ML19344D901
| ML19344D901 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/19/1980 |
| From: | Mills L TENNESSEE VALLEY AUTHORITY |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8008260293 | |
| Download: ML19344D901 (35) | |
Text
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i TENNESSEE VALLEY AUTHORITY CHATTANOOGA, TENNESSCE 37401 400 Chootnut Street Tower II August 19, 1980 Director of Nuclear Reactor Regulation Attention: :Mr. Schwencer, Chier Licensing Branch No. 2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555 Dear Hr. Schwencers In the Matter of the Application of
)
Docket Nos. 50-327 Tennessee Valley Authority
.)
50-328 This letter provides the information you requested.from TVA during the telephone conversation of today. This letter explains where in our license application each regulation in 10 CFR Parts 20, 50, and 100 is addressed.
Enclosed is an itemized review of the Sequoyah Nuclear Plant unit 1 compliance with all applicable 10 CFR regulations. TVA believes that Sequoyah units 1 and 2 do comply with the applicable regulaticns except
-in those osses where specific exemptions have been justified and approved by the staff. lie base our' confidence in this conclusion on the references in the enclosure, plus the lengthy review and licensing process that the Sequoyah Nuclear Plant has undergone. The design proccas of our own employees, the quality assurance progr1une of TVA and the NSSS vendor, the independent review of the NRC staff and ACRS, and the additional independent rinview of the Atomic-Safety and Licensing Boards provide reasonable assuranoc that the public health and safety. will be protected.
If I can be of Ibrther assistance, please get in touch with me at FTS 857 -2778.
Very truly yours, TENNE 3SEE VALLEY AUT!iORITY h.
L. M. Mills,.anager Nuclear Regulation and Safety Enclosure Dworn to and subscribed before me Ma day of 0
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ENCLOSURE 4
COMPLIANCE OF SEQUOYAH NUCLEAR PLANT UNIT 1 WITH THE NRC REGULATIONS OF 10 CFR PARTS 20, 50, AND 100 Regulation (10 CFR)
Comoliance 20.1(a)
This regulation merely states the general purpose fcc which the Part 20 regulations are established and does not impose any independent obligations on licensees.
20.1(b)
This regulation describes the overall purpose of the Part 20 regulatioons to control the possession, use and transfer of licensed material by any licensee, such that the total dose to an individual will not exceed the standards prescribed therein.
It does not impose any independent obligations on licensees.
20.1(c)
Conformance to the ALARA principle stated in this regulation is ensured by the implementation of TVA policies and appropriate Technical Specifications and health physics procedures. Chapters 11 and 12 of the FSAR describe the specific equipment and design features utilized in this effort.
20.2 This regulation merely establishes the applicability of the Pert 20 regulations and imposes no independent obligations on those licensees to which they apply.
20.3 The definitions contained in this regulation are adhered to in all appropriate Technical Specifications and procedures, and 11. applicable sections of the FSAR.
20.4 The Units of Radiation Dose specified in this regulation are accepted and conformed to in all applicable SONP procedures.
20.5 e Units of Radioactivity specified in this regulation are accepted and conformed to in all applicable SONP procedures.
20.6 This regulation governs the interpretation of regulations by the NRC and does not impose independent obligations on licensees.
20.7 This regulation sives the address of the NRC and does not impose independent obligations on licensees.
Rrgulation (10 :FR)
Compliance 20.101 The radiation dose limits specified in this regulation are complied with through the implementation of and adherence to administrative policies and controls and appropriate health physics procedures developed for this purpose.
Conformance is documented by the use of appropriate personnel monitoring devices and the maintenance of all required records.
20.102 When required by this regulation, the accumulated dose for any individual permitted to exceed the exposure limits specified 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 implementation of appropriate health physics 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 radioactive materials. The systems and equipment described in Chapters 11 and 12 of the FSAR provide the capability to minimize these bazards.
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. When necessary, as determined by SQNP administrative guidtlines, additional precautionary procedures are utilized to limit the potential for intake of radioactive materials.
20.103(c)
The SONP Respiratory Protection Procedure implements the requirements of this regulation by ensuring the proper use of approved respiratory protection equipment. The SQNP Respiratory Protection Procedure incorporates fully the stipulations 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 February 7, 1979.
