ML19338F498

Forwards Evaluation Results of Util Compliance W/Nrc Regulations Contained in 10CFR20,50 & 100.Affidavit Encl

ML19338F498
Person / Time
Site:	McGuire
Issue date:	10/13/1980
From:	Parker W DUKE POWER CO.
To:	Harold Denton, Youngblood B Office of Nuclear Reactor Regulation
References
NUDOCS 8010200430
Download: ML19338F498 (34)

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DuxE POWER COMPANY Powra Bcit.ntwo 422 SocTa Cucacu Stuzzt. CnAntoTTE. N. C. as24a wimm o. Paa=ca.sa. October 13, 1980

'/ ICE Pettiotmf TELEp=CNE!ASC4 704 Sveau Paoovcwon 373-4c 8 3 Mr. Harold R. Denton . Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. B. J. Youngblood, Chief Licensing Branch No. 1 Re: McGuire Nuclear Station Docket Nos. 50-369

Dear Mr. Denton:

i Enclosed is the results of an evaluation of compliance of McGuire Nuclear Station wiqh the regulations contained in Title 10, Code of Federal

Regulations, Parts 20, 50 and 100. Although the evaluation is generally

. applicable to both units, it has been prepared specifically to demonstrate f compliance of Unit I with the regulations.

If there are questions regarding this matter, please advise.

Ve truly yours,  !

-a wn h- '

William O. Parker, Jr.

GAC:scs Enclosure 84102004(78 A

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.Mr. Harold R, Denton, Director

-October 13, 1980 Page Two WILLIAM 0. PARKER, JR., being duly sworn, states that he is Vice President of Duke Power. Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this document entitled " Compliance of McGuire Nuclear Station Unit I with the NRC Regula-tions of 10CFR Parts 20, 50 and 100"; and that all statements and matters set fo .therein are me and correct to the best of his knowledge.

%& m W illiam O. Parker, Jr...Vi resident Subscribed and sworn to before me this 13th day of October, 1980.

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Rotary Public b My Commiss' ion Expires:

September 20, 1984

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' COMPLIANCE OF MCGUIRE NUCLEAR STATION UNIT 1 WITH THE NRC REGULATIONS OF 10 CFR PARTS 20, 50, AND 100 l

Regulation l (10 CFR) Compliance l

. 20.l(a) This regulation merely states the general purpose for which l the Part 20 regulations are established and does not impose l any independent obligations on licensees.

20.l(b) This regulation describes the overall purpose of the Part l

20 regulations.to control the possession, use and transfer of licensed material by any licensee, such that the total l dose to an individual will not exceed the standards pre-i scribed therein. It does not impose any independent obli-gations on licensees.

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l 20.l(c) Conformance to the ALARA principle stated in this regulation is ensured by the implementation of Duke policies and appro-f priate Technical Specifications and health physics procedures.

Chapters 11 and 12 of the FSAR describe the specific equip-ment and design features utilized in this effort.

20.2 This regulation merely establishes the applicability of the

. Part 20 regulations and imposes no independent obligations i

on those licensees to which they apply.

20.3 The definitions contained in this regulation are adhered to in all appropriate Technical Specifications and proce-dures, and in applicable sections of the FSAR.

20.4 The units of radiation dose specified in this regulation are accepted and conformed to in all applicable station procedures.

20.5 The units of radioactivity-specified in this regulation are accepted and conformed to. in all applicable station procedures.

20.6 This regulation governs the interpretation of regulations by l

the NRC and does not impose independent obligations on licen-

! sees.

20.7 This regulation gives the address of the NRC and does not im-

. pose independent obligations on licensees.

l 20.101 The radiation dose limits specified in this regulation are l complied with through.the. implementation of and adherence l

-to administrative policies and controls and appropriate health physics procedures developed for this purpose. Con-formance is documented by the use of appropriate personnel monitoring devices and the maintenance of all required records.

Regul,ations (10 CFR) Compliance 20.102 When required by this~ regulation, the accumulated dose for any individual permitted to exceed the exposure limits speci-fied in 20.101(a) is determined by the use of Form NRC-4.

Appropriate health physics procedures and administrative policies control this process.

20.103(a) Compliance with this regulation is ensured through the imple-mentation of appropriate health phvsics procedures relating to air sampling for radioactive materials, and bioassay of individuals for internal contamination'. Administrative policies and controls provide adequate margins of safety for the protection of individuals against intake of radio-active materials. The systems and equipment described in Chapters 11 and 12 of the FSAR provide the capability to minimize these hazards, 20.103(b) Appropriate process and engineering controls and equipment, as described in Chapters 11 and 12 of the FSAR, are installed and operated to maintain levels of airborne radioactivity as low as practicable. Wnen necessary, as determined by station administrative guidelines, additional precautionary procedures are utilized to limit the potential for intake of radioactive materials.

20.103(c} The McGuire respiratory protection program implements the requirements of this regulation by ensuring the proper use of approved respiratory protection equipment. The McGuire respiratory protection program incorporates fully the recom-mendations of Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection."

20.103(d) This regulation describes further restrictions which the Commission may impose on licensees. It does not impose any independent obligations on licensees.

20.103(e) The notification specified by this regulation was made as required, on November 13, 1978.

20.103(f) The respiratory. protection program is in full conformance with the requirements of 20.103(c).

20.104 Conformance with this regulation is assured by appropriate Duke Power Company policies regarding employment of indivi-daals under the age of 18 and the Duke Power Company System Health Physics Manual cestricting these individuals' access to restricted areas.

20.105(a) Chapter 11 of the FSAR provides the information and.related radiation dose assessments specified by this regulation.

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Regulation (10 CFR) Compliance 20.105(b) The radiation dose rate limits specified in this regulation are complied with through the implementation of procedures, Technical Specifications, and administrative policies which control the use and transfer of radioactive materials. Appro-priate. surveys and monitoring devices document this compliance.

20.106(a) Conformance with the limits specified in this regulation is assured through the implementation of procedures and applicable Technical Specifications which provide adequate sampling and analyses, and monitoring of radioactive materials in effluents

'before and during their release. The level of radioactivity in station effluents is minimized to the extent practicable by the use of appropriate equipment designed for this purpose, as described in Chapter 11 of the FSAR.

20.106(b) Duke Power Company has not and does not intend to include in 20.106(c) any license or amendment applications proposed limits higher than those specified in 20.106(a), as provided for in these regulations.

20.106(d) Appropriate allowances for dilution and dispersion of radio-active effluents are made in conformance whith this regula-tion, and are described in detail in Chapter 11 of the FSAR, and in appropriate reports required by the Technical Speci-fications.

20.106(e) This regulation provides criteria by which the Commission may impose further limitations on releases of radioactive materials made by a licensee. It imposes no independent obligations on licensees.

20.106(f) This regulation merely states that the provisions of 20.106 do not apply to disposal of radioactive material into sani-tary sewerage systems. It imposes no independent obligations on licensees.

20.107 This. regulation merely clarifies that the Part 20 regulations are not intended to apply to the intentional exposure of patients to radiation for the purpose of. medical diagnosis or therapy. It does not impose any independent obligations on licensees.

20.108 Necessary bioassay equipment and procedures, including Whole Body Counting, are utilized at McGuire Nuclear Station to

. determine exposure of individuals to concentrations of radio-

. active materials. Appropriate health physics procedures and administrative policies implement this requirement.

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Regul'ation (10 CFR) Compliance 20.203(e) The area and room posting requirements set forth in this regulation pertaining to radioactive materials are complied with through the implementation of appropriate station health physics procedures, and portions of the System Health Physics Manual.

20.203(f) The container labeling requirements set forth in this regu-lation are complied with through the implementation of appro-priate station health physics procedures, and portions of the System Health Physics Manual.

20.204 The posting requirement exceptions described in this regula-tion are used where appropriate and necessary at McGuire Nuclear Station. Adequate controls are provided within the station health physics procedures to assure safe and proper application of these exceptions.

