Alternative - Nonmetallic thermal insulation for austenitic stainless steel, including filler material for encapsulated insulation, complies with Regulatory Guide 1.36 Revision 0 with the exception of packaging and shipping requirements of Paragraph C. l of this guide. In lieu of controlled packaging and shipping, receipt inspection and tests are required by specification. This consists of visual inspection for physical or water damage to all cartons. Damaged canons are segregated. The potentially contaminated insulation is not accepted unless randomly selected samples from each carton are shown to be acceptable after being re-subjected to the production test outlined in RG 1.36.

Purchase and/or erection specifications include the requirements as delineated above. No nonmetallic insulation will be in direct contact with safety-related austenitic stainless steel fluid systems within the primary containment.

Regulatory Guide 4.15 Rev. I QualityAssurancefor Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment CGG CGG conforms to the intent of this Regulatory Guide as addressed in CCNPP chemistry procedures and the Offsite Dose Calculation Manual; Ginna's Process Control Program and applicable Ginna effluent and environmental radioactivity measurements; and NMPNS Units' I & 2 Offsite Dose Calculation Manuals and applicable effluent and environmental radioactivity measurements.

Regulatory Guide 7.10 Rev. I Establishing Quality Assurance Programsfor Packaging Used in the Transport of Radioactive Material CGG CCNPP Conforms. Ginna and NMPNS conform to the intent of this Regulatory Guide as addressed in each unit's Process Control Program.

QATR Revision 0 Page 5 of 17

ANSI N18.7-1976/ANS-3.2 Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants Regulatory Guide 1.33 Rev. 2 Quality Assurance Program Requirements (Operation)

CGG CGG conforms to Regulatory Guide 1.33 Revision 2 and its endorsement of ANSI N 18.7-1976/ANS-3.2, with the following exceptions:

In lieu of the requirements of ANSI N18.7-1976/ANS-3.2 Section 5.2.2, CGG does not require the supervisor in charge of the shift to be the Senior Reactor Operator (SRO) approving temporary changes to procedures; any active SRO (either on-shift or on-staff) may provide the SRO approval for procedure changes.

In lieu of the requirements of Regulatory Guide 1.33 Revision 2 Section C, Regulatory Position 4, the audit frequency for all safety-related functions is at least once every two years (except as otherwise required in regulations).

CGG conforms to ANSI N18.7-1976/ANS-3.2 Section 4.5, Audit Program, with the exception that a 90 day grace period may be applied to the 24 month frequency for performing internal audits.

In lieu of the master surveillance schedule requirements of ANSI NI 8.7-1976/ANS-3.2 Section 5.2.8, independent master schedules may exist for different programs such as inservice inspection (ISI),

pump and valve testing, and Technical Specification surveillance testing.

Implementation of ANSI N18.7-1976/ANS-3.2 Section 5.2.9, Plant Security and Visitor Control, is addressed in the respective site security plan.

In lieu of the requirements of ANSI N18.7-1976/ANS-3.2 Section 5.2.15, routine plant procedures do not receive biennial reviews, but are subject to programmatic controls which continually identify needed procedure revisions. Routine plant procedures that have not been used for two years are reviewed consistent with pre-evolution briefing requirements. Additionally, procedure use and adherence requirements state clear procedure compliance expectations and the need to stop the job if a procedure cannot be performed as written. Non-routine plant procedures whose usage may be dictated by an event (such as Emergency Operating Procedures, Abnormal Operating Procedures, and Emergency Response Plan Implementing Procedures) receive biennial reviews according to administrative procedures. Applicable plant procedures are reviewed following an unusual incident and following any modification to a system. The biennial audit of the Procedures Program includes a representative sample of plant procedures. The audit helps ensure the acceptability of the procedures and verify that the Procedures Program is being implemented effectively.

NMPNS In lieu of the requirements for the Onsite Operating Organization per ANSI N18.7-1976/ANS-3.2 Section 3.4.2, NMPNS will comply with those described in the applicable sections of the Nine Mile Point Units I and 2 SARs and Technical Specifications.

In lieu of the requirements for reviews and audits per ANSI N18.7-1976/ANS-3.2 Section 4, NMPNS will comply with those described in the applicable sections of the Nine Mile Point Units I and 2 SARs, Technical Specifications, and this QATR.

In lieu of the requirements for procedures per Regulatory Guide 1.33 Revision 2 Appendix A, NMPNS will comply with those described in the applicable sections of the Nine Mile Point Units I and 2 SARs, Technical Specifications, and this QATR.

In lieu of the administrative controls for written procedures specified in ANSI NI 8.7-1976/ANS-3.2 Section 5.3, NMPNS will comply with those controls described in the applicable sections of the Nine Mile Point Units I and 2 SARs, Technical Specifications, and this QATR.

QATR Revision 0 Page 6 of 17

ASME NQA-1-1994, Part I Quality Assurance Program Requirements for Nuclear Facilities Regulatory Guide 1.28 Rev. 3 Quality Assurance Program Requirements (Design and Construction)

CGG In lieu of ANSI/ASME NQA-1-1983 and the NQA-la-1983 Addenda endorsed by Regulatory Guide 1.28 Revision 3, CGG is committed to ASME NQA-1-1994 Part I and Regulatory Guide 1.28 Revision 3 with the following exceptions:

Consistent with the guidance found in Regulatory Guide 1.28 Revision 3, Section B Paragraph 8, personnel performing inspection and testing that are qualified to the guidance contained in Regulatory Guide 1.8 (or ANS-3.1) need not be qualified in accordance with the requirements of NQA-I.

In lieu of the internal audit requirements of Regulatory Guide 1.28 Revision 3, Regulatory Position 3.1, CGG will audit applicable elements of the QA Program at least once every 2 years, in accordance with ANSI N 18.7-1976/ANS-3.2. (See exceptions to ANSI N 18.7-1976/ANS-3.2)

In lieu of the annual evaluations of suppliers required by Regulatory Guide 1.28 Revision 3, Regulatory Position 3.2, CGG may perform ongoing evaluations of suppliers. Where applicable, this evaluation takes into account: (I) a review of supplier-furnished documents such as certificates of conformance, nonconformance notices, and corrective actions; (2) the results of previous source verifications, audits, and receiving inspections; (3) the operating experience of identical or similar products furnished by the same supplier; and (4) the results of audits from other sources (e.g.,

customer, ASME, or NRC audits). The results of the evaluations are reviewed and appropriate corrective actions taken. Adverse findings resulting from these evaluations are periodically reviewed in order to determine if, as a whole, they result in a significant condition adverse to quality and to provide input to support supplier audit activities conducted by the licensee or a third party auditing entity.

CGG's ISI Plans endorse ASME Code Section XI. The version of the ASME code endorsed is updated periodically. ASME Code Section XI references standards for the qualification and certification of nondestructive testing personnel.Section XI of the ASME Code contains specific requirements for nondestructive examination and also references the use of other supplementary standards for the qualification and certification of personnel performing nondestructive examinations. The applicable versions of the standards referenced in Section XI of the ASME code, as permitted for use by 10 CFR Part 50.55a, may be used for the qualification and certification of personnel performing nondestructive examinations required by Section III and Section Xl of the ASME Code in lieu of the standard identified in ASME NQA-1-1994 Supplement 2S-2, (SNT-TC-IA, June 1980) provided that other applicable rules contained in Section XI of the ASME Code are met.

Interpretation of ASME NQA-I-1994 Supplement 2S-3: Personnel who perform audits that are outside the scope of I OCFR50 Appendix B are not required to be qualified in accordance with ASME NQA-I-1994 Supplement 2S-3.

In lieu of the audit participation requirements of ASME NQA-1-1994 Supplement 2S-3 Section 3.3, CGG requires prospective Lead Auditors to demonstrate their ability to effectively implement the audit process and effectively lead an audit team. This demonstration process is described in administrative procedures and shall be evaluated and the results of the demonstration documented. The prospective Lead Auditor shall have participated in at least one nuclear quality assurance audit within the year preceding the individual's effective date of qualification. Upon successful demonstration of the ability to effectively lead audits, the individual may be certified as a Lead Auditor.

In lieu of the requalification requirements of ASME NQA-1-1994 Supplement 2S-3 Section 4.2, CGG requalifies Lead Auditors on the basis of the satisfactory performance of one audit, as observed by a qualified Lead Auditor.

A grace period of 90 days may be applied to the performance of annual evaluations of inspection, QATR Revision 0 Page 7 or 17

examination and testing personnel qualifications defined in ASME NQA-1-1994 Supplement 2S-1 Section 2.6.

A grace period of 90 days may be applied to the performance of annual lead auditor recertifications described in ASME NQA-I-1994 Supplement 2S-3 Sections 4.1 and 6.3.

A grace period of 90 days may be applied to the performance of triennial supplier audits described in Regulatory Guide 1.28 Revision 3 Section 3.2.

Interpretation of ASME NQA-1-1994 Supplement 4S-I: Section 2.3 requires that suppliers have a documented quality assurance program in accordance with ASME NQA-1-1994 Part 1. Since many existing suppliers of basic components maintain quality assurance programs based on ANSI N45.2 or NQA-I-(Various Years), it is alternately acceptable to require that suppliers have a quality assurance program in accordance with N45.2 or NQA-..

ASME NQA-1-1994 Supplement 7S-I Paragraph 8.2.4 states "...post-installation test requirements and acceptance documentation shall be mutually established by the Purchaser and Supplier." In exercising ultimate responsibility for its QA Program, CGG establishes post-installation test requirements, giving due consideration to supplier recommendations.

CCNPP In lieu of the reinforced concrete, concrete block, masonry, or equal construction requirements of ASME NQA-1-1994 Supplement 17S-I Section 4.4.1 (a), the records vault is entirely enveloped by a structurally sound, fire resistive building. Second, the vault rests on a reinforced slab on grade and its walls extend fully to the underside of the structural deck. Third, the walls of the vault are constructed of gypsum wallboard on metal studs per Underwriters Laboratory Test Number U412, assuring the equivalent of 2-hour fire resistive construction. This is equal construction to concrete block in terms of fire protection. The walls carry no structural load; hence, they provide equivalent structural integrity to that needed of concrete block. (See Note).

ASMENQA-1-1994 Supplement 17S-1 Section 4.4.1(b) requires floor and roof drainage control. If a floor drain is provided, a check valve (or equal) shall be included. In lieu of this requirement, the vault is contained within an environmentally protected building. As such, it has no roof, or need for floor drain.l (See Note).

Note: These responses have been forwarded to the NRC by the CCNPP letter dated 02/11/83 from Robert G. Nichols, Sr. Facilities Project Administrator, Real Estate and Office Services Department, to Terry L. Harpster, Chief QA Branch, Division of QA, Safeguards and Inspection Programs, IE, USNRC. These responses have also been accepted by the NRC in their letter dated 04/22/83 from Walter P. Haass, Deputy Chief, QA Branch, Division of Quality Assurance, Safeguards, and Inspections Programs, Office of Inspection and Enforcement.

In lieu of the storage facility requirements of ASMENQA-l-1994 Supplement 17S-I Section 4.4, CCNPP allows the following alternative storage requirements for organizations other than the records management organization. Organizations that originate records and do not transfer them to the records management within 30 days of completion shall establish one of the following three controls as alternatives to the requirements specified for the records management organization:

1. Duplicate Storage - Either A or B.

A. Within 30 days of completion of a record, a duplicate record file shall be established. This activity shall be controlled by procedures which provide for the following:

a. Assignment of responsibility for records.

b. Description of storage area.

c. Description of filing system.

d. An index of the filing system.

e. Rules governing access to and control of files.

f. Methods for maintaining control of and accountability for records removed from the file.

g. Method for filing supplemental information and disposing of superseded or obsolete records.

h. Method for preserving records to prevent deterioration.

i. Method for maintaining specially processed records that are sensitive to light, pressure, or QATR Revision 0 Page8 of 17

temperature.

j. Transfer of duplicates to the records management organization within two years of completion of records.