20.103(f)
The Respiratory Protection Program is in full conformance with the requirements of 20.103(c).
I I
20.104 Conformance with this regulation is assured by appropriate TVA policies regarding employment of individuals under the age of 18 and the SQNP Health Physics Manual restricting these individuals' access to restricted areas.
Fegulation (10 CFR)
Compliance 20.105(a)
Chapter 11 of the FSAR provides the information and related radiation dose assessments specified by this regulation.
20.105(b)
The radiation dose rate limits specified in this regulation are complied with through the implementation of SQNP procedures, Technical Specifications, and administrative policies which control the use and transfer of radioactive materials.
Appropriate surveys and monitoring devices document this compliance.
20.106(a)
Conformance with the limits specified in this regulation is assured through the implementation of SONP 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)
TVA has not and does not intend to include in any license 20.106(c) 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 radioactive effluents are made in conformance with this regulation, and are described in detail in Chapter 11 of the FSAR, and in appropriate reports required by the Technical Specifications.
20.106(s)
This regulation provides criteria by which the Commission may impose further 1:mitations 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 sanitary 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 SONP to determine exposure of individuals to concentrations of radiocctive materials.
Appropriate health physics procedures and administrative policies implement this requirement.
20.201 The surveys required by this regulation are performed at.. -. -
R:gulation (10 CFR)
Compliance t
adequate frequencies and contain such detail as to be consistent with the radiation hazard being evaluated. When necessary, the Radiation Special 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 SONP Health Physics Manual and applicable health physics procedures require these surveys and provide for their documentation in such manner as to ensure co=pliance with the regulaticns of 10 CFR Part 20.
20.202(a)
The SQNP Health Physics Manual and applicable health physics procedures set forth policies And practices which ensure that all individ::als are supplied with, and required to -
use, appropriate personnel monitoring equipment.
The Radiation Special 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 eccepted and conformed to in all applicable SONP procedures, Technical Specifications, and those portions of the SCNP Health Physics Manual in which its use is made.
20.203(a)
All materials used for labeling, posting, or otherwise designating radiation hazards or radioactive materials, and using the radiation symbol, conform to the conventional design prescribed in this regulation.
20 303(b)
This regulation is conformed to through the implementation of appropriate health physics procedures and portions of the 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" are conformed to by the implementation of the Technical Specifications and appropriate plant health physics procedures, as well as the SONP Health Physics Manual. The controls and other protective measures set forth in the regulation are maintained under the surveillance of the SQNP Health Physics group.
It should be noted that Technical Specification 6.12.1 provides alternate access control methods to be applied "in lieu of the ' control device' or ' alarm signal' required by paragraph 20.203(c)(2) of 10 CFR 20," which will prevent unauthorized entry into a high radiation area.
20.203(d)
Each Airborne Radioactivity Area, as deri 1 in this regulation, is required to be posted by provisions of the Health Physics Manual and apprcpriate health physics procedures. These procedures also provide for the
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Regulation (10 CFR)
Coeoliance surveillance require =ents necessary to deter =ine airborne radioactivity levels.
20.203(e)
The area and room posting requirements set forth in this regulation pertaining to radioactive =aterials are complied with through the implementation of appropriate health physics procedures, and portions of the SCNP Health Physics Manual.
20.203(f)
The container labeling require =ents set forth in this regulation are co: plied with through the implementation of appropriate health physics procedures, and portions of the SQNP Health Physics Manual".
20.204 The posting require =ent exceptions described in this regulation are used where appropriate and necessary at SONP.
Adequate controls are provided within the SONP health physics procedures to ensure safe and proper application of these exceptioas.
20.205 All of the requirements of this regulation pertaining to procedures for picking up, receiving, and opening packages of radioactive caterials are implemented by the SONP Heal *,h Physics Manual and appropriate health physics procedures.
These procedures also provide for the necessary documentation to ensure an auditable record of co=pliance.
20.206 The requirements of 10 CFR 19.12 referred to by this regulation are satisfied by the Radiological Hygiene Orientation conducted at SONP.
Appropriate health physics procedures set forth requirements for all radiation workers to receive this instruction on a periodic basis.