20.205 All of the requirements of this regulation pertaining to procedures for picking up, receiving, and opening packages of radioactive materials are implemented by the System Health Physics Manual and appropriate station health phy-sies procedures. These procedures alco provide for the necessary documentation to ensure an auditable record of compliance.

20.206 The requirements of 10 CFR 19.12 referred to by this regula-tion are satisfied by the orientation training conducted at McGuire Nuclear Station. Appropriate departmental procedures set forth requirements for all employees who frequent or work at McGuire Nuclear Station to receive this instruction on a periodic basis.

20.207 The storage and control requirements for licensed materials in unrestricted areas are conformed to and documented through the implementation of station health physics procedures and applicable portions of the System Health Physics Manual.

20.301 The general requirements for waste disposal set forth in this regulation are complied with through station operating pro-cedures, Technical Specifications, and the provisions of the station license.

20.302 No such application for proposed disposal procedures, as described in this regulation, has been made or is contem-plated by Duke Power Company.

20.303 No plans for waste disposal by release into sanitary sewerage systems, as provided for in this regulation, are contemplated by Duke Power Company, nor is this practice currently utilized.

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Regulation (10'CFR) Compliance 20.201 The surveys required by this regulation are performed at adequate frequencies and conta'.n such detail as to be con-sistent with the radiation hazard being evaluated. When necessary, the Radiation Work Permit system established at the station provides for detailed physical surveys of equipment, structures and work sites to determine appropriate levels of radiation protection. The Dukn Power Company System Health Physics Manual and applicable station health physics procedures require these surveys and provide for their documentation in such manner as to ensure compliance with the regulations of 10 CFR Part 20.

20.202(a) The System Health Physics Manual and applicable station health physics procedures set forth policies and practices which ensure that all individuals are supplied with, and required to use, appropriate personnel monitoring equipment.

The Radiation Work Permit system is established to provide additional control of personnel working in radiation areas and to ensure that the level of protection afforded to these individuals is consistent with the radiological hazards in the work place.

20.202(b) The terminology set forth in this regulation is accepted and conformed to in all applicable station procedures, Technical Specifications, and those portions of the System Health Physics Manual in which its use is made.

20.203(a) All materials used for labeling, posting, or otherwise desig-nating radiation hazards or radioactive materials, and using the radiation symbol', conform to the conventional design pre-scribed in this regulation.

20.303(b) This regulation is conformed to through the implementation of appropriate station health physics procedures and portions of the System Health Physics Manual relating to posting of radiation areas, as defined in 10 CFR Part 20.202(b)(2).

20.203(c) The requirements of this regulation for "High Radiation Areas" arr conformed to by the implementation of the Technical Sp.ecifications and appropriate station health physics proce-dures, as well as the System Health Physics Manual. The con-trols and other protective measures set forth in the regulation are maintained under the surveillance of the station Health Physics group.

20.203(d) Each Airborne Radioactivity Area, as defined in this regula-tion, is required to be posted by provisions of the System Health Physics Manual and appropriate station health physics procedures. These procedures also provide for the surveil-lance requirements necessary to determine airborne radio-activity levels.

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Regulation

'(10'CFR) Compliance 20.304 Disposal of wastes by burial in soil (i.e., onsite burial),

3 as provided for in this regulation,-is not performed or being contemplated by Duke Power Company.

20.305 Specific authorization, as described in this regulation,

< is not_ currently being sought by Duke Power Company for j treatment or disposal of wastes by. incineration.

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i 20.401 . All of the requirements of this regulation are complied j with through the implementation of appropriate Technical i Specifications and station procedures pertaining to records

[ of surveys, radiation monitoring and waste disposal. The i retention periods specified for such records are also pro-l vided for in these specifications _and in station and depart-

{- mental procedures.

t l 20.402 'McGuire Nuclear Station has established an appropriate inven-j . tory and ~ control program to ensure strict accountability for

! all licensed radioactive materials. Reports of theft or loss

! of licensed material are required by reference to the regula-

'tions of 10 CFR in the Technical Specifications.

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[ 20.403 Notifications of incidents, as described in this regulation, i

are assured by the requirements of the Technical Specifica-tions, the System Health Physics Manual and appropriate 1

station procedures, which also provide for the necessary

!- assessments to determine the' occurrence of such incidents, i

20.404 This regulation was_ deleted effective September 17, 1973 (38 Fed.1 Reg. 22220).

3 ~ 20.405 ReportsEof overexposures to radiation and the occurrence

'of excessive levels and concentrations, as required b'y-this ~ regulation, are_ provided. for by reference in the i-

-Technical Specifications and in appropriate health physics l_ procedures.

i 20.406 This regulation was ' deleted August 17, 1973, effective f~ September' 17, 1973 (38 Fed. Reg. 22220).

5 20.407 The personnel monitoring report required by this regulation

, is expressly provided for by the Technical Specifications.

, LAppropriate health physics procedures establish the data base from which this report is. generated.

~ 20.408 ~ The report-of' radiation exposure required by this regula-4 tion upon termination ofjan individual's employment or work ,

assignment is generated through the provisions of Duke

,- ~ Power Company procedures.

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e Regulation (10 CFR) Compliance 20.409 The notification and reporting requirements of this regula-tion, and those referred to by it, are satisfied by the provisions of Duke Power Company procedures.

20.501 This regulation provides for the granting of exemptions from 10 CFR Part 20 regulations, provided such exemptions are authorized by law and will not result in undue hazard to life or property. It does not impose independent obli-gations on liceusees.

20.502 This regulation describes the means by which the Commission may impose upon any licensee requirements which are in addition to the regulations of Part 20. It does not im-pose independent obligations on licensees.

20.601 This regulation describes the remedies which the Commission may obtain in order to enforce its regulations, and sets forth those penalties or punishments which may be imposed for violations of its rules. It does not impose any inde-

. pendent obligations on licensees.

n Regulation (10'CER) Compliance 50.1 This regulation states the purpose of the Part 50 -regula-tions and does not impose any independent obligations on licensees.

50.2 This regulation defines various terar and does not impose independent obligations on licensees.

50.3 This regulation governs the interpretation of the regula-tions by the NRC and does not impose independent obliga-tions on licensees.

50.4 This regulation gives the address of the NRC and does not impose independent obligations on licensees.

50.10 inese regulations specify the types of activities that may 50.11 not be undertaken without a license from the NRC. Duke Power Company does not propose to conduct any such activi-ties at McGuire Nuclear Station without an NRC license.

50.12 This regulation provides for the granting of exemptions from 10 CFR Part 50 regulations, provided such exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. It does not impose independent obligations on licensees.

50.13 This regulation says that a license applicant need not design against acts of war. It imposes no independent obligations on licenses.

50.20 These regulations merely describe the types of licenses 50.21 that the NRC issues. They do not address the substantive 50.23 requirement; that an applicant must satisfy to qualify for such licenses.

50.24 This regulation has been deleted, 35 Fed. Reg. 19655.

50.30 This regulation sets down procedural requirements for the filing of license applications, such as the number of copies of the application that must be provided the NRC. Duke Power Company has substantially complied with the procedural re-quirements in effect at the time when filing its license application and the amendments to it. In particular, 10 CFR 50.30(f) requires that a license application must be accompanied by any Environmental Report required pursuant to 10 CFR Part 51, and Duke Power Company has submitted an Environmental. Report covering McGuire Nuclear Station.

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s l Regulation (10'CFR) Compliance 50.31 These regulations merely permit more efficient organization l 50.32 . of the license application and impose no independent obliga-l tions on licensees.

50.33 This regulation requires the license application to contain certain general information, such as an identification of the applicant, information about the applicant's financial qualifications, and a list of regulatory agencies with juris-diction over the applicant's rates and services. This infor-mation was provided in the McGuire Nuclear Station operating l license application, i

l 50.33a .This regulation requires applicants for construction permits to submit information required for antitrust review. The antitrust review required by the Atomic Energy Act of 1954, l as amended, was performed at the construction permit stage.