B. Make arrangements with at least one other department that receives a copy of each document to subject this other copy to the controls specified above.

2. Fire-resistant Building Storage - Records shall be stored in steel cabinets located in a fire-resistant building or a non-combustible building with a fire suppression system. The procedural controls defined for duplicate storage shall be applied.

3. Non-fire-resistant Building Storage - Within non-fire-resistant facilities, records shall be stored in UL one-hour-minimum fire-mted storage cabinets and be subject to the procedural controls defined for duplicate storage. CCNPP defines a Fire-resistant Building as follows: A facility constructed to resist the initiation or spreading of fire; non-combustible and/or fire-suppressive materials used; building certified as fire-resistant by a person who specializes in the technical field of fire prevention and fire extinguishing.

ASME NQA-1-1994 Supplement 7S-I Section 3 could be interpreted to mean that all requirements of this standard are applicable to all safety-related items or services. CCNPP has two approaches for safety-related and designated non-safety related procurement. Controls established for Basic Component Purchases correspond to the requirements ofASME NQA-1-1994. The extent to which the individual requirements of ASME NQA-1-1994 are applied to Commercial Grade Purchases depends on the nature and scope of the work to be performed and the importance to nuclear safety and the items or services purchased.

QATR Revision 0 Page 9 of 17

ANSI N1OI.4-1972 Quality Assurance for Protective Coatings Applied to Nuclear Facilities Regulatory Guide 1.54 Rev. 0 Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants CCNPP Alternative - Through their response to Generic Letter 98-04 the following applies:

The requirements of IOCFR50 Appendix B are implemented through a design specification, which provides the appropriate technical and quality requirements for the Service Level I (safety related) coatings program which includes ongoing maintenance activities. For CCNPP, Service Level I coatings are subject to the requirements of ANSI N101.2-1972, ANSI N101.4-1972, ANSI N5.9-1967 or ANSI N5.12-1974, and Regulatory Guide (RG) 1.54, dated June 1973. The majority of the exposed surfaces within the containment are coated with materials qualified in accordance with ANSI N101.2 and are applied in accordance with RG 1.54. ANSI N101.4 provides an adequate basis for complying with quality requirements for protective coatings applied to various surfaces.

1. Procurement of Service Level I coatings used for new applications or repair/replacement activities are procured, utilizing plant procedures, from suppliers with a quality assurance program meeting the applicable requirements of 10 CFR Part 50 Appendix B. The applicable technical and quality requirements that the supplier is required to meet are specified by CCNPP in procurement documents.

Acceptance activities are conducted in accordance with procedures that are consistent with ASME NQA-1-1994 Part I requirements (e.g., receipt inspection, source surveillance, etc.). This specification of required technical and quality requirements combined with appropriate acceptance activities provides adequate assurance that the coatings received meet the requirements of the procurement documents.

2. The qualification testing of Service Level I coatings used for new applications or repair/replacement activities inside containment meets the applicable requirements contained in the standards and regulatory commitments referenced above along with meeting the requirements of ANSI NIOI.2-1972, ANSI N5.12-1974 and Regulatory Guide (RG) 1.54.

3. Design specifications and procedures, developed and approved in accordance with IOCFR Appendix B criteria, capture the surface preparation, application, and surveillance during installation of Service Level I coatings used for new applications or repair/replacement activities inside containment in order to meet the applicable portions of the standards and regulatory commitments referenced above.

Coating activities (and documentation of such) within both containments are performed in accordance with plant procedures. Where the requirements of the standards and regulatory commitments did not address or were not applicable to repair/replacement activities, these activities were performed in a manner consistent with the generally accepted industry practices for coatings repair/replacement.

These practices are typically described in various ASTM standards and coating practice guidelines by industry organizations issued subsequent to those to which CCNPP has a regulatory commitment.

GINNA Alternative - For new coatings and configuration changes to existing coatings, that have the potential to adversely affect a safety related function, the quality assurance requirements of IOCFR5O, Appendix B, in conjunction with engineering specifications, are used instead of the detailed requirements included in this Regulatory Guide and its referenced standard, ANSI NIO1.4-1972.

NMPI For NMP I, Service Level I coatings are subject to the requirements of ANSI NI 01.4. Adequate assurance that the applicable requirements for the procurement, application, inspection, and maintenance are implemented is provided by procedures and programmatic controls, approved under this Quality Assurance program.

Acceptance activities are conducted in accordance with procedures which are consistent with ASME NQA-1-1994 Part I requirements (e.g., receipt inspection, source surveillance). This specification of required and technical requirements combined with appropriate acceptance activities provides adequate assurance that the coatings received meet the requirements of the procurement documents.

NMP2 The requirements of IOCFR50 Appendix B are implemented through Specification NMP2-S401K, which provides the appropriate technical and quality requirements for the Service Level I (safety related) coatings program which includes ongoing maintenance activities. For NMP2, Service Level I coatings are subject to the requirements of ANSI N101.2, ANSI NIOI.4, ANSI N5.12, and Regulatory QATR Revision 0 Page IO of 17

Guide (RG) 1.54. The majority of the exposed surfaces within the drywell (i.e., primary containment liner, drywell head, biological shield wvall, pipe rupture restraints, pipe supports, piping, and concrete) are coated with materials qualified in accordance (with) ANSI NO1I.2 and are applied in accordance with RG 1.54. ANSI N101.4 provides an adequate basis for complying with quality requirements for protective coatings applied to various surfaces.

Acceptance activities are conducted in accordance with Specification NMP2-S40I K This specification of required and technical requirements combined with appropriate acceptance activities provides adequate assurance that the coatings received meet the requirements of the procurement documents.

NMPNS The requirements of I OCFR50 Appendix B are implemented through specification of appropriate technical and quality requirements for the Service Level I (safety related) coatings program which includes ongoing maintenance activities. Service Level I coatings for new applications or repair/replacement activities are procured Safety Related from suppliers with a Quality Assurance program meeting the applicable requirements of IOCFR50 Appendix B. The applicable technical and quality requirements which the supplier is required to meet are specified in procurement documents.

The qualification testing of Service Level I coatings used for new applications or repair/replacement activities inside the primary containment meets the applicable requirements contained in the standards and regulatory commitments referenced below. The surface preparation, application and surveillance during installation of Service Level I coatings used for new applications or repair/replacement activities inside the primary containment meet the applicable portions of the standards and regulatory commitments referenced below. Documentation of the completion of these activities is performed consistent with the applicable requirements.

QATR Revision 0 Page I I Of 17

ANSI N18.1-1971 Selection and Training of Nuclear Power Plant Personnel Regulatory Guide 1.8 Rev. 1 Personnel Selection and Training CGG CCNPP, Ginna, and Nine Mile Point Unit I conform to Regulatory Guide 1.8 Revision I and its endorsement ofANSI N18.1-1971, with the listed exceptions. Nine Mile Point Unit 2 conforms to ANSI/ANS-3.1-1978 as described in this section.

CCNPP In lieu of the requirements of ANSI N18.1-1971 Paragraphs 3.2.1 and 4.2.2, neither the Manager-Nuclear Operations or the General Supervisor-Nuclear Plant Operations needs to individually meet all of the requirements of both paragraphs 3.2.1 and 4.2.2. The Manager-Nuclear Operations will satisfy paragraph 3.2.1 and most of 4.2.2, except that he will not maintain a SRO license. Instead, the Manager-Nuclear Operations will hold or have held a SRO license. The General Supervisor-Nuclear Plant Operations will hold and maintain a SRO license. The General Supervisor-Nuclear Plant Operations satisfies paragraph 4.2.2, but he does not satisfy 3.2.1 because he does not report directly to the plant manager.

In lieu of the requirements of ANSI N18.1-1971 Paragraphs 3.2.2 and 4.3.2, CCNPP has three supervisory positions in its organization (Supervisors, and in some cases Assistant General Supervisors and General Supervisors) that are organizationally equivalent, when supervising technicians/repairmen, to the positions described in Paragraph 3.2.2. All of these individuals do not need to possess the four years of craft/discipline experience required by Paragraph 4.3.2. Instead, at least the first line supervisor shall possess four years experience in the craft/discipline he supervises.

The other supervisors in the organization may be selected to fill supervisory positions based on possessing a minimum of an Associate's Degree, with four years of related technical experience, and demonstrated supervisory ability. Additionally, all first line and intermediate supervisors shall have at least a high school diploma or equivalent.

GINNA In lieu of the SRO license requirements of ANSI N18.1-1971 Section 4.2.2, the operations manager or operations middle manager shall hold a SRO license.

NMPI Each member of the Nine Mile Point Unit I staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for, the Manager Operations who, in lieu of meeting the SRO license requirements of ANSI N18.1-1971, shall 1) hold an SRO license at the time of appointment, or 2) have held an SRO license a NMPNS Unit I or at a similar unit, or 3) have been certified for equivalent SRO knowledge; and the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

As a minimum, either the Manager Operations or the General Supervisor Operations shall hold an SRO license.

NMP2 In lieu of ANSI N18.1-1971, each member of the Nine Mile Point Unit 2 staff shall meet or exceed the minimum qualifications of ANSI/ANS-3.1-1978 for comparable positions, except for the radiation protection manager who shall meet or exceed the qualification of Regulatory Guide 1.8 Revision 1.

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ANSI N45.2.1-1973 Cleaning ofFluidSystems andAssociated Components During Construction Phase of Nuclear Powver Plants Regulatory Guide 1.37 Rev. 0 Quality Assurance Requirementsfor Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Poiter Plants CGG CGG conforms to Regulatory Guide 1.37 Revision 0 and its endorsement of ANSI N45.2.1-1973, with the following exceptions:

Regulatory Guide 1.37 Revision 0 Paragraph C.3 requires the water quality for final flushes...shall be at least equivalent to the quality of the operating system water. CGG is committed to the stated requirement except for the oxygen content.

ANSI N45.2.1-1973 Section 3.1.2 gives the requirements for Class B cleanness. Upgraded piping systems and components constructed of carbon steel materials will meet Class B cleanliness requirements, except for final flushing/cleaning, which may exhibit rust staining in accordance with Class C cleanness requirements.

CCNPP In lieu of the requirements for demineralized water in ANSI N45.2.1-1973 Section 3.2, CCNPP specifications for demineralized water are consistent with guidelines provided by the Nuclear Steam Supply System supplier. CCNPP specifications are generally more restrictive than those specified by ANSI N45.2.1-1973.

GINNA For new construction activities, the cleanliness requirements of ANSI N45.2.1-1973 as modified by Regulatory Guide 1.37 Revision 0 are followed. Consistent with Position C.2 of Regulatory Guide 1.37 Revision 0, the cleanliness requirements of this standard are used when applicable to maintenance on operating systems. The cleanliness requirements applied to operational systems are established in station procedures.

NMPNS Applicable parts of Regulatory Guide 1.37 Revision 0 will be applied to nuclear safety-related activities pertaining to major plant modifications and maintenance within the scope of Section 5.2.10 of ANSI NI 8.7-1976/ANS-3.2 only (i.e., those under the direction of Nuclear Engineering).

Regulatory Guide 1.37 Revision 0 Paragraph C.4 gives precautions on chemical compounds that could contribute to intergrannular cracking or stress corrosion cracking in austenitic stainless steels and nickel-based alloys. At NMPNS expendable materials, i.e., inks and related products, temperature indicating sticks, tapes, gummed labels, wrapping materials (other than polyethylene), water soluble dam materials, lubricants, NDT penetrant materials, and couplants that contact stainless steel or nickel allow surfaces, are in accordance with the Nine Mile Point Unit 2 USAR position for Regulatory Guide 1.38 Revision 2.