20.207 The storage and control require =ents for licensed =aterials in unrestricted areas are conforced to and documented through the i=plementation of SCNP health physics procedures and applicable portions of the SQNP Health Physics Manual.
20.301 The general requirenents for waste disposal set forth in this regulatien are co: plied with through SONP surveillance instructions, the Technical Specifications, and the provisivas of the statf en license. Chapter 11 of the FSAR describes the Solid Wat ta Disposal Systen installed at SONP.
20.302 No such application for proposed disposal procedures, as described in this regulation, has been made er is conte = plated by TVA.
20.303 No plans for waste disposal by release into sanitary sewerage systens, as provided for in this regulation, are contemplated by SONP, nor is this practice currently utilized.
R;gulation (10 CFR)
Compliance 20 304 Disposal of wastes by burial in soil (i.e., onsite burial),
as provided for in this regulation, is not performed or being contemplated by SQNP.
20.305 Specific authorization, as described in this regulation, is not currently being sought by TVA for treatment or disposal of wastes by incineration.
20.401 All of the requirements of this regulation are complied with through the implementation of appropriate Technical Specifications and health physics procedures pertaining to records of surveys, radiati.on monitoring and waste disposal.
The retention periods specified for such records are also provided for in these specifications and procedures.
20.402 SQNP has established an appropriate inventory 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 regulations of 10 CFR in the Technical Specifications.
20.403 Notifications of incidents, as described in this regulation, are assured by the requirements of the Technical Specifications, the SONP Health Physics Manual and appropriate plant procedures, which also provide for the necessary assessments to determine the occurrence of such incidents.
20.404 This regulation was deleted effective September 17, 1973 (38 Fed. Reg. 22220).
20.405 Reports cf overexposures to radiation and the occurrence of excessive levels and concentrations, as required by this regulation, are provided for by reference in the Technical Specifications and in appropriate health physics procedures.
20.406 Th!s regulation was deleted August 17, 1973, effective September 17, 1973 (38 Fed. Reg. 22220).
20.407 The personnel monitoring report required by this regulation is expressly provided for by the Technical Specifications.
Appropriate health physics procedures establish the data base from which this report is generated.
20.408 The report of radiation exposure required by this regulation upon termination of nn individual's employment or work assignment is generated through the provisions of TVA procedures.
20.409 The notification and reporting requirements of this regulation, and those referred to by it, are satisfied by the provisions of TVA procedures.
R:gulaticn
_ 10 CFR)_
Compliance
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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 hasard to life or property.
It does not impose independent obligations on licensees.
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 inpose independent obligations on licensees.
20.601 This regulation describes the remedies which the Co==ission may obtain in order to enforce its regulations, and sets forth those penalties or punish =ents which may be imposed for violations of its rules.
It does not impose any independent obligations on licensees.
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Regulation
'( " CFR)
Compliance 50.1 This regulation states the purpose of the Part 50 regulations and does not impose any independent obligations on licensees.
50.2 This regulation defines various terms and does not impose independent obligations on licensees.
50.3 This regulation governs the interpretation of the regulations by the NRC and does not impose independent ogligations on licensees.
"50.4 This regulation gives the address of the NRC and does not impose independent obligations on licensees.
50.10 These regulations specify the types of activities that may 50.11 not be undertaken without a license from the NRC. TVA does not propose to conduct any such activities at SQNP without an NRC license.
50.12 This regulation provides for the granting of exemotions 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 ogligations 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 requirements 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.
TVA has substantially complied with the procedural requirements 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 TVA has submitted a Final Environmental Statement covering SONP.
50 31 These regulations merely permit more efficient organization 50.32 of the license application and _mpose no independent obligations on licensees.
_R.
50.33 This r gulction requir:s tha license cpplication to contain certain general information, auch as an identification of the applicant, information about the applicant's financial qualifications, and a list of regulatory agencies with jurisdiction over the applicant's rates and services. This inforsrtion was provided in SQNP operating license atplicatien.
50.33a This regulation requires aoplicants for construction per=its to submit information required for antitrust review. The antitrust review required by the Atomic Energy Act of 1954, as amended, was performed at the construciton permit Stage.
50.34(a)
This regulation governs the contents of the Preliminary Safety Analysis Report and is relevant to the construciton permit stage rather than the gperating license stage.