I 50.34(a) This regulation-governs the contents of the Preliminary Safety Analysis Report and is relevant to the construction l

permit stage rather than the operating license stage.

l 50.34(b) A Final Safety Analysis Report (FSAR) has been prepared I

and submitted, which addresses in the chapters indicated j the information required:

(1) site evaluation factors - Chapter 2 (2) structures, systems, and components - Chapters 3, 4, 5, 6, 7, 8, 9, 10,.11, 12, and 15 (3) radioactive effluents and radiation protection -

Chapters 11 and 12 l (4) design and performance evaluation - ECCS performance j is discussed and shown to meet the requirements of 10 CFR 50.46 in Chapters 6 and 15 (5) results of research programs - Chapter 1 (6) (i) organizational structure - Chapter 13 l (ii) managerial and administrative controls -

l Chapters 13 and 17. Chapter 17 discusses compliance with the quality assurance re-quirements of Appendix B.

(iii) . plans for preoperational testing and initial operations - Chapter 14-Y

l Regulation (10'CFR) Compliance (iv) plans for conduct of normal operations -

Chapter 13 and 17. Surveillance and periodic testing is specified in the Technical Speci-fications.

(v) plans for coping with emergencies - Emergency Plan (Chapter 13) l (vi) Technical Specifications - prepared in conjunction l

with the Staff (Chapter 16)

(vii) not applicable, since the operating license appli-cation was filed before February 5,1979 l (7) technical qualifications - Chapter 16 - (Eventually j superceded by Standard Technical Specifications).

(8) operator qualification program - Chapter 13 and Duke Power Company document "McGuire Nuclear Station -

l Response to TMI Concerns" dated May 23, 1980. The

! latter document indicates changes made to the operator l

requalification program as a result of NUREG-0660.

50.34(c) A physical security plan was prepared and submitted as re-quired by this regulation for McGuire Nuclear Station.

50.34(d) A safeguards contingency plan has been prepared and sub-mitted as required by this regulation for McGuire Nuclear Station.

! 50.35 This regulation is relevant to the construction permit stage l rather than the operating stage.

l 50.36 Technical Specifications have been prepared and implemented, including items in each of the categories specified, including:

(1) safety limits and limiting safety settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

l 50.36a The McGuire Nuclear Station Technical Specifications include

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specifications which require compliance with 10 CFR 50.34a (releases as low as is reasonably achievable), and that en-sure that concentrations of radioactive effluents released

.. to unrestricted areas are within the limits specified in l 10 CFR 20.106. 'The reporting requirements of 10 CFR 50.36a (a)(2) are also included in these specifications.

50.37 This regulation requires the applicant to agree to limit access to Restricted Data. Duke Power Company's agreement to do so is contained in the operating license application for McGuire Nuclear Station.

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' - Regulation-  ;

(10'CFR) Compliance l

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50.38 -

This regulation prohibits the NRC from issuing a license >

to foreign-controlled entities. Duke Power Company's  ;

. statement that it is not owned, controlled, or dominated by an alien, foreign corporation, or foreign government is

. contained in the operating license application for McGuire '

Nuclear Station.

i 50.39 This regulation provides that applications and related g documents may be made available for-public inspection.

i This imposes no direct obligations on applicants and

' licensees.

l 50.40 This regulation provides considerations to " guide" the Commission in granting licenses r.s follows:

50.40(a) The design and operation of the facility is to provide reasonable assurance that the applicant will comply with NRC regulations, including those in 10 CFR Part 20, and

that the health and safety of the public will not be en-3 dangered. The basis for Duke Power Company's assurance 1

that the regulations will be met and the public protected

is contained in this enclosure and in the license application  ;

and the related correspondence over the years. Moreover, 1 the lengthy process by which the plant is designed, con-

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• structed, and reviewed, including reviews by Duke Power i Company's own staff, the NRC staff, the ACRS, and NRC i licensing 'ooards, provides a great deal of assurance that the public health and safety will not be affected. In ,

particular, the Atomic Safety and Licensing' Board, after  !

an extensive-review, concluded that Duke Power Company had the commitment and technical qualifications necessary to operate McGuire Nuclear Station' safely and in compliance with all applicable radiological health and safety re-quirements (see below).

50.40(b) Another consideration is that the applicant be technically and financially qualified. Both Duke Power Company's 1

technical qualifications and its financial qualifications i were reviewed in hearings before the Atomic Safety and Licensing Board at both the. construction permit and ope- t rating license stages. Favorable initial decisions were

- issued as a result of bath proceedings.

!50.40(c) Another consideration'is that the issuance of the license is not to be inimical to the common defense and security

! or to the health and safety of the public. -The indivi-l' dual showings of compliance with.particular regulations contained-in this enclosure as well as the contents of-j the entire'FSAR and related correspondence over the years, I

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e Regulation ,

(10'CIR) Compliance plus the lengthy process of design, construction, and review by Duke Power Company, its NSSS vendor, and the government, provide Duke Power Company with considerable assurance that the license will not be inimical to the health and safety of the public. As for the common defense and the security, there is considerable assurance that the license will not be

. inimical in that Duke Power Company has a viable security plan for McGuire Nuclear Station that Duke Power Company is not controlled by agents of foreign countries, and that Duke

?Jwer Company has. agreed to limit access to Restricted Data (see above).

50.40(d) The final 50.40 " consideration" is that the applicable re-quirements of Part 51 have been satisifed. Part 51 concerns compliance with the National Environmental Policy Act of 1969.

Duke Power Company has submitted an Environmental Report which has been reviewed by the NRC staff. The results of the staff review are contained in the Final Environmental Statement for McGuire Nuclear Station, NUREG 0063, April, 1976.

50.41 This regulation applies to class 104 licensees, such as those for devices used in medical therapy. McGuire Nuclear Statina has not applied for a class 104 license, and so 50.41 is not applicable.

50.42 Section 50.42 provides additional " considerations" to " guide" the Commission in issuing Class 103 licenses. The two con-siderations are: (a) that the proposed activities will serve a useful purpose proportionate to the quantities of special nuclear material or source material to be utilized and (b) that due account will be taken of the antitrust ad-vice provided by the Attorney General under subsection 105c of the Atomic Energy Act. The "useful purpose" to be served is the production of electric power. The need for the power was determined by the licensing board at the construction permit stage. Although conditions affecting the need for power are constantly chao;$ng, Duke Power Company periodi-cally makes load projections, and in Duke Power Company's judgment the need for McGuire Nuclear Station is still substantial. As for the amount of special nuclear material or source material used, there is no reason to believe that their proportion in relation to the power produced is sub-stantially greater than.that of other commercial power reactors in this country. As for the antitrust advice of the Attorney General, as noted above, the antitrust review was done at the construction permit stage.

50.43 This regulation imposes certain duties on the NRC and addresses the applicability of the Federal Power Act and the right of government agencies to obtain NRC licenses.

It imposes no direct obligations on licensees.

Reguration

, (10 CFR) Compliance I

50.44 The McGuire Nuclear Station combustible gas control system is described in FSAE Section 6.2.5. The system is designed to maintain the hydrogen concentration in containment at a.

1 safe level following a LOCA, without purging the containment

' atmosphere, as specified in 10 CFR 50.44(e). The system con-

sists of internal recombiners, a hydrogen analyzer, and a hydrogen skimmer system. The containment recirculation

( system and hydrogen purge system complement the recombiner t system. McGuire Nuclear Station meets-the requirements of NUREG-0660 and NUREG-0694. The requirements of 10 CFR j 50.44 are satisfied.

i 50.45 This regulation provides standards for construction permits rather than operating licenses and is therefore not material i to this operating license proceeding.