ANSI N45.2.1-1973 Section 3.2 gives water quality requirements for cleaning. Due to seasonal conditions, freshwater from Lake Ontario will have an allowable upper pH limit of 8.5 QATR Revision 0 Page 13 of 17

ANSI N45.2.2-1972 Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase)

Regulatory Guide 1.38 Rev. 2 Quality Assurance Requirementsfor Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants CGG Ginna conforms to Regulatory Guide 1.38 Revision 2 and its endorsement of ANSI N45.2.2-1972. In lieu of Regulatory Guide 1.38 Revision 2 and ANSI N45.2.2-1972, NMPNS is committed to ANSI/ASME NQA-2-1983, Part 2.2 for nuclear safety-related permanent plant modifications and maintenance activities. CCNPP conforms to Regulatory Guide 1.38 Revision 2 and its endorsement of ANSI N45.2.2-1972, with the following exceptions:

CCNPP In lieu of the requirements of ANSI N45.2.2-1972 Section 2.4, CCNPP requires that only persons who are responsible for approving items for acceptance shall be qualified in accordance with ASME NQA-1-1994 Supplement 2S-I and that personnel who verify that storage areas meet requirements will be qualified to ASME NQA-1-1994 Supplement 2S-1 or 2S-3.

In lieu of the requirements of the second sentence of ANSI N45.2.2-1972 Section 2.4, CCNPP uses personnel qualified in accordance with ASME NQA-1-1994 Supplement 2S-3 to perform auditing and monitoring functions.

In lieu of the requirements of ANSI N45.2.2-1972 Section 2.7, The level of protective measures defined by Section 2.7 are applied to Basic Component purchases. Personnel of CCNPP-Engineering Services or Procurement and Warehouse Services, will determine the level of protective measures to be applied to Commercial Grade purchases.

In lieu of the requirements of ANSI N45.2.2-1972 Section 3, CCNPP has replaced Section 3 with the following:

1. Packaging for Shipment to CCNPP - Personnel of CCNPP-Engineering Services or Procurement &

Warehouse Services shall ensure that procurement documents for Basic Component and Commercial Grade item purchases either indicate that the normal methods of packaging and shipment used by industry in general are acceptable for the items being procured or specify the level of protection assigned to the item and the requirement that the vendor conform to applicable requirements for items in that classification defined in Regulatory Guide 1.38 Revision 2.

2. The normal methods of packaging used by the industry in general are acceptable for items being procured as Commercial Grade.

3. Packaging for Storage by CCNPP - In general the packaging used by the vendor to ship items for all types of purchases to CCNPP need not be retained after the item is received by CCNPP, provided that the item is stored in an area that meets the requirements for a storage area for the level of protection assigned to the item. Special or unique items, however, may require special protective measures. For such unusual items, the Department that initiated the purchase, together with CCNPP-Engineering Services or Procurement & Warehouse Services shall identify if any of the requirements of Section 6.4.2 of ANSI N45.2.2-1972 apply.

In lieu of the requirements of ANSI N45.2.2-1972 Section 4, CCNPP has replaced Section 4 with the following:

1. Shipping to CCNPP - CCNPP will invoke the requirements for shipping specified in Section 4 of ANSI N45.2.2-1972 on Basic Component purchases only when CCNPP-Engineering Services or Procurement & Warehouse Services personnel have specified in procurement documents that the item shall be packaged in conformance with ANSI N45.2.2-1972, Section 3.8. CCNPP will not invoke the requirements of ANSI N45.2.2-1972 Section 4, on Commercial Grade item purchases.

2. Shipping from CCNPP - Items shipped from CCNPP need not conform to any of the requirements of ANSI N45.2.2-1972, but the organization that packs and handles the item shall provide roughly the same level of protection that the item was given during shipment to CCNPP.

In lieu of the requirements of ANSI N45.2.2-1972 Section 6.4, CCNPP does not require items to be QATR Revision 0 Page 14 of17

stored in the packing used for shipment if the storage level in the area provides the same protection as the level of packing assigned to the items. Caps, covers, etc. will be required only if specified by CCNPP-Engineering Services or Procurement & Warehouse Services personnel during the procurement process. If an item is taken from one storage area to another, however, the persons who move it are responsible for ensuring, as applicable, that additional packing is supplied to give adequate protection during transportation.

In lieu of the requirements of ANSI N45.2.2-1972 Section 7.3.3, CCNPP controls for the use of hoisting equipment are compatible with the Standards listed in Subsection 7.3.3 of ANSI N45.2.2-1972, although at the discretion of the Plant General Manager, they need not be compatible with documents referred to in these documents.

ANSI N45.2.3-1973 Housekeeping During the Construction Phase of Nuclear Power Plants Regulatory Guide 1.39 Rev. 0 Housekeeping Requirements for Water-Cooled Nuclear Power Plants CGG In lieu of Regulatory Guide 1.39 Revision 0, Ginna conforms to Regulatory Guide 1.39 Revision 2. In lieu of Regulatory Guide 1.39 Revision 0 and ANSI N45.2.3-1973, NMPNS conforms to ANSI/ASME NQA-2-1983, Part 2.3. CCNPP conforms to Regulatory Guide 1.39 Revision 0 and its endorsement of ANSI N45.2.3-1973, with the following exceptions:

CCNPP In lieu of the requirements of ANSI N45.2.3-1973 Section 2.1, CCNPP has established three classes for cleanliness requirements. There is no class equivalent to the ANSI Zone 1. Requirements of ANSI Zones 4 and 5 have been consolidated into CCNPP's class 3. Where required, smoking restrictions are posted for CCNPP's class 3 areas.

In lieu of the requirements of ANSI N45.2.3-1973 Section 2.1, CCNPP has established the following methods for personnel and material accountability:

1. Written accountability.

2. Where possible tethering of tools and materials to permanent plant structures or persons.

3. Post-maintenance close-out inspections.

ANSI N45.2.4-1972/IEEE 336-1971 Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations Regulatory Guide 1.30 Rev. 0 Quality Assurance Requirementsfor the Installation, Inspection, and Testing of Instrumentation and Electric Equipment CGG CGG conforms to Regulatory Guide 1.30 Revision 0 and its endorsement of ANSI N45.2.4-1972/IEEE 336-1971, with the following exceptions:

In lieu of the requirements of the last paragraph of ANSI N45.2.4-1972 Section 6.2.1, the calibration program at CGG does not use calibration stickers on installed plant instrumentation that contain the date of calibration and identity of person that performed the calibration. Calibrations of instruments are scheduled and tracked by a computer database.

QATR Revision 0 Page 15 of 17

ANSI N45.2.5-1974 Supplementary QA Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants Regulatory Guide 1.94 Rev. I Quality Assurance Installation, Inspections, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants CGG Not applicable to Ginna. This Regulatory Guide applies to plants in the construction phase and was issued after Ginna was built. In lieu of Regulatory Guide 1.94 Revision I and ANSI N45.2.5-1974, CCNPP conforms to ANSI N45.2.5 Draft 3, Revision I (l 1/3). NMPNS conforms to Regulatory Guide 1.94 Revision I and its endorsement of ANSI N45.2.5-1974, with the following exceptions:

NMPNS Applicable parts of ANSI N45.2.5-1974 will be applied to nuclear safety-related activities pertaining to major plant modifications only (i.e., those under the direction of Nuclear Engineering)

In lieu of the requirements for assembly and erection under Section 5.3, bolt holes generally will not be burned (oxygen cut). If holes must be burned, the following criteria will be followed: (a) after cutting, the edges of the cut will be ground or reamed back a minimum of 1/32 in, and (b) the final bolt hole dimensions will not exceed those given in the Specification for Structural Joints using ASTM A325 or A490 bolts.

In lieu of the criterion for determining correct bolt length as at least two threads extending beyond the face of the nut under Section 5.4, the criterion for NMPNS is one thread extending beyond the face of the nut.

In lieu of the requirements for welding inspection under Section 5.5, all reinforcing bar splices made by arc welding, except those splices welded to metal embedments, will be selected on a random basis for radiography and inspected in accordance with AWS D12.1. Splices welded to metal embedments will be inspected in accordance with AWS D12.1.

In lieu of the requirements for evaluating mechanical (Caldweld) splice test results under Section 6.2.2, exceptions regarding mechanical splicing of safety-related reinforcing bars are in accordance with the Nine Mile Point Unit 2 USAR position for Regulatory Guide 1.1 0 (superseded by Regulatory Guide 1.136).

ANSI N45.2.8-1975 Supplementary QA Requirements for Installation, Inspection, and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants Regulatory Guide 1.116 Rev. 0 Quality Assurance Requirements for Installation, Inspections, and Testing of Mechanical Equipment and Systems CGG Ginna Conforms to Regulatory Guide 1.116 Revision 0 and its endorsement of ANSI N45.2.8-1975.

In lieu of Regulatory Guide 1.116 Revision 0 and ANSI N45.2.8-1975, CCNPP conforms to ANSI N45.2.8 Draft 3, Revision 2 (09/73). In lieu of Regulatory Guide 1.116 Revision 0 and ANSI N45.2.8-1975, NMPNS is committed to ANSI/ASME NQA-2-1983, Part 2.8, for nuclear safety-related permanent plant modifications and maintenance activities.

QATR Revision 0 Page 16 of 17

ANSI/ANS 51.1-1983 Nuclear Safety Criterlafor the Design of Stationary Pressurized Water Reactor Plants Regulatory Guide 1.26 Rev. 3 Quality Group Classif cations and Standardsfor Water-, Steam-, and Radioactive-Wvaste-Containing Components of Nuclear Power Plants CCNPP Alternative - Calvert Cliff's QA Program is applied to structures, systems, components, and activities that have been designated SR because they prevent accidents or mitigate the consequences of postulated accidents that could cause undue risk to the health or safety of the public. The QA Program is also applicable to designated NSR structures, systems, components, activities, and services as required by regulations. Designated NSR program requirements are based on a graded approach to Quality Assurance required to meet applicable regulatory designated requirements and guidance. The level of QA Program controls placed on designated NSR items are defined in QA Program documents and/or implementing procedures. The controls from other sections of this QA Policy are selected as necessary to meet the particular regulations being implemented. Controls have been established for specifying on a Quality List (Q-List) all SR structures, systems, components, and activities that are subject to the requirements of the QA Program.

GINNA Alternative - A classification process is established to identify SSCs that are safety related (SR), safety significant (SS), or Non-Nuclear Safety (NS). Criteria are based on information contained in the Updated Final Safety Analysis Report (UFSAR), licensing commitments, guidelines contained in NRC regulatory guides, and functional guidance derived from ANSI/ANS 51.1-1983. For changes to Ginna Station, safety classification and corresponding QA program applicability are determined using approved procedures.

The Nuclear Policy Manual provides a method of applying a graded QA Program to systems, components, items, and services which are not classified as safety related (SR), but are considered necessary for reliable plant operation.

NMPNS Alternative - Safety-related structures, systems, and components are identified in Q-Lists, which are developed and maintained for each Unit, and are consistent with the SAR commitments. Appropriate elements of the QATR are applied to the Fire Protection Program, emergency plans, radiation protection procedures and radioactive waste shipment programs for the Station.

The scope of the program and the extent to which its controls are applied are established as follows:

1. NMPNS uses the criteria specified in Nuclear Implementing Procedures for identifying structures, systems and components to which the QA Program applies.

2. This identification process results in a Q-List and Safety Classification Determinations, which identify safety-related items for each Unit. The Q-List is a controlled document. Safety-related items are determined by an engineering analysis of the function(s) of plant structures, systems and components in relation to safe operation and shutdown.

3. The controls specified in the QA Program described in this QATR are applied to safety-related items, and others as specified by NMPNS.

EPRI NP-5652 (NCIG-07)

Guidelinefor the Utilization of Commercial Grade Items in Safety-Related Applications (NCIG-07)

Generic Letter 89-02 Actions to Improve the Detection of Counterfeit and Fraudulently Marketed Products CGG GINNA conforms to Generic Letter 89-02, and its endorsement of EPRI NP-5652 (NCIG-07).