50.34(b)
A Final Safety Analysis Report (FSAR) has been perpared and submitted, which addresses in the chapters indicated the inforamtion 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 (4) design and performance evaluation - ECCS performance is discussed and shown to meet the requirements of 10 CFR 50.46 in Chapters 6 and 15.
($) results of research programs - Chapter 1 (6)
(i) organizational structure - Chapter 13 (ii) managerial and administrative controls -
Chapters 13 and 17.
Chapter 17 discusses compliance with the quality assurance requriements of Appendix B.
(iii) plans for preoperational testing and initial operations - Chapter 14 (iv) plans for conduct of normal operations - Chapter 13 and 17.
Surveillance and periodic testing is specified in the Technical Specifications.
(v) plans for coping with emergencies - Emergency Plan (Chapter 13)
(vi)
Technical Specifications - prepared in conjunction with the Staff (Chapter 16).
(vii) not cpplic:blo, sinos the op;rsting lic;nsa application was filed before February 5,1979 (7) technical qualifications - Chapter 13 (8) operator requalification program - Chapter 13 and letters of 10-31-79, 7-1-80, and 7-31-80. The latter letters describe a requalification program meeting the new requirements of NUREG-0660.
50 34(c)
A physical security plan has been prepared and is implemented for Sequoyah Nuclear Plant unit 1.
50 34(d)
A safeguards contingency plan for TVA has been prepared and is implemented for SQNP.
50 35 This regulation is relevant to the construction permit stage rather than the operating stage.
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.
50 36(a)
The Environmental Technical Specifications, Part I, include specifications which require compliance with 10 CFR 50 34a (releases as low as is reasonably achievable), and that ensure that concentrations of radioactive effluents released to unrestricted areas are within the limits specified in 10 CFR 20.106. The reporting requirements of 10 CFR 50 36a(a)(2) are also lancluded in these specifications.
50.37 This regulation requires the applicant to agree to limit access to Restricted Data. TVA's agreement to do so is the operating license application for Sequoyah Nuclear Plant units 1 and 2.
50.38 This regulation prohibits the NRC from issuing a 11. nse to foreign-controlled entities. TVA's statement tLit it is not owned, controlled, or dominated by an alien, foreign corporation, or foreign government is in the operating license application for Sequoyah Nuclear Plant.
50 39 This regulation provides that applications and related documents may be made available for public inspection.
This imposes no direct obligations on applicants and licensees.
50.40 This regulation provides considerations to " guide" the Commission in grating licenses as follows:
50.40(a)
The design and operation of the facility is to provide.
50.40(c)
Tha d; sign cnd Cper2 tion of tha fccility in to provida 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 endangered. The basis for TVA's assurance that the regulations will be met and the public protected is contained fn this enclosure and in the license application and the related correspondence over the years. Moreover, the lengthy process by which the plant is designed, constructed, and reviewed, including reviews by TVA's own staff, the NRC staff, the ACRS, and NRC licensing boards, provides a great deal of assurance that the public health and safety will not be endangered.
In particular, the Atomic Safety and Licensing Board, after an extensive review, concluded that T'f A had, the commitment and technical qualifications necessary to operate Sequoyah Nuclear Plant safely and in compliance with all applicable radiological health and safety requirements (see below).
50.40(b)
Another consideration is that the applicant te technically and financially qualified. Both TVA's technical qualifications and its financial qualifications were reviewed in hearings before the Atomic Safety and Licensing Board.
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 individual showings of compliance with particular regulations contained in this enclosure, as well as the contents of the entire FSAR and related correspondence over the years, plus the lengthy process of design, construction, and review by TVA, its NSSS vendor, and the government, provide TVA 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 TVA has an approved security plan for Sequoyah Nuclear Plant that TVA is not controlled by agents of foreign countries, and that TVA has agreed to limit access to Restricted Data (see above).
50.40(d)
The final 50.40 " consideration ~ is that the applicable requirements of Part 51 have been satisfied. Part 51 concerns compliance with the National Environmental Policy Act of 1969. TVA has submitted a Final Environmental Statement, the AEC staff has reviewed and accepted the TVA Environmental Statement pursuant to 10 CFR 50, Appendix D.
Environmental Technical Specifications have been issued for Sequoyah Nuclear Plant.