I 50.46 FSAR Sections 6.3 and 15.4.1 describe the Emergency Core [

Cooling System and the methods used to analyze ECCS per-l . formance following a postulated loss of coolant accident. ,

In'FSAR Section 15.4.1, Duke Power Company provided the re-sults of a LOCA-ECCS analysis for McGuire Nuclear Station using an NRC approved evaluation model, which is in com-pliance with Appendix K to 10 CFR 50. The analysis, based

! on an overall peaking factor (Fq) of 2.32, provided results i in compliance with the. criteria of 10 CFR 50.46(b). The Fq l limit will be reflected in the Technical Specifications.

i 50.50 This regulation provides that the NRC will issue a license ,

i upon determining that the application meets the standards and requirements of the Atomic Energy Act and the regula-j . tions and that the necessary notifications. to cther agencies

.or bodies have been duly made. It imposes no direct obli-4 gations on licensees.

1 50.51 This. regulation specifies the maximum duration of licenses.

Compliance will be affected simply by the Commission's writing.the license so as to comply.

50.52 .This regulation provides for. the combining in a single

-license of a number of activities. It imposes no inde-pendent obligation on the licensee. i 50.53 This regulation provides that licenses are,not to be issued for activities that are not under or within the jurisdiction l

'of the United States. ' The operation of McGuire -Nuclear Sta-t tion will be within the United States and subject to the

. jurisdiction of.the United States, as is evident from the i

. description.of the facility in the operating license appli-cation.

Regulation

,fl0 CER) Compliance 50.54 This regulation specifies certain conditions that are incorporated in every license issued. Compliance is cffect-ed simply by including these conditions in the license when it is issued. Indeed, much of 50.54 merely provides that other provisions of the law apply, which would be the case even without 50.54.

50.55 This regulation addresses conditions of construction permits, not operating licenses, and so it is not relevent at this point.

50.55a(a)(1) Various chapters of the FSAR discuss design, fabrication, erection, construction. testing, and inspection of safety-related equipment. For example, Chapter 14 provides infor-mation on testing of safety-related systems. Chapter 17 pro-vides information concc -i ing .the Quality Assurance Program that was utilized. As a further example of a specific system, Chapter 5, Section 5.2, " Integrity of the Reactor Coolant System Boundary," discusses the design of the reactor coolant system.

50.55a(a)(2) This paragraph is a general patagraph leading into paragraphs (c) through (i) of the regulation.

50.55a(b)(1) These paragraphs provide guidance concerning the approved 50.55a(b.,2) Edition and Addenda of Section III and XI of the ASME B&PV Code.

50.55a(c) Design and fabrication of the reactor vessel was carried out in accordance with Class 1, ASME Section III (1971) through Summer 1971 Addenda. Information can be found in Chapter 5 of the-FSAR.

50.55a(d) Reactor coolant system piping meets the requirements of Class 1, ASME,Section III, 1971 of the FSAR. Information can be found in Chapter 5 and Chapter 3.

50.55a(e) Reactor Coolant pumps meet the requirements of Class 1, ASME Section III (1971). Information can be found in Chapter 5 and Chapter-3 of the FSAR.

50.55a(f) Noting the construction permit date of February,1973, the valves within the reactor coolant system pressure boundary were designed and fabricated in accordance with the re-quirements of ASME Section III, 1971 edition.

50.551(g) Inservice Inspection (ISI) requirements are delineated in this part and are specified in the Technical Specifi-cations, paragraph 4.0.5. The McGuire inservice inspection

Regulation (10 CFR) Compliance program is delineated in Section 5.2.8 of the FSAR. Inser-vice testing of pumps and valves is described and exceptions have been requested in a document submitted for staff review on November 14, 1978 (MC-1WP/lWV-780V).

The staff's SER for McGuire Nuclear Station, Section 5.2.3 provides a discussion of inservice testing of pumps and valves. Additional information on ISI can be found in FSAR Section 5.2.8.

50.55a(h) As discussed in Chapter 7, the protection systems meet IEEE 279-1971.

50.55a(i) Fracture toughness requirements are set forth in Appendices G and H of 10 CFR 50. Technical Specifications require the use of reactor vessel material irradiation surveillance specimens and updating of the "heatup" and "cooldown" curves given in the Technical Specifications. Further iltformation is given in FSAR Section 5.2.4.4. concerning the irradiation surveillance program.

50.55b This regulation has been revoked. 43 Fed. Reg. 49775.

50.55e this regulation is only proposed, 39 Fed. Reg. 26297, and applies to fuel reprocessing plants.

50.56 This regulation provides that the Commission will, ia the absence of good cause shown to the contrary, issue an ope-rating license upon completion of the construction of a facility in compliance with the terms and conditions of the construction permit. This imposes no independent obliga-tions on the applicant.

50.57(a) This regulation requires the Commission to make certain findings before the issuance of an operating license. These findings for McGuire Nuclear Station can be made for the reasons given in this enclosure generally. Specifically:

(1) Construction of the facility has been substantially completed in conformity with the construction permit and the application as amended. Conformance of the facility to the NRC rules and regulations and the Act, as implemented by the regulations, has been demonstrated by the application.

(2) The Technical Specifications and resulting operating procedures provide assurance that the unit will operate in conformity with the applicationias amended and with the rules and regulations, with the noted exceptions to 10 CFR 50.

Regulation-(10'CFR) Compliance (3) The application demonstrates that the facility can be operated without endangering the health and safety of the public and in compliance with the regulations, as noted above.

(4) The application demonstrates that Duke Power Company is technically and financially qualified to operate the unit.

(5) The applicable provisions of 10 CFR 140 have been satisfied.

(6) The McGuire Security Plan assures that special nuclear material is being appropriately safeguarded. The application demonstrates that the operation of the unit will not be inimical to the health and safety of the public.

50.57(b) The license, as issued, will contain appropriate conditions to assure that items of construction or modification are com-pleted on a schdule acceptable to the Commission.

50.57(c) This regulation provides for a low-power testing license.

i Such a license has been requested for licGuire Nuclear Station.

50.58 This regulation provides for the review and report of l the Advisory Committee on Reactor Safeguards. The ACRS

! has reviewed the operating license application for McGuire Nuclear Station in accordance witht its usual practice.

l l 50.59 This regulation provides for the licensing of certain changes, tests, and experiments as a licensed facility. Technical Specifications and procedures provide implementation of this regulation.

50.60 This regulation has been deleted, 40 Fed. Reg. 8790.

50.65 This regulation has been deleted, 43 Fed. Reg. 6915.

50.70 The Commission has assigned resident inspectors to McGuire Nuclear Station. Duke Power Company has provided office space in accordance with the requirements of this section.

! Duke Power Company permits access to the station to NRC inspectors in accordance with 10 CFR 50.70(b)(3).

50.71 Records are and will be maintained in accordance with the requirements of sections (a) through (e) of this regula-tion and the license. Section (e) requires that the FSAR

i Regulation

,(10'CFR) Compliance

-be_ updated within 24 months after date of-issuance of operatiag

- license and~ annually thereafter. Such updates will be made.

50.72 Notification of.significant events to the NRC will be made in'accordance with the requirements in this regu-

.lation.

50.80 - This regulation provides that licenses may not ce trans-ferred'without NRC consent. No application for' transfer of a license is involved in the McGuire Nuclear Station

• proceeding.

50.81 This regulation permits the creation of mortgages, pledges,

'and liens on licensed facilities, subject to certain provi-

-sions. These provisions concern the requirements and re-- ,

strictions on creditors and do not impose independent obli-

.gations on licensees.

50.82 This regulation provides for the termination of licenses.

It does not apply to McGuire Nuclear Station because Duke Power Company has not requested the termination of a license.

50.90 This -regulation governs applications for amendments to licenses. Future requests for license amendments will be made in accordance with these requirements.

~ 50.91 ' This regulation provides guidance to the NRC in issuing license amendments.

50.100' These regulations govern the revocation, suspension, and-50.101 modification _of licenses by the Commission under unusual 50.102 circumstances. No such circumstances are-present in the 50.103 McGuire Nuclear Station proceeding, and these ' regulations are not applicable.

50.109 . This regulation specifies the conditions under which the NRC may require the backfitting of a facility. This reg-ulation imposes no independent obligations on a licensee.

unless the NRC proposes a.backfitting requirement.