NMPNS and CCNPP reference EPRI NP-5652 and the supplement thereof as basis documents in procurement procedures. There is no formal commitment to GL 89-02.

QATR Revision 0 Page 17 of 17

ATTACHMENT (2)

QA PROGRAM COMPARISON MATRIX Constellation Generation Group March 29, 2005

i}, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

Current QAP Section / (paragraph)**

CGG QATR Section / (paragraph)

Change Discussion..

R/N/I Basis for

'(Note I)

',Reduction in-'

Comm'itmen'ft Meeting.

____A ppendix B -

L.B; L Organization and

-Responfsibilities_____

1) All levels of organization..

A.2. I-A.2.4 Note 2 N

2) Middle level management is..

A.2.5 Note 3 N

3) First line craft and non-craft..

A.3(1) All employees of CGG Note 3 N

4) Adherence to procedures is..

A.3 (4) Adherence to procedures...

Verbatim

-N

5) Procedures may be deviated..

A.3 (5) Procedures may be deviated Verbatim N

6) The Corporate Organization..

A.2 Note 4 N

7) The management team listed A.1(1) The President of...

Note 3 N

8) Reporting to the Vice Presidents...

A.2 Note 4 N.

9) Vendors, contractors or..

B.4 (I and 8.4) A.3(1 and 3)

Note 2 N

10) Two advisory groups perform...

A.2(6.2 and 6.3 on page 6) (NSRB Note 2 N

and PORC)

11) CCNPP's QA Program for A.2.1 Note 2 N

12) The Senior Vice President-A.2.2 Note 2 N

13) Quality Assurance matters that C.2(7) If a difference of opinion...

Note 2 N

14) The Vice-President-CCNPP is...

A.2.3 Note 2 N

15) The Plant General Manager is...

A.2.5.a Note 3 N

16) The Manager-Nuclear Operations Not included Note 4 N

17) The Manager-Nuclear Not included Note 4 N

Maintenance Presidens...4

18) The Manager-Inrtegrated Work Not included Note 4 N

Management

19) The Manager-Nuclear Training..

A.2.5.c Note 3 N

20) The Vice-President-NTS Not included Note 4 N

21) The Managcr-CCNPP-ES, is...

A.2.5.g Note 3 N

22) The Manager-Fleert LicNot included Note 4 N

23) -The Director-Fleet Fuels Not included Note 4 N

24) The Director-Fleet PRA Services Not included Note 4 N

Page I of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

25) The Director-Fleet Policies and Not included Note 4 N

Procedures...

26) The Manager-Q&PA is...

A.2.4 Note 3 N

27) 2. Ensuring that QA compliance A.2.4, B.14(3)

Note 3 N

28) 3. Taking necessary corrective...

A.4(2) Nuclear operations and Note 3 N

support

29) CCNPP has established that the A.2.4, TableA.7.2 (Reg Guide 1.8, Note 3 N

Manager...

Rev 1)

30) CCNPP personnel who report...

C.2.(1) A program of planned...

Note 3 N

31) The General Manager-Fleet Not included Note 4 N

Production Operations...

32) The Manager-Projects is...

A.2.5.b Note 3 N

33) The Manager-Supply Chain...

A.2.5.h Note 3 N

34) The Manager-Not included Note 4 N

Security/Emergency Preparedness...

35) The Manager-CGG-IT...

Not included Note 4 N

I.B.2; Quality Assurance Pro' r

1) The QA program consists of...

A.1 Note 3. QATR does not discuss N

revision process.

2)The QA Policy identifies NRC...

A.1 (4) The QA Program has also...

Note 3 N

3) Nuclear Program Directives...

A.1(5) Nuclear directives address Note 3 N

4) Calvert Cliff's QA Program is...

Table A.7.2, page 15, R.G. 1.26 Note 2 N

5) Controls have been established...

Table A.7.2, page 15, R.G. 1.26 Note 2 N

6) The Statement of Authority, in...

A.I(1) The President of...

Note 3 N

7) The QA Policy is approved by...

A.I(l) The President of...

Note 3 N

8) The QA Policy has established...

A.1(4) The QA Program has also...

Note 3 N

9) The QA Program has also...

A.1(4) The QA Program has also...

Note 2 N

10) Nuclear Program A.2.4 Note 3. The first sentence is not N

Managers/Directors ensure...

covered.

11) The Sr. VP-Nuclear Operations C.2(9) The effectiveness of Note 3 N

12) If a difference of opinion arises C.2(7) If a difference of opinion...

Verbatim N

13) To ensure that important...

A.5 Note 2 N

14) The QA Program was developed A.7 Note 2 N

15) (list of regulations, Reg. Guides, Table A.7.2 Note 2 N

and standards)

Page 2 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

16) The QA Policy and revisions...

B.14(3.5) Procedures that...

Note 3.

N

17) Nuclear Program Directives...

A.1(5) Nuclear directives address...

Note 3. QATR does not have QA N

review of directives.

18) QA Program documents ensure B.I(1) The QA Program provides...

Note 3 N

19) Procedures require that CCNPP A.2(6) Procedures require that...

Verbatim N

20) The maintenance and repair of...

B.1 Note 3 N

21) When necessary, non-plant...

B.1 Note 3 N

22) Controls are established in QA...

B.4 and B.7 Note 3 N

23) Equipment manufacturers and...

B.4(2) CGG qualifies suppliers...

Note 2 N

I.B.3, Design Control I) Facility changes are controlled...

B.2(1) Station modifications are...

Note 3 N

2) Controls of facility changes, tests, B.2(1) Station modifications are...

Note 3. Does not discuss N

and...

Modification Evaluations or Equivalency Change Evaluations.

3) The process for controlling facility B.2(2)

Note 3 N

changes...

Addendum A.7.1, section 1.7.e

4) RDO's, either on contract or within

CCNPP,

1.

B.2(4) Design control measures...

Note 3 N

2.

B.2(5) Design inputs (such as...

Note 2 N

3.

B.2(5) Design inputs (such as...

Note 2 N

4.

B.2(4.1) Design inputs are...

Note 3 N

5.

B.2(4.4) Personnel other than...

Note 2 N

6.

B.(4.2) Design interfaces are...

Note 2 N

7.

B.3(2) The extent of and methods...

Note 2 N

8.

B.3(6) Persons representing...

Note 3. QATR allows the design N

supervisor to review in certain cases. Supplement 3S-1 in NQA-1

9.

B.2(2) Configuration changes,...

Note 3 N

10.

B.2(4.3) The design is suitable...

Note 3 N

11.

B.3 Note 3 N

12.

B.2(4.3) The design is suitable...

Note 3 N

13.

B.2(8) Computer programs used...

Note 2 N

1.B.4, Procurement Document Control

1) Controls have been established...

B.4(1) Procurement documents...

Note 2 N

Page 3 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

2) The degree to which these...

B.4(1) Procurement documents...

Note 3 N

3) Qualified PWS personnel...

B.1(2) Personnel performing...

Note 3 N

4) All changes made to procurement..

B.4(l) Procurement documents...

Note 3 N

5) Bids submitted to supply safety-...

B.4(l) Procurement documents...

Note 3 N

6) Personnel in Q&PA evaluate...

B.4(2) CGG qualifies suppliers...

Note 3 N

7) Controls have been established to B.4(3) Procurement of safety-related Note 2 N

ensure that, before placement of a

8) Controls have been established to B.4(12) Appropriate controls are Note 2 N

ensure that items or services...

imposed...

9) Procedures require that...

B.4(8), B.4(12), B.5(2)

Note 3 N

L.B.5i Instructions, Procedures, and D raw ings

1) Controls delineate the sequence...

B.14(l) Documents are controlled..

Note 2 N

2) Controls require that:

B.14(l) Documents are controlled..

Note 3 N

3) Controls ensure that:

B.14(l) Documents are controlled..

Note 3. QATR deletes the concept N

of Control Procedures.

4) The Plant General Manager may..

B.14(4) The Plant General Manager Note 2 N

5) Procedures listed in Technical...

B.14(5) Procedures listed in...

Note 2 N

6) Temporary changes to...

B.14(6) Temporary changes to...

Verbatim N

7) Editorial corrections to...

B.14(2) Changes to controlled...

Note 3 N

8) As used in this document,:

B.14(2) Changes to controlled...

Note 3 N

I.B.6, Document Control I) Requirements have been...

B.14(3) Procedures are established..

Note 3 N

2) Revisions to the QA Policy are...

B.14(l) Documents are controlled..

Note 3 N

3) Alterations to the UFSAR are...

B.14(l) Documents are controlled..

Note 3 N

4) Alterations to the ISFSI updated B.14(l) Documents are controlled..

Note 3 N

SAR...

5) Alterations to the Operating B.14(l) Documents are controlled..

Note 3 N

License...

6) Alterations to the ISFSI Materials B.14(l) Documents are controlled..

Note 3 N

License,

7) Alterations to the Emergency B.14(l) Documents are controlled..

Note 3 N

Response Plan

8) Alterations to the Security Plan...

B.14(l) Documents are controlled..

Note 3 N

9) Directives are required to:

A,1(5) Nuclear directives address...

Note 2 N

Page 4 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

10) Nuclear Program Directives are..

A,1(5) Nuclear directives address...

Note 3 N

11) Administrative and Technical... B

.14(3) Procedures are established..

Note 3. QATR does not discuss N

QA review of Control Procedure

12) Administrative procedures are... B

.14(3) Procedures are established..

Note 3 N

13) During the review of each...

B.14(1) Documents are controlled..

Note 3 N

14) Organizations that issue...

B.14(l), B.14(2)

Note 3 N

15) When changes to drawings or...

B.14(1) Documents are controlled..

Note 3 N

I1.B.7, Control of Purchased Material, Equipment and Services

1) Procurement and Warehouse...

B.l(l) The QA Program provides...

Note 3 N

2) The controls include:

B.4(3) Procurement of safety-related Note 3 N

3) The vendor evaluation is...

B.4(2) CGG qualifies suppliers...

Note 2 N

4) The National Institute of...

B.4(6) The National Institute of...

Note 2 N

5) Q&PA performs evaluations...

B.4(5) CGG qualified suppliers...

Note 2 N

6) Since CCNPP accepts the...

B.5(3) Since CGG accepts the...

Verbatim N

7) A survey may be performed of...

B.5(3) Since CGG accepts the...

Verbatim N

8) Vendor controls evaluated to be...

B.5(3) Since CGG accepts the...

Verbatim N

9) The depth of vendor evaluation...

B.5(3) Since CGG accepts the...

Verbatim N

10) Upon completion of the...

B.4(3) Procurement of safety-...

Note 3 N

11) An auditing organization such...

B.4(2) CGG qualifies suppliers...

Note 2 N

12) When required by operational...

B.4(3) Procurement of safety-...

Note 3 N

13) Vendor surveillance and...

B.4(5) CGG qualified suppliers...

Note 3 N

14) The depth and frequency of...

B.4(5) CGG qualified suppliers...

Note 3 N

15) When a vendor's certificates...

B.5(4) When a supplier's certificates Note 2 N

16) Procurement and Warehouse...

B.5(5) Receipt inspections are...

Note 3 N

17) Additionally, P&WS is...

B.5(5) Receipt inspections are...

Note 3 N

18) Standard receiving inspection...

B.5(5) Receipt inspections are...

Note 3 N

19) Special receiving inspection...

B.5(5) Receipt inspections are...

Note 3 N

20) A written record of the results...

B.5(5) Receipt inspections are...

Note 3 N

21) All SR and designated NSR...

B.5(5) Receipt inspections are...

Note 3 N

22) Non-conforming items are...

B.5(5) Receipt inspections are...

Note 3 N

23) CCNPP procurement...

BA.4(1) Procurement documents...

Note 2 N

24) Vendor inspection records or...

B.13(9) Nonconforming items may Note 3 N

25_ eno_ rq eteevains______________B138_ono fom n be...

__I___

Note 3 N

25) Vendor requested deviations...