50.41 This regulation applies to class 104 licensees, such as those for devices used in medical therapy. Sequoyah Nuclear Plant has not applied for a class 104 license, and so 50.41
is nst applicible.
50.42 Section 50.42 provides additional " considerations" to
" guide" the Commission in issuing Class 103 licenses.
The two considerations 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 advice 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 changing, TVA periodically makes load projections, and in TVA's judgnent the need for Sequoyah Nuclear Plant 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 substantially 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.
50.44 The Sequoyah Nuclear Plant combustible gas control system is described in FSAR Section 6.2.5.
The system is designed to maintain the hydrogen concentration in containment at a safe level following a LOCA, without purging the containment atmosphere, as specified in 10 CFR 50.44(e).
The system consists of internal recombiners,'two hydrog'en analyzers, and a hydrogen skimmer system. The containment recirculation system and purge system complement the recombiner system. The Sequoyah Nuclear Plant system neets the requirements of NUREG-0660 and NUREG-0694. The requirements of 10 CFR 50.44 are satisified.
50.45 This regulation provides standards for construction permits rather than operating licenses and is therefore not material to this operating license proceeding.
50.46 FSAR Sections 6.3 and 15.4.1 describe the Emergency Core Cooling System and the methods used to analyze ECCS performance following a postulated loss of coolant accident.
By letter dated July 31, 1980, TVA provided the results of a LOCA-ECCS analysis for Sequoyah Nuclar Plant using an NRC approved evaluation model, which is in compliance with Appendix K to 10 CFR 50.
The analysis, based on an overall peaking factor (Fq) of 2.237, provided results in.
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complianca with tha critcria of 10 CFR 50.46(b). Tha Fq limit will be reflected in the Technical Specifications.
50.50 This regulation provides that the NRC will issue a license upon determining that the application meets the standards and requirements of the Atomic Energy Act and the regulations and that the necessary notifications to other agencies or bodies have been duly made.
It imposes no direct obligations on licensees.
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 nu=ber of activities. It imposes no independent obligation on the licensee.
50.53 This regulation provides that licenses are not to be issued for activities that are not under or within the jurisdiction of the United States. The operation of Sequoyah Nuclear Plant will be within the United States and subject to the jurisdiction of the United States, as is evident from the description of the facility in the operating license application.
50.54 This regulation specifies certain conditions that are incorporated in every license issued. Compliance is effected simply by including these conditions in the license when it is issued, and in fact the low-power license already in effect for Sequoyah unit 1 expressly says that it is subject to the conditions in Section 50.54 of 10 CFR Part 50 (see paragraph 2.D of Sequoyah unit 1 license for fuel loading and low-power testing).
Indeed, much of 50.54 merely provides that other provisions of the law apply, which would be the case even without 50.5h.
50.55 This regulation addresses conditons 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 information on testing of safety-related systems.
Chapter 17 provides information concerning 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.
I 50.55a(a)(2)
This paragraph is a general paragraph leading into paragraphs (c) through (i) of the regulation..
50.55a(b)(1)
Thssa paragraphs provida guidanca concarning tho approvsd 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 ASME Section III (1968) Class A.
Information can be found in Chapter 5.
50.55a(d)
Reactor coolant system piping meets the requirements of USAS B31.1.
Information can be found in Chapter 5 50.55a(e)
Reactor Coolant pumps meet the requirements of ASME Section III (1968) Class A.
Information can be found in Chapter 5.
50.55a(f)
Noting the construction permit date of May 1970, the valves within the reactor coolant system pressure boundary were designed and fabricated in accordance with the requirements of ANSI B 16.5, MSS-SP-66, and ASME Section III, 1968 edition.
50.55a(g)
Inservice Inspection (ISI) requirements are delineated in this part and are specified in the Technical Specifications, paragraph 4.0.5.
As permitted by this part and the Technical Specifications, certain exemptions have been requested and granted for the inservice inspection of various systems and the inservice testing of various pumps and valves. By letters dated 8-23-77 and 12-6-78, the Sequoyah Nuclear Plant Preservice and Inservice Inspection Program was docketed. Requests for relief have been incorporated into the FSAR, see o 5.9 In Supplement 1 of the Staff's SER for Sequoyah Nuclear Plant, Section 5.2.6 provides a discussion of preservice inspection and inservice testing of pumps and valves.