50.110 This regulation governs enforcement of the Atomic Energy Act, the Energy Reorganization Act of 1974, and the NRC's regulations and orders. No enforcement action is at issue in the McGuire Nuclear Station proceeding,.and so this regulation is not applicable.

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/ Appendix A l GDC l< Section 3.1 of the FSAR describes the design provisions

l. made to ensure that these requirements are met. Codes'and standards utilized for the unit are specified throughout the FSAR. . Chapter 17 describes the quality. assurance program and the_ provisions for maintenance of records.

GDC:2 FSAR'Section 3.1-addresses the. design considerations for natural' phenomena,which are described in' detail.in Chapters 2.and 3. -Appropriate considerations have been made in the.-

design basis.for historical data,' combined effects of normal and accident conditions with the effects of . natural phenomena, Land the importance of thessafety functions to be performed. i

-GDC 3. FSAR Section 3.1 describes in general the~ measures which l- have been taken to minimize the probability and effects of fires and: explosions. Section 9.5.1 describes the fire

. detection and protection systems. In addition,. improvements to the fire protection systems have_been and are being made in accordance with NRC requirements based on Appendix A to .

BTP APCSB 9.5-1. These modifications will be completed as indicated in Table 9.5-1 of Supplement 2 to the SER.

GDC 4- FSAR Section~3.1 describes'the design features used to accommodate >the effects of and compatibility with the environmental conditions' associated with all modes of operation and postulated accidents. Chapter 3 provides

-information concerning the specific design features for

~

protection against missles,' jet impingement and pipe rupture. Provisions for. qualification of equipment for ,

all postulated environments is' described in several.  !

sections of the FSAR. A NUREG-0588 review has confirmed that electrical equipment has been adequately demonstrated to be qualified for is expected service environments. This Jevaluation was provided to the staff in an August 13, 1980 ,

letter. ' Additional information will be provided by j October ~15,'1980. i l

GDC 5: As described'in FSAR Section 3.1, those structures,

, systems and components which are shared with~ Unit 1 are tabulated in FSAR-Section 1.2.2.12. It is concluded that~ safety functions are not significantly impaired by-

.such sharing. j l _GDC 10 FSAR Section 3.1 indicates'that the reactor core and L associated systems are designed to function throughout-l' theEdesign lifetime _without exceeding fuel: damage limits, using protection criteria specified~in Section-3.1 and '

Chapters 4,"7, and~15.

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(10'CFR) Compliance GDC 11 FSAR Section 3.1 indicates that prompt compensatory reacti-vity feedback effects are assured by unit design and opera-tional limit considerations. The core inherent reactivity feedback characteristics and reactivity control methods are described in F,SAR Section 4.3.

GDC 12 FSAR Section 3.1 describes the inherent and design features j which eliminate or limit the various types of oscillations.

Core stability is further described in Section 4.3.

GDC 13 As indicated in FSAR Section 3.1, and described in more de-tail in Chapter 7, instrumentation and control systems have

been provided to monitor and maintain plant variables in-cluding those variables which affect the fission process, integrity of the reactor core, the reactor coolant pressure boundary, and the containment, over their prescribed ranges for normal operation, anticipated occurrences, and under accident conditions.

GDC 14 FSA3 Section 3.1 indicates that the reactor coolant pressure boundary has been designed to accommodate the system temper-tures and pressures attained under all expected operational modes and anticipated transients, and to maintain stresses within applicable limits.

GDC 15 As indicated in FSAR Section 3.1, the reactor coolant system and associated auxiliary, control and protection systems are designed to ensure the integrity of the reactor coolant pressure boundary with adequate margins during normal ope-rations and anticipated transients. The design codes used for the Reactor Coolant System are described in Chapter 5.

Details concerning the protection systems are provided in Chapter 7.

GDC 16 As described in FSAR Section 3.1 and Chapter 6, an ice con-denser containment structure is provided. It is designed to sustain, without loss of required integrity, all effects of gross equipment failures, up to and including the rupture of the largest pipe in the reactor coolant system. The con-tainment and its associated engineered safety features thus meet the required functional capability of this criteria.

GDC 17 As described in FSAR Section 3.1, onsite and offsite power systems are provided which can independently supply'the electric power required for the operation of safety-related systems. This capability is maintained even with the failure of any single active component in either system. Chapter 8 provides the design details of the power systems and their compliance with this criterion.

m Regulation (10'CTR)- Compliance GDC 18 As described in FSAR Section 3.1 and Chapter 8, the redun-dant electric power systems important to safety are contin-uously monitored and energized during normal plant operation from redundant offsite power sources. Radundant onsite diesel generators provide automatic backap power sources.

g Periodic tests of the diesel generators, the transfer system and the station batteries are made, as required by Technical Specifications.

GDC 19 FSAR Section 3.1 describes the main control room, which con-tains the controls and instrumentation necessary for safe operation of the unit during normal and cccident conditions.

. Sufficient shielding, distance, structural integrity, and ventilation systems are provided to ensure that control room personnel will not receive radiation exposures in excess of the criterion for the duration of the accident.

In the event that access to the main contial room is restricted, an auxiliary control room is provided, within the protected envelope, which may be used to bring the reactor to cold shutdown.

GDC 20 FSAR Section 3.1 discusses the design criteria for the protection system and engineered safety features actuation, to ensure that the requirements of this criterion are met.

Further details are supplied in Chapter 7.

GDC 21 As indicated in FSAR Section 3.1, the protection system is ,

designed for the high functional reliability and inservice .

testability commensurate with the safety functions La be per-fo rmed. This section, as well as Chapter 7, describe in de-tail the design features provided to ensure redundancy and testability.

GDC 22 FSAR Section 3.1 indicates that the protection system has been designed to provide sufficient resistance to a broad class of accident conditions or postulated events. Chapter 7 provides further design details concerning this resistance such that independence is maintained.

GDC 23 As indicated in FSAR Section 3.1, the protection system is designed with due consideration of the most probable failure modes of the components under various perturbations of energy sources and the environment. Further details are supplied in Chapter 7.

GDC124 FSAR Section 3.1 discusses separation of the protection and control systems, such that the failure of any signal control system component or channel or the failure or removal from

. ~

-Regulation -

(10'CFR) Compliance-service of any protection system component or channel which is common to the protection and control systems, leaves in-tact a system satisfying all. redundancy, reliability, and independence requirements of.the protection system. Details concerning separation of protection and control systems are

- provided in-Chapter 7.

GDC 25 FSAR Section 3.1 indicates that the protection system has l been designed to assure that specified acceptable fuel de-I sign limits are not exceeded'in the event of any siugle i reactivity control malfunction, including an accidental withdrawal of control cluster groups. Further details cre provided FSAR Sections 4.3.1.4, 7.2.2.2.3, and 7.7.2.2.

l GDC 26 As indicated in FSAR Section 3.1, two independent reactivity control systems of different design principles are provided.

One of the systems uses control rods; the second system employs dissolved boron as a chemical shim. Reactivity control system redundancy and capability are described further in l Sections 4.3.1.5 and 7.7.2.2.

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l~ GDC 2) As described in FSAR.Section 3.1, means are provided for j . shutdown reactivity for cooling'the core under any antici-pated condition and with appropriate margin for contin-gencies. Combined use of rod cluster control and chemical shim control permit the necessary shutdown margin to be maintained during the long term xenon decay and plant cool-down. These means are discussed in detail in FSAR Sections 4.3 and 7.2.

GDC 28 FSAR Section 3.1 indicates that core reactivity is controlled by a chemical poison dissolved _in the coolant, rod cluster assemblies and burnable poison rods. The maximum reactivity insertion. rates due-to withdrawal of a bank or rod cluster control assemblies or by' boron dilution are limited. The maximum worth of control rods and the maximum rates of reactivity insertion employing control rods are limited to values which prevent rupture of the coolant pressure boundary or disruption of the core internals to a degree which would

- impair core cooling capacity.' Further details are provided

! in Section 4.3.