IB. 13 (8) Nonconforming items...

INote 3 N

Page 5 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

I.B.8, Identification and Control of Materials, Parts, and Components

1) Procurement and Warehouse...

B.6(1) The identification and control Note 2 N

2) Requirements for identification...

B.6(2) 1. Traceability to...

Note 3 N

3) Assigned Procurement and...

B.6(3) 2. Specification of the...

Note 3 N

4) CCNPP contractors and their...

B.6(l) The identification and...

Note 2 N

5) Identification of items important...

B.6(2) 1. Traceability to...

Note 2 N

6) SR and designated NSR items...

B.5(5) Receipt inspections are...

Note 2 N

7) If an item is found to be or is...

B.5(5) Receipt inspections are...

Note 2 N

8) Acceptance documentation is...

B.6(2) 1. Traceability to...

Note 3 N

9) After completion of tests and...

B.15(l) Organizations performing..

Note 2 N

I.B.9, Control of Special Processes ;.:;_-_--___-;

1) Controls have been established...

B.1 l(1) Special processes are...

Note 3 N

2) CCNPP contractors and their...

B.l 1(1) Special processes are...

Note 3 N

3) Procedures, equipment, and...

B.1 l(1) Special processes are...

Note 3 N

4) These procedures are prepared...

B.1 l(1) Special processes are...

Note 3 N

5) Methods of Nondestructive...

B. 11(2) Processes subject to...

Note 3 N

6) Special processes are performed...

B.1 1(3), B.1 1(6)

Note 3 N

7) Qualification records of...

B.1 1(6) Special process personnel...

Note 3 N

8) The Director-Q&PA provides...

C.2(l) A program of planned and...

Note 2 N

I.B. I 0, Inspection I) Activities that affect the quality...

B.12(1) Inspections required to...

Note 2 N

2) Controls exercised during inspections ensure that:

1.

B.12(l) Inspections required to...

Note 2 N

2.

B.12(6) Inspections are performed... Note 2 N

_3.

B. 1 (3) Personnel performing...

Note 3 N

4.

B.9(3) M&TE is calibrated at...

Note 2 N

5.

B.12(2) Inspections are applied to...

Note 3 N

6.

B.12(2) Inspections are applied to...

Note 2 N

7.

B.1(3) Personnel performing...

Note 3 N

8.

B.12(5) Hold points are...

Note 3 N

9.

B.1 1(5) Personnel with expertise...

Note 3 N

10.

B.12(6) Inspections are performed...

Note 3 N

3) Inspection procedures,...

B.12(2) Inspections are applied to...

Note 3 N

Page 6 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

4) The Director-QPA is responsible...

B.12(6) Tnspections are performed... Note 3 N

5) Other inspections are conducted...

B.12(6) Inspections are performed... Note 3 N

6) In-service inspections are B...B.

1 2(2) Inspections are applied to...

Note 3 N

7) In-service inspections and...

B.12(2) Inspections are applied to...

Note 3 N

L.B.A 1,TestControl

1) To demonstrate the ability of SR.

B.8(1) Testing is performed to...

Note 2 N

2) Whenever testing is required to...

B.8(1) Testing is performed to...

Note 2 N

3) CCNPP-Engineering Services, B.8(2) Tests are performed in...

Note 3 N

4) Written test procedures are...

B.8(3) Test procedures and...

Note 3 N

5) Test results are documented and... B.8(l) Testing is performed to...

Note 2 N

6) Results of completed tests on SR..

Addendum A.7.1 Note 3 N

7) Results of testing performed as...

B.5(5) Receipt inspections are...

Note 2 N

LB.12, Control of Measuring and Test -.

Equipment

1) Calibration controls have been...

B.9(l) M&TE is identified,...

Note 2 N

2) Calibration controls require...

B.9(l) M&TE is identified,...

Note 2 N

3) Written procedures are prepared... B.9(l) M&TE is identified,...

Note 2 N

4) Manufacturer's procedures are...

B.9(1) M&TE is identified,...

Note 3 N

5) Measuring and test equipment B.9(2), B.9(3)

Note 2 N

6) When equipment is found out...

B.9(6) When M&TE used for...

Note 2 N

7) Test and measuring equipment...

B.9(l) M&TE is identified,...

Note 3. "Segregation" in QAP N

inferred from "controlled" in QATR.

8) The status of each item...

B.9(1) M&TE is identified,...

Note 3 N

9) Measuring and test equipment is B.9(l) M&TE is identified,...

Note 3. Details in this QAP N

controlled...

paragraph inferred from "controlled" in QATR

10) Measuring and test equipment...

B.9(3) M&TE is calibrated at...

Note 2 N

I.B.13, Handling, Storage, and;:-,-

r Shipping

.~

I) Appropriate and special...

B.7(i) Activities with the...

Note 2 N

2) Procedures have been...

B.7(2) Procedures are used to...

Note 2 N

3) Special handling, preservation, B.7(2) Procedures are used to...

Note 3 N

4) Controls have been established...

B.7(3) Controls have been...

Note 2 N

I.B.14, Inspection, Test, and Page 7 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

0 erating Status l_-_'

-l

1) Controls have been established...

B.I0(l) Systems are established...

Note 2 N

2) Procedures/instructions are...

B.10(3) The sequence of...

Note 2 N

3) Senior shift personnel are 13...

B.10(3) The sequence of...

Note 3 N

4) The Manager-Q&PA is...

C.2(1) A program of planned and... Note 3. QATR does not specify N

"senior shift personnel" for tagging.

5) The bypassing of required...

B.10(4), B.10(1)

Note 2 N

I.B. 15, Nonconforming Materials, Parts, or Components

1) Controls have been established...

B.13(5) Material, parts, Note 2 N

2) Issues affecting nuclear plant items B.13(5) Material, parts, Note 3 N

are...

3) Any individual identifying...

B.13(7) Any individual identifying Note 2 N

4) Nonconforming items are...

B.13(7) Any individual identifying Note 2 N

5) Nonconformance control...

B.13(7) Any individual identifying Note 2 N

6) Nonconforming items are...

B.13(8) Nonconforming items Verbatim N

7) Reworked, repaired, and...

B.13(8) Nonconforming items are Verbatim N

8) Nonconforming items may be...

B.13(9) Nonconforming items Verbatim N

I.B.16, Corrective Actions -'

1) Controls have been established...

B.13(1) Controls have been...

Note 2 N

2) Corrective actions are...

B.13(2) For significant...

Note 3 N

3) Issues identified as potentially...

B.13(1 and 2)

Note 3 N

Addendum A.7.1, section 1.7

4) Corrective action verification is B.13(3) Corrective action Note 3. Does not require N

performed...

verification...

verification by individuals not involved with the corrective actions. NRC-endorsed NQA-1 does not require independent verification.

5) Significant issues require a root...

B.3.13(2) For significant conditions...

Note 2 N

6) Issues are periodically analyzed...

B.13(4) Issues are periodically...

Verbatim N

1.B.17, Quality Assuri'nce Records

1) Controls have been established...

B..15(1) Organizations performing..

Note 3 N

2) Quality assurance records are...

B..15(2) Records to be controlled...

Note 3 N

Page 8 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

3) Lifetime records, maintained for...

B.15(2) Records to be controlled...

Note 3 N

4) The following records shall be...

B.15(2) Records to be controlled...

Note 3 N

5) Non-permanent records, which...

B.15(3) Requirements and...

Note 3 N

6) The following records shall be...

B.15(3) Requirements and...

Note 3 N

7) Records of radioactive...

B.15(4) Records of radioactive...

Note 2 N

8) Procurement documents specify...

B.4(8) The control of...

Note 3 N

9) Inspection and test records...

B.15(3) Requirements and...

Note 3 N

10) Controls have been provided...

B.15(6) Controls have been...

Note 2 N

1.B.18, Audits -

1) Internal audits are performed by...

C.2(1, 2, 3, 4, and 5)

Note 3 N

2) Vendor audits are performed to...

B.4(5) CGG qualified suppliers...

Note 3N

3) Audits are performed in...

Table A.7.2 Note 3 N

NQA-1, Supplement 18S-1, section 4

4) Reports are analyzed and...

C.2(6) Assessment results are...

Note 2 N

5) To ensure that CCNPP's Q&PA...

C.2(9) The effectiveness of...

Note 2 N

Table lB-I, CCNPP Position on Table A.7.2 Note 2 N

Industry Standards and Regulatory Guides CCNPP is adopting an existing R

Note 5 NMPNS exception: ASME NQA-1-1994 Supplement 7S-1 Paragraph 8.2.4 states, "...post installation test requirements and acceptance documentation shall be mutually established by the Purchaser and Supplier." In exercising ultimate responsibility for its QA program, CGG establishes post-installation test requirements, giving due consideration to supplier recommendations.

CCNPP is adopting the existing R

Note 6 NMPNS exception /interpretation Page 9 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

that in lieu of the master surveillance schedule requirements of ANSI Ni 8.7-1976/ANS-3.2 Section 5.2.8, independent master schedules may exist for different programs such as inservice inspection (ISI), pump and valve testing, and Technical Specification surveillance testing.

CCNPP is adopting NMPNS's R

Note 7 interpretation that implementation of ANSI N18.7-1976/ANS-3.2 Section 5.2.9, Plant Security and Visitor Control, is addressed in the respective site security plan.

CCNPP is adopting the NMPNS R

Note 8 exception to Regulatory Guide 1.37 Revision 0. RG 1.37 Paragraph C.3 requires the water quality for final flushes...shall be at least equivalent to the quality of the operating system water. CGG is committed to the stated requirement except for the oxygen content.

CCNPP is adopting the NMPNS R

Note 9 exception to ANSI N45.2.1-1973.

Section 3.1.2 gives the requirements for Class B cleanness. Upgraded piping systems and components constructed of carbon steel materials will meet Class B Page 10 of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

cleanliness requirements, except for final flushing/cleaning, which may exhibit rust staining in accordance with Class C cleanness requirements.

Attachment A-1, Bases for QA Policy Not applicable Note 3. Reduced level of detail.

N Revisions Figure IB-i, Organization Chart Not included Note 4 N

Addendum I B-i, Review Fu'nctionsSection I PORC, -

Addendum-A.7.1 1.1 1.1 Note 2 N

1.2 1.2 Note 3. Minimum is 5 members, N

not 7. Specific qualification areas are generalized. (TVA SER) 1.3 1.3 Note 2 N

1.4 1.4 Note 2 N

1.5 1.5 Note 2 N

1.6 1.6 Note 2 I

1.7 1.7 Note 3. This does not include N

review of Security and EP plans, and the Process Control Program; and it transfers the separate requirement for review of proposed modifications to be part of the 50.59 process. (TVA SER) 1.8 1.8 Note 2 N

1.9 1.9 Note 2 I

Section 2, Qualified Reviewers Addendum A.7. 1 2.1 2.1 Note 2 N

2.2 2.2 Note 2 N

2.3 2.3 Note 2 N

2.4 2.4 Note 2 N

2.5 2.5 Note 2 N

Section 3, NSRB Addendum A.7. 1 -

3.1 3.1 Note 2 N

3.2 3.2 Note 2 N

Page II of 29

, QA Program Comparison Matrix Calvert Cliffs Nuclear Power Plant, Inc.

3.3 3.3 Note 2 N

3.4 3.4 Note 2 N

3.5 3.5 Note 2 N

3.6 3.6 Note 2 N

3.7 3.7 Note 2 N

3.8 3.8 Note 2 N

3.9 3.9 Note 2 N

3.10 3.10 Note 2 N

3.11 3.11 Note 2 N

Notes:

1. R = Reduction in Commitment N = Not a Reduction in Commitment I = Increase in Commitment

2. Wording in the CGG QATR establishes equivalent requirements or commitments. Any wording difference is the result of format needs associated with changing to Standard Review Plan 17.3 format.