GDC 29 As indicated in FSAR Section 3.1, the protection and reacti-

vity control systems are designed to assure extremely high probability of' performing their required ' safety functions in the event of anticipated operational occurrences. The L protection. system is further discussed in Section 7.2. The reactivity control systems areLdiscussed in Sections 4.2.3 and 7.'7.

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< (10'CFR) - Compliance 7 - GDC 30 As described in FSAR Section 3.1, reactor coolant pressure

, boundary components are designed, fabricated, inspected, and tested in conformance-with ASME Section III. Major

- ' components are classified as seismic Class 1 and are i accorded the quality measures appropriate to this class-ification. The evaluations of reactor coolant pressure i

boundary components are discussed in Section 5.2.

~i GDC 31 As-indicated in FSAR Section 3.1, close control is maintain-ed over material selection and fabrication for the reactor

coolant system to ensure that the boundary behaves in a t nonbrittle manner. -The materials testing is consistent with 10 CFR 50, Appendixes G and H. These tests ensure the selection of materials with proper toughness properties

and margins as well as verify the integrity of the reactor coolant pressure boundary. Operating procedures and Tech-1 nical Specifications ensure operation within the pressure-temperature-limit relative to this criterion.

l GDC 32 FSAR Section 3.1 describes how the design of the reactor i- vessel and its arrangement in the system provides the capa-bility for accessibility during' service life to the entire

internal surfaces of the vessel and certain external zones

) of the vessel. The reactor arrangement within the contain-ment provides sufficient space for inspection of the external surfaces of the reactor coolant piping, except for the area-

, of pipe within-the primary shielding concrete. Additional details can be found in Section 5.2.

~

GDC 33 As indicated in FSAR Section 3.1, the chemical and volume control system provides a means of reactor coolant makeup and adjustment of- the boric acid concentration. A high de-gree of functional reliability and safe response to probable modes of failure is-assured by provision of standby compo-nents. Details of system design are included in Section 9.3 and details of'the electrical power systems are given in Sections 8.2 and 8.3.

GDC 34 FSAR Section-3.1 indicates that the residual heat removal 1

' system, in conjunction with the steam and power conversion system, is designed to transfer the fission product decay _

heat and other residual heat from the reactor core within acceptable limits. Suitable redundancy is accomplished l below 350 F-with the two residual heat removal pumps with

means available for draining and monitoring of leakage, two residual heat exchangers, and the associated piping and cabling. The residual heat removal system is able to ope-

- rate on either onsite or offsite electrical power. Suitable

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Regulation (10'CFR) Compliance redundancy'above 350 F is provided by the steam genera-

, tors, aaxiliary feed pumps, and attendant piping. Details

of the residual heat removal system design are in FSAR Section 5.5.7.

GDC 35 FSAR Section 3.1 describes the use of passive accumulators with two centrifugal charging pumps and two low head safety injection pumps to provide redundancy for failure of any -

component in any system. The primary function of the emer-gency core cooling system is to deliver borated cooling water to the reactor core in the event of a loss-of-coolant accident. This limits the fuel clad temperature and thereby ensures that the core will remain substantially intact and in place, with its essential heat transfer geometry pre-served. Further details are provided in Chapters 6 and 7.

GDC 36 As described in FSAR Section 3.1, design provisions are made for inspection to the extent practical of all com-

- ponents of the emergency core cooling system. An inspec-tion is performed periodically to demonstrate system readi-ness. To the extent possible, the critical parts of ;he reactor vessel internals, injection nozzles, pipes, valves, and pumps are inspected visually or by boroscopic examina-

tion for erosion, corrosion, and vibration wear evidence.

< Nondestructive inspection is performed where such techni-ques are desirable and appropriate. Technical Specifica-tions require inservice inspection in accordance with appli--

cable ASME Codes. Details of the inspection programs are provided in Chapters 5 and 6.

i GDC 37 FSAR Section 3.1 indicates that the components of the emer-gency core cooling system located outside the containment will be accessible for leaktightness inspection during

. appropriate periodic tests. Each active component of i

the system may be individually actuated on the normal power source at any time during plant operation to demon-strate operability._ The centrifugal charging pumps are part of'the charging system, and this system is in continuous ope-ration-during plant operations. Actuation circuits are test-ed and remote-operated valves are exercised periodically.

. The testing is described in. detail in FSAR Sections 6.3.4, 7.3.2.2.5, and per Technical Specification surveillance re-quirements.

GDC 38 As indicated in roar Section 3.1, the containment spray sys-tem, ice condenser, and RHR spray system are provided to re-move heat from the containment following a loss-of-coolant accident. An air return system is used to circulate. air and steam through the containment after the initial blow-down. This maintains proper mixing of the containment air I" )

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Regulation (10'CFR) Compliance and steam with the heat removal media for the necessary heat

. renoval. .The loss of a single active component was assumed in the design of these. systems. Emergency power system arrangements ensure the proper functioning of these systems.

Two electrical buses, each connected to both onsite and off-site power, feed 'the pump motors and the necessary valves.

- Further details are provided in Sections 6.2 and -8.3.

4 GDC 39 As indicated in Section 3.1,_the ice condenser design in-

' cludes provisions for visual inspections of the ice bed flow channels, doors, and cooling equipment. The air re-turn fan system provides for visual inspection of the fans

, and the associated backflow dampers and for duct systems

. - that are not embedded in concrete. The containment spray f- system and the residual heat removal system (RHR) sprays are I

designed such that active and passive components can be readily inspected to. demonstrate system readiness. Pressure contained systems are inspected for leaks from pump seals, i valve packing, flange joints, and relief valves. During

operational testing of the containment spray pumps and RHR pumps, the portions of the systems subjected to pressure are inspected'for leaks. System design details are given in Section 6.2.

GDC 40 As described in FSAR Section 3.1, the containment heat re-moval systems described in Section 6.2 are designed to per-

~

mit periodic testing so that proper operation can be. assured.

In some cases whole systems can-be operated for test purposes.

-In others, individual components are operated for functional tests so that plant operations are not disrupted. The ice condenser contains no active components required to function-during an accident condition. Samples of the ice are taken periodically and testei for boron concentration. The lower inlet- door. opening force is measured when the reactor is in the shutdown condition. The position of the lower inlet doors is monitored at all times. Top deck door and inter-mediate deck doors are tested for operability during the shutdown condition. All active components of the contain-ment spray system and the residual heat' removal spray system are tested in place after installation. These spray systems receive initial flow tests to assure proper dynamic function-ing. Further testing; of the active components is conducted after component maintenance. Air test lines, located upstream

, lof the spray' isolation valves, are provided for testing to

. assure that spray' nozzles are not obstructed. Testing of

transfer between normal and emergency, power supplies is also conducted. Air return fans and their associated backflow dampers are tested.for cperability while the reactor is. .

shutdown for refueling. .

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Regulation (10 CFR) Compliance GDC 41 As indicated by FSAR Section 3.1, the shield building, sur-rounding the primary co;tainment, serves as a secondary containment. The annulus ventilation system (Section 6.2) maintains this secondary containment at a negative pressure during the entire post-accident period. The annulus venti-lation system also collects and processes the secondary con-tainment atmosphere. After processing, the portion of this processed air necessary to assure a negative pressure is ex-hausted through the plant vent. The remainder is recircula-ted and distributed in the secondary contaxament. The auxi-liary building serves to collect any equipment leakage during the recirculation of containment sump water. The auxiliary building ventilation system (Section 9.4.2) is isolated by an accident signal. The auxiliary building filtered exhaust system (Section 9.4.2) processes any inleakage prior to re-lease to the environment. Postaccident hydrogen control with-in the containment is provided by electrical hydrogen re-combiners (Section 6.2). Distribution of the atmosphere within the containment is provided by the air return fan system (Section 6.6). The system also takes a suction in each com-partment to prevent stagnation and excessive accumulation of hydrogen.

GDC 42 FSAR Section 3.1 indicates that the annulus ventilation system and the hydrogen recombiners are designed to permit appropriate periodic inspection cf the important components. Additional discussion is provided in FSAR Sections 9.4 and 6.2.

GDC 43 FSAR Section 3.1 indicates that the annulus ventilation system and the hydrogen recombiner system are designed to permit periodic pressure testing and functional testing of their com-ponents. Further details are provided in Sections 6.2 and 9.4.

GDC 44 FSAR Section 3.1 describes how a Seismic Category I Component

' Cooling System (CCS) (Section 9.2) is provided to transfer heat from the Reactor Coolant System, reactor support equip-ment and engineered safety equipment to a Seismic Category I Nuclear Service Water System (NSW) (Section 9.2). The CCS serves as an intermediate system and thus a , barrier between potentially or normally radioactive fluids and the lake / pond water which flows in the NSW System. The CCS consists of two independent engineered safety subsystems, each of which is capable of serving all necessary loads under normal or accident conditions. In addition to serving as the heat sink for the CCS, the NSW System is also used as heat sink for the containment and engineered safety equipment through use of compartment and space coolers. The NSW System con-sists of two independent loops, each of which is capable of

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. Regulation :

4 (10 *C3R) Compliance

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providing all necessary heat sink requirements. .The NSW <

System transfers' heat to.the. ultimate heat sink (Section

9. :2) . Electric power is . discussed in Chapter 8. ,

' i GDC 45 .As indicated.in FSAR Section~3.1, the integrity and capabi-lity of the component cooling water system (Section 9.2).and i nuclearlservice water. system (Section 9.2) are monitored

. -during normal operation by alternating operation of the.

l systems.between1the redundant system components. Nonsafety i related systems may be isolated temporarily for inspection.

All major-conponents' will be visually inspected on a periodic basis. The component cooling' and nuclear service water pumps

are arranged such that any pump.may be isolated for inspection

[

and maintenance while maintaining full plant operational capa-

.bilities.

GDC 46 As' described in FSAR Section'3.1, the cooling water systems.

1 are pressurized during plant operations; thus, the structural j - and leak 6ight integrity of each system and the operability and performance of their active components are continuously demon-i- -strated. In addition, normally idle portions of the piping

!~ system and idle components are tested .during pisnt shutdv;n.

l The emergency functions of the systems are periodicallv tested +

.out to the final actuated device.

j.

• For details, see the write-ups on Electric Power (Chapter 8),

, Component Cooling Water (Section 9.2), Nuclear Service Water.

l (Section 9.2),'and. Instrumentation and Controls (Chapter 7).

l GDC 50 FSAR Section 3.1 indicates that the-containment structure,' .

including access openings and penetrations, is= designed with sufficient conservatism to accommodate, without exceeding the O design leakage rate, the transient peak pressure and tempera-

, ture associated with a postulated' reactor coolant piping Lbreak-up and including a double-ended rupture of the largest .

j. . reactor coolant pipe. Containment design basis is discussed

further in Sections ~3.8-and 6.2.

i iGDC 51- 'As; discussed in FSAR Section 3.1, the design condition for the containment pressure boundary is based on the parameters de-i

. rived from the-design basis accident. . For this; design condi-

. tion, the steel liner material behaves in:a nonbrittle manner

! . and has the . capability- to minimize the propagation of any ' unde-

, .tected. flaw; Detailed information on the steel liner material

~

'is found in Section 3.8.2. Additional information was sub-

. mitted by letter dated September. 10, 1980.

l GDC 52 As indicated.in FSAR 3.1, the. containment design permits over-

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pressure strength testing during' construction and permits pre--

operational integrated leakage rate testing at calculated peak f

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. (10'CFR) Compliance l . accident pressure and at reduced pressure, in accordance with Appendix J 10CFR50. The containment and other equipment which may be subjected to containment test conditions are designed

so that periodic integrated leakage rate testing can be con-

ducted at calculated peak accident pressure. The preoper-ational integrated leak tests at peak pressure verify that the containment, including.the isolation valves and the resilient' penetration seals, leaks less than the. allowable value of 0.25 weight percent per day at peak accident

}; ' pressure. Details concerning the conduct of periodic inte-

} grated leakage rate tests are in Section 6.2.1.4.

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GDC 53 FSAR Section 3.1 indicates that the containment and the I containment isolation system (Section 6.2) are designed so that: (1) integrated leak tests can be run during a plant lifetime (see compliance to Criterion 52), (2) visual inspections can be made of all important areas, such as pene-

! trations, (3) an appropriate surveillance program can be main-a tained (Section 6.2),-(4) periodic testing at containment j peak accident pressure of the leaktightness ' of isolation valves and penetrations which have resilient seals and ex-2 pansion bellows is possible, and (5) the operability of

the containment isolation system can be demonstrated peri-

[ odically. .In' testing locally the resilient seals and expan-l sion bellows leakages, the guidelines for Type B tests in Appendix J of 10CFR50 will be followed.

GDC 54 As described in FSAR Section 3.1, the containment isolation features are-classified as Seismic Category I. These com-ponents require quality assurance measures which enhance

, reliability. The containment isolation design provides for a double barrier at the containment penetration in those fluid systems that are not required to function following a design basis event. All piping systems penetrating the con-4 tainment, in so far as practical, have been provided with tests vents and test connections or have other provisions t- to allow periodic leak testing as required. Section 6.2.4.4 has further details on testing. See Section 6.2.4. for general containment isolation details.

l GDC 55 As indicated in FSAR Section 3.1, the reactor coolant pres-I' sure boundary is defined as those piping systems and compo-nents which contain reactor ccolant at design pressure and 4

temperature. With the exception of the reactor coolant sampling lines, the entire reactor coolant pressure boundary, as defined above, is located entirely within the containment structure. All sampling lines are provided with remotely operated valves for isolation in the event of a failure.

These valves also close automatically on a containment iso-lation signal. All other piping and components which may contain reactor coolant are low pressure, low temperature i

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Regulation (10 CFR) Compliance systems which would yield minimal environmental doses in the event of failure. The sampling system and low-pressure sys-tems are described in Section 9.3. An analysis of manfunc-tions in these systems is included in Chapter 15.

GDC 56 As. indicated in FSAR Section 3.1, at least two barriers are provided between the atmosphere outside the containment and'

the containment atmosphere, the reactor coolant system, or l closed systems which are assumed vulnerable to accident

[ forces. Redundant valving is provided for piping that is open to the atmosphere and to the containment atmosphere.

Additional details can be found in Section 6.2.4.

GDC 57 FSAR Section 3.1 indicates that those lines that penetrate the containment, do not commuicate with either the reactor coolant pressure boundary or the containment atmosphere, and l

are not affected by loss-of-coolant accident forces are de-

, fined as closed systems. All lines penetrating the contain-l ment are designed to meet GDC Criterion 57.

GDC 60 As described in FSAR Sectiun 3.1, provisions for liquid, gaseous, and solid radioactive waste processing is provided.

The principles of filtration, demineralization, evaporation, solidification and storage for decay are utilized as des-ibed in Sections 11.2, 11.3, and 11.5. Process monitoring l

ts provided to control this equipment and regulate releases l to the environment as described in Section 11.4.

l l GDC 61 FSAR Section 3.1 indicates that systems which may contain j radioactivity are designed to ensure adequate safety under L normal and puntulated accident conditions. Components are l designed and located such that appropriate periodic inspec-l tion and testing may be performed. All areas of the plant I

are designed with suitable shielding for radiation protec-

, . tion based on anticipatec* radiation dose rates and occupancy l as discussed in Section _2.1. Individual components which j contain significant radioactivity are located in confined

areas which are adequately ventilated through appropriate

! filtering systems. The spent fuel cooling systems provide cooling to remove residual heat from the fuel stored in the

! spent fuel pool. The system is designed for testability to permit continued heat removal. The spent fuel pool is de-signed such that no postulated accident could cause excessive

loss of coolant inventory. Radioactive waste treatment sys-tems are located in the auxiliary building, which contains

! or confines leakage under normal and accident conditions.