3. Methodology details (the "how") in the current program description are not included in the CGG QATR. The intent of the CGG QATR is to establish appropriate/sufficient requirements (the "what") to provide a quality assurance program meeting I OCFR50, Appendix B, but to allow flexibility in how the requirement is met. Therefore, this change in methodology detail (the "how") is not considered to be a reduction in commitment. 10CFR50.54(a)(3) applies.

4. The CGG QATR does not assign responsibilities beyond those listed in Section A.2 and A.3. The responsibilities listed in Revision 59 of the CCNPP QA Policy are documented in lower tier documents.

5. As stated in the exception, the licensee is ultimately responsible for its QA program and makes the final determination for post installation test requirements and acceptance documentation.

6. Independent master schedules for different programs meet the intent of ANSI N18.7-1976/ANS-3.2 Section 5.2.8 to ensure that safety-related structures, systems, and components continue to operate, keeping parameters within normal bounds or will act to put the plant in a safe condition if they exceed normal bounds.

7. This is conservatively called a reduction in commitment because CCNPP is adopting an exception to a standard. Plant security and visitor control requirements have changed significantly since the ANSI standard was published. Site security plans are maintained current with the changing regulatory requirements independent of the QA program.

8. RG 1.37 Revision 0 Paragraph C.3 refers to ANSI N45.2.1-1973 Section 3.2, Water Quality Requirements, for cleaning/flushing. Oxygen content of the cleaning/flushing water does not have any significant effect on cleaning operations.

9. ANSI N45.2.1-1973 Section 3.1.2 gives the requirements for Class B cleanness. Piping systems and components constructed of carbon steel materials upgraded to Class B cleanness will be allowed to exhibit rust staining in accordance with Class C cleanness requirements due to limitations of the original design.

Page 12 of 29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC Current QATR CGG QATR Change Discussion

- R/N/l Basis for

- Section /

Section / (Paragraph)

(Note 1)

Reduction in (Paragraph)

Commitment Meeting

_______________________________________Appendix B:

Introduction N/A Level of detail not required in the QATR N

B.0 Policy, A.1 Notes 2 & 3 N

B.1.1 Policy, A.1 Note 2 N

B.1.2.1 A.2.1-A.2.4 Note 3 N

B.1.2. 1.1 A.2.1-A.2.5 Notes 3 & 4 N

B. 1.2.1.2 B.4 Notes 2 & 3 N

B.2.1 A.1, Policy Note 2 N

B.2.2.1 A.2.1, A.2.2 Note 2.

N B.2.2.2 A.1 Note 2 & 3. Report Format uses SRP 17.3 N

B.2.2.3 A.1 Notes 2 & 3 N

B.2.2.4 A.1 Note 2 N

B.2.2.5 A.3, A.5 Notes 2 & 3 N

B.2.2.6 A.3, A.5 Notes 2 & 3 N

B.2.2.7 A.1, A.2 Note 2. Note 5N B.2.2.8 A.1 Note 2 N

B.2.2.9 A.1, A.3 Notes 2 & 3 N

B.2.2. 10 A.7 Note 2 N

B.2.2.11 A.1 Note 2 N

B.2.2.12 B.1 Notes 2 & 3 N

B.2.2.13 B.2 Note 6 N

B.2.2.14 A.5 Notes 2 & 3 N

B.2.2.15 A.1,C Notes2&3 N

B.2.2.16 Addendum A.7.1, 3.0 Notes 2 & 3. CCNPP's NSRB process was adopted for the Fleet N

B.2.2.17 Addendum A.7.1, 1.0 Notes 2 & 3. TVA SER N

B.2.2.18 Addendum A.7.1, 4.0 Notes 2 & 3. Note 7 N

B.2.2.19 A.6 Notes 2 & 3 N

B.3.1 B.2, B.3 Note 2 N

B.3.2.1 A.2 Note 2 N

B.3.2.2 B.2 Notes 2 & 3 N

B.3.2.3 B.2 Note 2 N

Page 13 of 29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC B.3.2.4 B.2 Notes 2 & 3 N

B.3.2.5 B.2 Notes 2 & 3 N

B.3.2.6 B.2 Notes 2 & 3 N

B.3.2.7 B.3 Notes 2 & 3 N

B.3.2.8 C.2 Notes 2 & 3 N

B.3.2.9 B.3 Notes 2 & 3 N

B.3.2.10 B.3 Note 2 N

B.3.2.11 B.3 Note 2 N

B.3.2.12 B.3 Note 2 N

B.3.2.13 B.2 Notes 2 & 3 N

B.3.2.14 B.2 Notes 2 & 3 N

B.3.2.15 NA Note 5 N

B.4.1 B.4, B.5 Note 2 N

B.4.2.1 A.2 Note 2 N

B.4.2.2 B.4 Notes 2 & 3 N

B.4.2.3 NA Level of detail not required in QATR N

B.4.2.4 B.4 Notes 2 & 3 N

B.4.2.5 B.4 Note 2 N

B.4.2.6 B.4 Notes 2 & 3 N

B.4.2.7 B.4 Notes 2 & 3 N

B.4.2.8 B.14 Notes 2 & 3 N

B.4.2.9 NA Note 5 N

B.5.1 B.14 Note 2 N

B.5.2.1 A.2 Note 2 N

B.5.2.2 B.14 Notes 2 & 3 N

B.5.2.3 A.1, A.3 Notes 2 & 3 N

B.5.2.4 B.14 Note 2 N

B.5.2.5 NA Note 5 N

B.5.2.6 B.14 Notes 2 & 3 N

B.5.2.7 B.14 Notes 2 & 3 N

B.5.2.8 B.14 Note 2 N

B.5.2.9 Addendum A.7.1, 2.0 Notes 2 & 3 N

B.6.1 B.14 Note 2 N

B.6.2.1 A.2 Note 2 N

Page 14 of 29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC B.6.2.2 B.14 Notes 2 & 3 N

B.6.2.3 B.14 Notes 2 & 3 N

B.6.2.4 B.14 Note 2 N

B.6.2.5 B.14 Notes 2 & 3 N

B.6.2.6 B.14 Notes 2 & 3 N

B.7.1 B.4, B.5 Note 2 N

B.7.2.1 A.2 Note 2 N

B.7.2.2 B.4 Notes 2 N

B.7.2.3 B.4, B.5 Notes 2 & 3 N

B.7.2.4 B.5 Note 2 N

B.7.2.5 B.4, B.5 Notes 2 & 3 N

B.7.2.6 B.4 Notes 2 & 3 N

B.7.2.7 B.5 Note 2 N

B.7.2.8 B.5 Note 2 N

B.7.2.9 B.4 Note 2 N

B.7.2.10 NA Note 5 N

B.8.1 B.6 Note 2 N

B.8.2.1 A.2 Note 2 N

B.8.2.2 B.6 Note 2 N

B.8.2.3 B.6 Notes 2 & 3 N

B.8.2.4 B.6 Notes 2 N

B.9.1 B.1 Note 2 N

B.9.2.1 A.2 Note 2 N

B.9.2.2 B.11 Note 2 N

B.9.2.3 B.11 Note 2 N

B.9.2.4 B.1 Note 2 N

B.9.2.5 B.11 Note 2 N

B.9.2.6 B.1 Note 2 N

B.9.2.7 B.15 Notes 2 & 3 N

B.9.2.8 C.2 Notes 2 & 3 N

B.9.2.9 B.15 Notes 2 & 3 N

B.10.1 B.12 Note 2 N

B.10.2.1 A.2 Note 2 N

B.10.2.2 B.12 Note 2 N

Page 15 of 29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC B.10.2.3 A.5 Notes 2 & 3 N

B.10.2.4 B.12 Note 2 N

B.10.2.5 B.12 Note 2 N

B.10.2.6 B.12 Notes 2 & 3 N

B.10.2.7 B.12 Notes 2 & 3 N

B.10.2.8 B.15 Notes 2 & 3 N

B.10.2.9 NA Note 5 N

B.1 1.1 B.8 Note 2 N

B.11.2.1 A.2 Note 2 N

B.11.2.2 B.8 Note 2 N

B.11.2.3 B.8 Note 2 N

B.1 1.2.4 B.8, C.2 Notes 2 & 3 N

B..11.2.5 B.15 Notes2&3 N

B.11.2.6 NA Note 5 N

B.12.1 B.9 Note 2 N

B.12.2.1 A.2 Note 2 N

B.12.2.2 B.9 Notes 2 & 3 N

B.12.2.3 B.9 Note 2 N

B.12.2.4 B.9 Note 2 N

B.12.2.5 B.9 Note 2 N

B.12.2.6 B.9 Note 2 N

B.12.2.7 B.9 Note 2 N

B.12.2.8 B.9 Note 2 N

B.12.2.9 NA Note 5 N

B.13.1 B.7 Note 2 N

B.13.2.1 A.2 Note 2 N

B.13.2.2 B.7 Note 2 N

B.14.1 B.10 Note 2 N

B.14.2.1 A.2 Note 2 N

B.14.2.2 B.10 Note 2 N

B.14.2.3 B.10 Note 2 N

B.14.2.4 B.10 Note 2 N

B.14.2.5 B.10 Note 2 N

B.14.2.6 B.13 Notes 2 & 3 N

Page 16 of29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC B.14.2.7 NA Note 5 N

B.15.1 B.13 Note 2 N

B.15.2.1 A.2 Note 2 N

B.15.2.2 B.13 Note 2 N

B.15.2.3 B.13 Note 2 N

B.15.2.4 B.13 Note 2 N

B.15.2.5 B.13 Note 2 N

B.15.2.6 B.13 Notes 2 & 3 N

B.15.2.7 C.2 Notes 2 & 3 N

B.15.2.8 B.13 Notes 2 & 3 N

B.15.2.9 B.13 Notes 2 & 3 N

B.15.2.10 B.13 Note 2 N

B.15.2.11 B.13 Notes 2 & 3 N

B.15.2.12 B.13 Notes 2 & 3 N

B.15.2.13 B.13 Note 2 N

B.15.2.14 B.13 Note 2 N

B.15.2.15 NA Note 5 N

B.16.1 A.6 Note 2 N

B.16.2.1 A.2 Note 2 N

B.16.2.2 B.13 Notes 2 & 3 N

B.16.2.3 B.13 Note 2 N

B.16.2.4 B.13 Notes 2 & 3 N

B.16.2.5 NA Note 5 N

B.17.1 B.15 Note 2 N

B.17.2.1 A.2 Note 2 N

B.17.2.2 B.15 Notes 2 & 3 N

B.17.2.3 B.15 Notes 2 & 3 N

B.17.2.4 B.15 Notes 2 & 3 N

B.17.2.5 B.15 Note 2 N

B.17.2.6 B.15 Note 2 N

B.17.2.7 B.15 Notes 2 & 3 N

B.17.2.8 B.15 Notes 2 & 3 N

B.18.1 C.2 Note 2 N

B.18.2.1 A.2 Note 2 N

Page 17 of 29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC B.18.2.2 C.2 Notes 2 & 3 N

B.18.2.3 C.2 Notes 2 & 3. Adopted Ginna policy of 90 day grace period.

N B.18.2.4 Addendum 7.2.1, 3.0 Notes 2 & 3 N

B.18.2.5 C.2 Notes 2 & 3 N

B.18.2.6 C.2 Notes 2 & 3 N

B.18.2.7 C.2 Notes 2 & 3 N

B.18.2.8 C.2 Notes 2 & 3. Entergy SER for Lead Auditor Qualification N

B.18.2.9 C.2 Notes 2 & 3 N

B.18.2.10 C.2 Note 2 N

B.18.2.11 C.2 Notes 2 & 3 N

B.18.2.12 C.2 Notes 2 & 3. Adopted Ginna policy for biennial audits of Fire R

Note 8 Protection Program.

Table B.1 NA Level of Detail and references not required in the QATR N

Table B-2, 1 Table A.7.2 Note 2 N

Table B-2, 2 Table A.7.2; Notes 2 & 3. See submittal letter and FENOC SER to reversion to N

RG 1.33 Rev 2 1976 versus 1982 edition of ANSI 18.7/ANS 3.2.