The auxiliary building ventilation system includes char-coal filtration which minimizes radioactive material release associated with a postulated spent fuel handling accident.

Fuel storage and handling is discussed in Section 9.1 and radioactive waste' management in Chapter 11.

-Regulation (10'CFR) Compliance GDC 62 As noted in Section 3.1, the restraints and interlocks pro-vided for safe handling and storage of new or spent fuel are discussed in Section 9.1. The center-to-center distance between the adjacent spent fuel assemblies is sufficient to ensure suberiticality, even if unborated water is 2 sed to fill the spent fuel storage pool. The design of the spent fuel storage rack assembly is such that it is impos-sible to insert the spent fuel assemblies in other than pre-scribed locations, thereby preventing any possibility of accidental criticality. Layout of-the fuel handling area is such that the spent fuel ca ks will never be required to traverse the spent fuel storage pool during removal of the spent fuel assemblies.

GDC 63 FSAR Section 3.1 and Chapters 9, 11, and 12 describe-the monitoring capability in the fuel storage and waste handling areas and indicates that the operator will take appropriate sctions if an alarm from any of these monitors is received.

GDC 64 FSAR Section 3.1 indicates the facility contains means for monitoring the containment atmosphere and all other impor-tant areas during both normal and accident conditions to de-tect and measure radioactivity which could be released under any conditions. The monitoring system includes area gamma monitors, atmospherie monitors and liquid monitcrs with full indication in the control room. Alarms are provided to warn of high activity. Section 11.4 discusses the process and effluent and area radiological monitoring systems. Section 11.6 describes the offsite monitoring program.

Appendix B Chapter 17 of the FSAR describes in detail the provisions of  !

the quality assurance program which has been implemented to  !

meet all applicable requirements of Appendix B.

Appendix C This Appendix provides a guide for establishing the applicant's financial qualificatioa. Duke Power Company's financial quali-fications were fully litigated before the Atomic Safety and Licensing Board, and the Board expressly found that Duke Power Company had satisfied the burden of proving that it has rea-sonable assurance of having the funds that it needs to operate the facility in compliance with the Commission's regulations.

Appendix D This Appendix has been superseded by 10 CFR Part 51. As noted in the discussion for 10 CFR 50.40(d), the require- l ments of Part- 51. have been satisfied.

4 Appendix E This Appendix specifies requirements for emergency plans. An emergency plan was prepared for McGuire Nuclear Station in accordance with the provisions of this Appendix. This Emer-gency Plan was reviewed by the NRC staff and the staff found 1 that the emergency plan provided reasonable assurance that

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Regulation (10'CFR) Compliance appropriate measures can and will be taken in the event of

-an emergency to protect public health and safety and prevent damage to property.

In response to new criteria for emergency planning developed subsequent to the~ event at Three Mile Island unit 2, the emergency plan has been extensively modified and improved.

This revised plan, which meets the criteria in NUREG-0654

~has been submitted to the NRC Staff.

Appendix F This Appendix applies to fuel reprocessing plants and related waste management facilities, not to power reactors and is therefore not applicable to this proceeding.

Appendix G' Fracture toughness requirements of this Appendix and program requirements given in Appendix H form the basis for Technical Specification surveillance requirements dealing with the use of surveillance specimens. Additional information to demon-strate compliance can be found in FSAR Chapter 5, Section 5.4.3.7, concerning the irradiation surveillance program.

Heatup and cooldown limits consistent with the equirements of this Appendix are established in the Technical Specifi-cations. Several exemptions to Appendix G were granted by the NRC Staff. These are discussed in detail in Supplement 2 to the SER.

Appendix H Reactor vessel material surveillance program requirements are delineated in this part. Technical Specifications operating procedures have been established to implement these requirements with a further requirement to update the "heatup" and "cooldown" curves provided in the Technical l Specifications. Further information is provided in FSAR.

Chapter 5, Section 5 2.4.3. One exemption to Appendix H was granted by the NRC Staff. This is discussed in detail in Supplement 2 to the SER.

Appendix I This Appendix provides numerical guides for design objectives and ILkiting conditions for operation to meet the criteria "as low as is reasonably achievable" for madioactive material in light-water-cooled nuclear power reactor efflue.ats. Duke Power Company filed with the Commission in June 4, 1976, and February 7,1977, the necessary information to permit an 1

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evaluation of McGuire Nuclear Station respect to the require-ments of Sections II. A, II.B, and II.C of Appendix I. In this submittal Duke Power Company provided the necessary information to show conformance with the Commission' September 4,1975 amendment to Appendix' I rather than perform a detailed j cost-benefit analysis required by Section II.D of Appendix I. 1

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Regulation (10'CFR) Compliance Appendix J Reactor containment leakage testing for water cooled power reactors is delineated in this Appendix. These requirements are given in Technical Specifications 3/4.6.1.2, 3/4.6.1.3, and 3/4.6.1.6. Additional information concerning compliance can be found.in FSAR Chapter 3, Sections 3.1, 3.8.2.7. and FSAR Chapter 14, Section 14.1.3.

Appendix K This Appendix specifies features of acceptable ECCS evalua-tion models. As noted above for 50.46, the analysis for McGuire Nuclear Station has been conducted using a model which has been accepted by the Commission staff as meeting the requirements of this Appendix.

Appendix L This Appendix covers information requested by the Attorney

. General for anti-trust review of license applications. As noted above, the ' anti-trust review for McGuire Nuclear Station took place at the construction permit stage.

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Appendix M This Appendix covers standarization of design and is not applicable to McGuire Nuclear Station.

Appendix N This Appendix covers standardization of nuclear power plant designs and is not applicable to McGuire Nuclear Station.

Appendix 0 This Appendix covers standardization of design and is not applicable to McGuire Nuclear Station. ,

Appendix P This Appendix is proposed, 39 Fed. Reg. 26293, and it applies to fuel reprocessing plants. Accordingly, it is not applica-ble to McGuire Nuclear Station. ,

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Appendix Q This Appendix governs preapplication early review of site <

suitability issues and is not applicable to McGuire Nuclear l Station.

I Appendix Q This Appendix is proposed, 39 Fed. Reg. 26297, and it would I (Proposed) apply to fuel reprocessing plants, not power reactors. l 1

100.1 This regulation is explanatory and does not impose indepen-dent obligations on licensees.

100.2 This regulation is explanatory. McGuire Nuclear Station is

, not novel in design and is not unproven as a prototype or pilot plant.

100.3 This regulation is explanatory and does not impose indepen-dent obligations on licensees.

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. w Regulation (10 CFR) Compliance 100.10 The factors listed related to both the unit design and the site have been provided in the application. Site specifics, including seismology, meteorology, geology, and hydrology, are presented in Chapter'2 cf the FSAR. The exclusion area, low population zone, and population center distance are pro-vided and described. The FSAR also descrioes the character-istics of reactor design and operation.

100.11 lui exclusion area has been established, as described in FSAR Section 2.1.2.2. The low population zone required by 100.11 (a)(2) has been established, as described in FSAR Section 2.1.3.3, as the area within a radial distance of 5.5 miles from the centerline of the reactors. As indicated in Sec- 1 tion 2.1.3.5, the nearest population center, as defined by 10 CFR 100.3(c), based on the 1970 census, is Charlotte, North Carolina, which is 11 miles south-southeast of the site.

The FSAR accident analyses, particularly those_in Chapters 6 and 15, demonstrate that offsite deses resulting from postulated ~ accidents would not exceed the criteria in this section of the regulation.

Appendix A Appendix A to 10 CFR Part 100 provides seismic and geologic siting criteria for nuclear power plants. The compliance of

- the McGuire Nuclear Station site with this Appendix is dis-cussed in the McGuire Nuclear Station Safety Evaluation Re-Port, NUREG-0422.

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