Table B-2,3 Table A.7.2, RG 1.28 See submittal letter and Exelon SER for details committing to NQA-1, N

Rev 3 1994, Part I NMPNS is adopting the existing CCNPP exception that in lieu of the R

Note 9 requalification requirements of ASME NQA-1-1994 Supplement 2S-3 Section 4.2, CGG requalifies Lead Auditors on the basis of the satisfactory performance of one audit, as observed by a qualified Lead Auditor.

Table B-2, 4 Table A.7.2, Notes 2 & 3 N

RG 1.1 16RevO Table B-2, 5 Table A.7.2, RG 1.37 Notes 2 & 3 N

Rev 0 Table B-2, 6 Table A.7.2, RG 1.94 Notes 2 & 3 N

Rev I Table B-2, 7 Table A.7.2, RG 1.30, Notes 2 & 3 N

Rev 0 Table B-2, 8 Table A.7.2, RG 1.28 Notes 2 & 3 N

Rev 3 Table B-2, 9 Table A.7.2, RG 1.37 Notes 2 & 3 N

Page 18 of 29

, QA Program Comparison Matrix Nine Mile Point Nuclear Station, LLC l RevO I

Table B-2, 10 Table A.7.2, RG 1.94, Notes 2 & 3 Rev I Table B-2, 11 Table A.7.2, BTP Notes 2 & 3.

CMEB 9.5-1 NOTES:

1. R = Reduction in Commitment N = Not a Reduction in Commitment I = Increase in Commitment

2. Wording in the CGG QATR establishes equivalent requirements or commitments. Any wording difference is the result of format needs associated with changing to Standard Review Plan 17.3 format.

3. Methodology details (the "how") in the current program description are not included in the CGG QATR. The intent of the CGG QATR is to establish appropriate/sufficient requirements (the "what") to provide a quality assurance program meeting I OCFR50, Appendix B, but to allow flexibility in how the requirement is met. Therefore, this change in methodology detail (the "how") is not considered to be a reduction in commitment. I OCFR50.54(a)(3) applies.

4. The CGG QATR does not assign responsibilities beyond those listed in Section A.2 and A.3. The station specific responsibilities listed in the current NMPNS QATR are documented in unit Safety Analysis Reports and in lower tier documents.

5. Appropriate elements of this QATR are applied to the Fire Protection Program. Commitments to apply specific QA program elements to the Fire Protection Program are contained in the applicable unit's Safety Analysis Report (or will be incorporated as part of the implementation of this program following NRC approval.) Entergy SER.

6. QA program elements associated with computer programs used in safety-related design and Station operation are appropriately applied in accordance with NQA-1, 1994, Part 1.

7. Commitments in the current QATR associated with ISEG functions will be incorporated in the applicable unit Safety Analysis Report as part of the implementation of this program following NRC approval. Entergy SER.

8. In lieu of the 12-month, 24-month and 36-month fire protection and loss prevention audits, CGG will perform a biennial audit of the facility fire protection program and implementing procedures that combines the scope of the three audits into one. The biennial requirement includes an inspection and audit of the fire protection and loss prevention program by non-licensee personnel. The personnel may be representatives of a fire insurance brokerage firm or other qualified individuals. The NRC approved the biennial audit frequency for Ginna in a IOCFR50.54 Quality Assurance Program Change Review dated March 22, 1995.

9. The purpose of the training specified in paragraph 3.2 of the Standard is to ensure that candidates understand the fundamentals of auditing and the requirements for activities to be audited. The fact that persons have not maintained their proficiency does not mean that they need complete re-training; it means only that they have not been able to review and study the applicable Codes, Standards, Procedures, instructions, and other documents related to QA programs and program auditing. CGG considers that the satisfactory performance of an audit under the observation and guidance of a qualified Lead Auditor should ensure that persons with lapsed certification will review and understand the pertinent documents.

Page 19of29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC Current QAPSO CGG QATR-

Change Discussion R/N/I Basis for SecSection

/ (Paragraph)

(Note 1)

Reduction in (Paragraph)

Commitment

-Meeting,~-

'.'Appendix B..

17.1.1(1)

A.l(l)

Note 2 N

17.1.1(2-3)

A.1(2)

Note 2, Note 3 N

17.1.1(4-5)

Table A.7.2 Note 2 N

RG 1.26 17.1.1(6)

N/A Note 3, ASME NQA-I-1994 Part I or industry standard definitions N

apply 17.1.2 The A.2(thru 5.i)

Notes 3 & 4, Entergy, FENOC and TVA SERs N

organization 17.1.2 REVIEW A.2 Review of Note 2 N

AND AUDIT...

Operations 17.1.2 PORC...

Addendum A.7.1, 1.0 Note 3, TVA SER N

17.1.2 NSRB...

Addendum A.7.1, 1.0 Note 3, CCNPP's NSRB process was adopted for the fleet N

17.1.2 DELEG.

A.3(1,3)

Note 2 N

OF WORK...

17.1.3(1)

A.3(1)

Note 2 N

17.1.3(2)

A.l(l,5)

Note 2 N

17.1.3(3-4)

A.3(2-3)

Note 2 N

17.1.4(1-2)

A.4(1-2)

Note 2 N

17.1.5 A.5 Note 3, Reduced site-specific detail - Entergy and FENOC SERs N

17.1.6 A.6(l)

Note 2 N

N/A A.6(2)

A discussion on the Employee Concerns Program was added to the I

QA program 17.1.7(1)

A.7(1)

Note 2 N

17.1.7(2)

Table A.7.2 Note 2 N

90-day grace._._.

Table 17.1.1-1 CGG QA Policy Note 2 N

Table 17.1.1-2 N/A Note 3, ASME NQA-1-1994 Part I or industry standard definitions N

app ly Table 17.1.7-1 Table A.7.2 Added - "In lieu of the SRO license requirements of ANSI N18.1-N RG 1.8 Rev. I RG 1.8 Rev. 1 1971 Section 4.2.2, the operations manager or operations middle Page 20 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC manager shall hold a SRO license." This was copied from Ginna's Technical Specifications.

Table 17.1.7-1 Table A.7.2 Added detail from 17.1.1(4-5) - see above N

RG 1.26 Rev. 3 RG 1.26 Rev. 3 Table 17.1.7-1 Table A.7.2 Note 3, See submittal letter and Exelon SER for details of N

RG 1.28 Rev. 2 RG 1.28 Rev. 3 Committing to RG 1.28 Rev. 3 and NQA-1-1994 Part 1.

Ginna is adopting the existing CCNPP exception that in lieu of the R

Note 5 requalification requirements of ASME NQA-1-1994 Supplement 2S-3 Section 4.2, CGG requalifies Lead Auditors on the basis of the satisfactory performance of one audit, as observed by a qualified Lead Auditor.

Ginna is adopting an existing NMPNS exception: ASME NQA R

Note 6 1994 Supplement 7S-1 Paragraph 8.2.4 states, "...post installation test requirements and acceptance documentation shall be mutually established by the Purchaser and Supplier." In exercising ultimate responsibility for its QA program, CGG establishes post-installation test requirements, giving due consideration to supplier recommendations.

Table 17.1.7-1 Table A.7.2 Added - In lieu of the requirements of the last paragraph of ANSI N

RG 1.30 Rev. 0 RG 1.30 Rev. 0 N45.2.4-1972 Section 6.2.1, the calibration program at CGG does not use calibration stickers on installed plant instrumentation that contain the date of calibration and identity of person that performed the calibration. Calibrations of instruments are scheduled and tracked by a computer database. See the Entergy SER.

Table 17.1.7-1 Table A.7.2 Ginna is committing to a QA Standard approved by the NRC that is N

RG 1.33 Rev. 0 RG 1.33 Rev. 2 more recent than the current QA standard and dropping the patchwork of substituted ANSI criteria.

Ginna currently has a five-year procedure review frequency. ANSI N

N18.7-1976/ANS-3.2 requires a biennial procedure review frequency. CGG is adopting Calvert Cliffs' existing exception to ANSI N18.7-1976/ANS-3.2 for biennial procedure reviews. To mitigate the administrative burden for Ginna associated with the change from a five-year review frequency to the revised biennial Page 21 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC requirement, Ginna plans on a two-year transition period. The two-year transition period is needed to mitigate the administrative burden of Ginna associated with changing the review frequencies and processes upon implementing the CGG QATR. This transition period will also allow equalizing the workload for the new review process.

Ginna is adopting the temporary procedure change exception approved in the Entergy SER.

Ginna is adopting the existing NMPNS exception /interpretation that in lieu of the master surveillance schedule requirements of ANSI N18.7-1976/ANS-3.2 Section 5.2.8, independent master schedules may exist for different programs such as inservice inspection (ISI),

pump and valve testing, and Technical Specification surveillance testing.

Ginna is adopting NMPNS's interpretation that implementation of ANSI N18.7-1976/ANS-3.2 Section 5.2.9, Plant Security and Visitor Control, is addressed in the respective site security plan.

I N

R R

Note 7 Note 8 Table 17.1.7-1 Table A.7.2 No change N

RG 1.36 Rev. O RG 1.36 Rev.

Table 17.1.7-1 Table A.7.2 Ginna is adopting two NMPNS exceptions:

RG 1.37 Rev. 0 RG 1.37 Rev. 0 Regulatory Guide 1.37 Revision 0 Paragraph C.3 requires the water R

Note 9 quality for final flushes...shall be at least equivalent to the quality of the operating system water. CGG is committed to the stated requirement except for the oxygen content.

ANSI N45.2.1-1973 Section 3.1.2 gives the requirements for Class B R

Note 10 cleanness. Upgraded piping systems and components constructed of carbon steel materials will meet Class B cleanliness requirements, except for final flushing/cleaning, which may exhibit rust staining in accordance with Class C cleanness requirements.

Table 17.1.7-1 Table A.7.2 No change N

RG 1.38 Rev. 2 RG 1.38 Rev. 2 Page 22 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC Table 17.1.7-1 Table A.7.2 No change, Ginna conforms to RG 1.39 Rev. 2 N

RG 1.39 Rev. 2 RG 1.39 Rev. 0 Table 17.1.7-1 Table A.7.2 No change N

RG 1.54 Rev. 0 RG 1.54 Rev. 0 Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-1-1994 Part I replaces N

RG 1.58 Rev. 1 RG 1.28 Rev. 3 the commitment to RG 1.58 Rev. 1. Applicable exceptions were retained.

Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-1-1994 Part I replaces N

RG 1.64 Rev. 2 RG 1.28 Rev. 3 the commitment to RG 1.64 Rev. 2 Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-1-1994 Part I replaces N

RG 1.74 Rev. 0 RG 1.28 Rev. 3 the commitment to RG 1.74 Rev. 0 Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-1-1994 Part I replaces N

RG 1.88 Rev. 2 RG 1.28 Rev 3 the commitment to RG 1.88 Rev. 2. The two exceptions are no longer needed.

Table 17.1.7-1 Table A.7.2 No change N

RG 1.94 Rev. I RG 1.94 Rev. I Table 17.1.7-1 Table A.7.2 No change N

RG 1.16 Rev. 0 RG 1.16 Rev.

Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-1-1994 Part I replaces N

RG 1.123 Rev. 1 RG 1.28 Rev. 3 the commitment to RG 1.123 Rev. 1 Table 17.1.7-1 Table A.7.2 No change N

RG 1.143 Rev. I RG 1.143 Rev. 1 Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-1-1994 Part I replaces N

RG 1.144 Rev. I RG 1.28 Rev. 3 the commitment to RG 1.144 Rev. 1. The 90-day grace period to perform triennial supplier audits is retained.

Table 17.1.7-1 Table A.7.2 The commitment to RG 1.28 Rev. 3 and NQA-I -1994 Part I replaces N

RG 1.146 Rev. 0 RG 1.28 Rev. 3 the commitment to RG 1.146 Rev. 0. The lead auditor scoring exception is no longer needed.

Table 17.1.7-1 Table A.7.2 No change N

RG 1.152 Rev. 0 RG 1.152 Rev. 0 Table 17.1.7-1 Table A.7.2 No change N

RG 4.15 Rev. l RG 4.15 Rev. I Table 17.1.7-1 Table A.7.2 No change N

RG 7.10 Rev. 1 RG 7.10 Rev. I Table 17.1.7-1 Table A.7.2 No change N

Page 23 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC IOCFR21 IOCFR21 Table 17.1.7-1 Table A.7.2 Note 3, SRP 17.3 addresses IOCFR50, Appendix A, Criterion I only.

N

IOCFR50, IOCFR50, Appendix Ginna conforms to Criterion 1. For existing systems, Ginna's Appendix A A, Criterion I UFSAR addresses the criteria applied.

Table 17.1.7-1 Table A.7.2 No change N

IOCFR50, 10CFR50, Appendix B Appendix B Table 17.1.7-1 Table A.7.2 No change N

I OCFR50.55a I OCFR50.55a Table 17.1.7-1 Table A.7.2 No change N

IOCFR50.55(e)

IOCFR50.55(e)

Table 17.1.7-1 Table A.7.2 No change N

IOCFR55 IOCFR55 Table 17.1.7-1 Table A.7.2 No change N

Reg. Pos. 2&4 BTP CMEB 9.5-1 Reg.

of BTP CMEB Pos. 2&4 per SRP 9.5-1 9.5.1 Table 17.1.7-1 Table A.7.2 No change N

GL 89-02 &

GL 89-02 EPRI NP5652 Table 17.1.7-1 Addendum A.7.1 Note 3, CCNPP's NSRB process was adopted for the fleet N

ANSI/ANS-3.1-Section 3.3 1987, Sect. 4.7 17.2.1(1-2)

B.1(2-4)

Note 2 N

17.2.1(3)

B.1(4)

Note 2 N

17.2.2(1)

B.2(2)

Note 2 N

17.2.2(2)

B.2(3)

Note 2 N

17.2.2(3-6)

B.2(4-7)

Note 2 N

17.2.2(7)

Addendum A.7.1 Note 2 N

Sections 1.7, 2.1 17.2.2(8)

B.2(9)

Note 2 N

N/A B.2(8)

New NQA-1-1994 Part I requirement I

17.2.3(1)

B.3(1)

Note 2 N

17.2.3(2)

B.3(8)

Note 2 N

17.2.3(3)

B.3(7)

Note 2 N

Page 24 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC 17.2.3(4,7)

B.3(2)

Note 2 N

17.2.3(5-6)

B.3(5-6)

Note 2 N

17.2.4(1)

B.4(7)

Note 2 N

17.2.4(2)

B.4(4)

Note 2 N

17.2.4(3)

A.5, Table A.7.2 Note 2, NQA-1-1994 Part I Basic Requirement 2 and Supplements N

NQA-1 -1994 establish qualification requirements 17.2.4(4)

B.4(7)

Note 2 N

17.2.4(5)

B.4(8)

Note 2 N

17.2.4(6)

B.14(2)

Note 2 N

17.2.4(7)

B.4(1,11,12)

Note 2 N

17.2.4(8)

B.4(2-3)

Note 2 N

17.2.4(9)

B.5(3)

Note 2 N

17.2.4(10)

B.4(5), B.5(2,4)

Note 2 N

17.2.4(11)

B.4(7)

Note 2 N

N/A B.4(6)

NRC letter dated 5/16/79 (Walter P. Haas to Thomas E. Reaves Jr.)

N N/A B.4(9)

Additional detail added N

17.2.5(1)

B.5(2),

Note 2, Note 3 - Grand Gulf SER N

Table A.7.2 RG 1.28 Rev. 3 17.2.5(2)

B.5(2)

Note 2 N

17.2.5(3)

B.4(4)

Note 2 N

17.2.5(4)

B.4(4)

Note 2 N

17.2.5(5)

B.5(5), A.5 Note 2 N

17.2.6(1)

B.6 Note 2 N

17.2.6(2)

B.2(1), B.5(4)

Note 2 N

17.2.7(1)

B.7(1-2)

Note 2 N

17.2.7(2)

B.7(2)

Note 2 N

17.2.7(3)

B.2(l), B.4(8)

Note 2 N

17.2.7(4)

B.7(2)

Note 2 N

N/A B.7(3)

Additional detail added I

17.2.8(1)

B.8(1)

Note 2 N

17.2.8(2)

B.8(3)

Note 2 N

17.2.8(3)

B.8(1-2)

Note 2 N

17.2.8(4)

B.4(1,8)

Note 2 N

17.2.8(5)

B.8(3-4)

Note 2 N

Page 25 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC 17.2.8(6)

B.8(2)

Note 2 N

17.2.9(1)

B.9(1)

Note 2 N

17.2.9(2)

B.9(3)

Note 2 N

17.2.9(3)

B.9(2-4), B.15(2),

Note 2 N

B.4(8) 17.2.9(4)

B.9(6)

Note 2 N

17.2.9(5)

B.14(3)

Note 2 N

17.2.9(6)

B.8(2)

Note 3 N

17.2.10(1)

B.10(3)

Note 2 N

17.2.10(2)

B.10(3)

Note 2 N

17.2.10(3)

B.10(3)

Note 2 N

17.2.10(5)

B.10(3)

Note 2 N

.17.2.10(5)

B. I0(1, 3-4)

Note 2 N

17.2.10(6)

B.8(3)

Note 2 N

17.2.11(l)

B.111-2, 4)

Note 2 N

17.2.11(2)

B.11(2)

Note 2 N

17.2.11(3)

B.4(8)

Note 2 N

17.2.11(4)

B. 1(4,6)

Note 2 N

17.2.12(1)

B.12(1)

Note 3 N

17.2.12(2)

B.12(2,4)

Note 2 N

17.2.12(3)

B.12(5)

Note 2 N

17.2.12(4)

B.12(2)

Note 2 N

17.2.12(5)

B.12(1,6)

Note 2 N

17.2.12(6)

B.4(1,8)

Note 2 N

17.2.13(1-2)

Table A.7.2 Defined in NQA-1-1994 Part I Introduction N

NQA-1-1994 17.2.13(3)

B.13(1)

Note 2 N

17.2.13(4)

B.13(2-3)

Note 2 N

17.2.13(5)

A.4(2)

Note 2 N

17.2.13(6)

B.13(2-3)

Note 2 N

17.2.13(7-8)

B.13(5-6)

Note 2 N

17.2.13(9-10)

B.13(8)

Note 2 N

17.2.13(11)

B.13(4)

Note 2 N

N/A B.13(7)

Additional detail added N

Page 26 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC 17.2.14(1)

B.14(l)

Note 2 N

17.2.14(2)

Table A.7.2 Note 2, Covered by NQA-1-1994 Part 1, Basic Requirement 5 N

NQA-1-1994 17.2.14(3)

B.14(4-5)

Note 3, TVA SER N

Addendum A.7.1 Sections 1.7, 2.1 17.2.14(4)

Table A.7.2 Note 3, Entergy SER N

N18.7-1976 17.2.14(5)

B.14(1-3)

Note 2 N

17.2.14(6)

B.14(3)

Note 2 N

17.2.14(7)

B.4(1,8)

Note 2 N

17.2.14(8)

B.14(1-2)

Note 2 N

N/A B.14(6)

From ANSI N18.7-1976/ANS-3.2 with related exception from Table I

A.7.2 17.2.15(1-3)

B.15(1-3)

Note 2 N

17.2.15(4-8)

Table A.7.2 Note 2, Covered in NQA-1-1994 Part I Supplement 17S-I N

NQA-1-1994 N/A B.15(4)

Additional detail added N

N/A B.15(5)

Applicable to CCNPP only N

N/A B.15(6-8)

Additional detail added N

17.3.1(1)

C.2(l)

Note 2 N

17.3.1(2)

C.1(3)

Note 2 N

17.3.1(3)

C.2(1,3,6)

Note 2 N

17.3.1(4)

Addendum A.7.1 Note 2 N

Section 3.9 17.3.1(5)

C.l(l), C.2(l)

Note 2 N

17.3.2(1l)

C.1(2), A.3 Note 2 N

17.3.2(2)

C.2(l)

Note 2 N

17.3.2(3)

C.2(1-3)

Note 3, Reduced the level of detail N

17.3.2(4)

C.2(2-3)

Note 2 N

17.3.2(5)

C.2(3-4)

Note 2 N

j 17.3.2(6-8)

N/A Note 3, Reduced the level of detail N

17.3.2(9)

C.2(1),

Note 2, NQA-1-1994 Part I Section 3.1 N

Table A.7.2 NQA-1-1994 Page 27 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC 17.3.2(10)

Table A.7.2 Note 2, NQA-1-1994 Part I Sections 3.2 and 4 N

NQA-1-1994 17.3.2(11)

C.2(6),

Note 2, NQA-1-1994 Part I Section 6 N

Table A.7.2 NQA-1-1994 17.3.2(12)

B.4(2-5,7), B.5(2)

Note 2 N

17.3.2(13)

N/A Note 3, Reduced the level of detail N

17.3.2(14)

N/A Note 3, Removed site-specific detail N

Table 17.3.2-1 C.2(4)

Note 2 N

NOTES:

1. R = Reduction in Commitment N = Not a Reduction in Commitment I = Increase in Commitment

2. Wording in the QATR establishes equivalent requirements or commitments.

3. Methodology details (the "how") in the current program description are not included in the CGG QATR. The intent of the CGG QATR is to establish appropriate/sufficient requirements (the "what") to provide a quality assurance program meeting IOCFR50, Appendix B, but to allow flexibility in how the requirement is met. Therefore, this change in methodology detail (the "how") is not considered to be a reduction in commitment. IOCFR150.54(a)(3) applies.

4. The CGG QATR does not assign responsibilities beyond those listed in Section A.2 and A.3. Station specific responsibilities are listed in the Ginna UFSAR and in lower tier documents.

5. The purpose of the training specified in paragraph 3.2 of the Standard is to ensure that candidates understand the fundamentals of auditing and the requirements for activities to be audited. The fact that persons have not maintained their proficiency does not mean that they need complete re-training; it means only that they have not been able to review and study the applicable Codes, Standards, Procedures, instructions, and other documents related to QA programs and program auditing. CGG considers that the satisfactory performance of an audit under the observation and guidance of a qualified Lead Auditor should ensure that persons with lapsed certification will review and understand the pertinent documents.

6. As stated in the exception, the licensee is ultimately responsible for its QA program and makes the final determination for post installation test requirements and acceptance documentation.

7. Independent master schedules for different programs meet the intent of ANSI N18.7-1976/ANS-3.2 Section 5.2.8 to ensure that safety-related structures, systems, and components continue to operate, keeping parameters within normal bounds or will act to put the plant in a safe condition if they exceed normal bounds.

8. This is conservatively called a reduction in commitment because Ginna is adopting an exception to a standard. Plant security and visitor control requirements have changed significantly since the ANSI standard was published. Site security plans are maintained current with the changing regulatory requirements independent of the QA program.

Page 28 of 29

, QA Program Comparison Matrix R.E. Ginna Nuclear Power Plant, LLC

9. RG 1.37 Revision 0 Paragraph C.3 refers to ANSI N45.2.1-1973 Section 3.2, Water Quality Requirements, for cleaning/flushing. Oxygen content of the cleaning/flushing water does not have any significant effect on cleaning operations.

10. ANSI N45.2.1-1973 Section 3.1.2 gives the requirements for Class B cleanness. Piping systems and components constructed of carbon steel materials upgraded to Class B cleanness will be allowed to exhibit rust staining in accordance with Class C cleanness requirements due to limitations of the original design.

Page 29 of 29