Additional information on ISI can be found in FSAR Chapter 5.
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 information is given in FSAR Section 5.4.3.6 concerning the irradiation surveillance program.
50.55b This regulation has been revoked. 43 Fed. Reg. 49775.
50.55e This regulation is only proposed 39 Ped. Reg. 26297, and applies to fuel reprocessing plants.
50.56 This regulation provides that the Commission will, in the
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absence of good cause shown to the contrary, issue an operating license upon completion of the construciton of a facility in compliance with the terms and conditions of the construction permit.
This i= poses no independent obligations on the applicant.
50.57(a)
This regulation requires the co= mission to =ake certain findings before the issuance of an operating license. These findings have already been =ade for Sequoyah Nuclear Plant in connection with the issuance of license DPR-77 for fuel loading and low-power testing, and they can also be made for full power operation for the reasons given in this enclosure generally. Specifically:
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(1) Construction of the facility has been substantially co=pleted in conformity with the construction permit and the applicatien as amended. Conferr.ance of the facility to the NFC rules and regulations and the Act, as i=oleJnented by the regulations, has been demonstrated by the application.
(2) The Technical Specifications and resulting operating procedtres provide assurance that the unit will operate in conformity with the application as a= ended and with the rules and regulations, with the noted exceptions to 10 CFR 50, as documented in License DPR-77.
(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 TVA is technically and financially qualified to operate the unit.
(5) The applicable provisions of 10 CFR 140 have been satisfied.
(6) The approved 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 completed on a schedule acceptable to the Co==ission.
Regulation (10CFR)
Compliance 50.57(c)
This regulation provides for a low-power testing license.
Such a license has alreadv been issued for SONP.
50.58 This regulation provides for the review and report of the Advisory Committee on Reactor Safeguards. The ACRS has reviewed the operating license application for SQNP in accordance with its usual practice.
50.59' This regulation provides for the licensing of certain changes, tests, and experiments at 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 the SQNP.
TVA has provided office space in accordance with the requirements of this section. TVA 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 regulation and the license. Section (e) requires that the FSAR be updated by July 22, 1982, 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 regulation.
50.80 This regulation provides that licenses may not be transferred without NRC consent. No application for transfer of a license is involved in the SONP proceeding.
50.81 This regulation permits the creation of mortgages, pledges, and liens on licensed facilities, subject to certain provisions. TVA licensed facilities are not mortgaged, pledged, or otherwise encumbered as those terms are used in this regulation.
50.82 This regulation provides for the termination of licenses.
It does not apply to SQNP because TVA has not requested the termination of a license..
-,c y
Ecgulation (10CFR)
Compliance 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, 50.101 and modification of licenses by the Commission under 50.102 unusual circumstances. No such circumstances are present 50.103 in the SQNP 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 regulation imposes no independent obligations on a licensee unless the NRC proposes a backfitting requirement, and so this regulation is not applicable.
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 SONP proceeding, and so this regulation is not applicable.
Appendix A GDC 1 Section 3.1.2.1 of the FSAR describes the design provisions 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 3.1.2 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, and the importance of the safety functions to be performed.
GDC3 FSAR Section 3.1.2.1 describes in general the measures which 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 by November 1, 1980.
This schedule is in accordance with the Commissioner's Order dated May 23, 1980.
Regulation (10CFR)
Compliance GDC 4 FSAR Section 3 1.2.1 describes the design features used to accommodate the effects of and be compatible with the environ = ental 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 most electrical equipment has been adequately demonstrated to be qualified for its expected service environments. Justification in the form of additional evaluations and documentation has been provided to the staff for operation of the unit pending completion of the environmental qualification review in letters dated 6/16/80, 7/28/80, 8/11/80, and 8/18/80.
GDC 5 As described in FSAR Section 3 1.2.1, those structures, systems and components which are shared with Unit 1 are tabulated in FSAR Section 1.2.2.16.
It is concluded that safety functions are not significantly impaired by such sharing.
GDC 10 FSAR Section 3.1.2.2 indicates that the reactor core and associated systems are designed to function throughout the design lifetime without exceeding fuel damage limits, using protection criteria specified in Section 3.1.2.2 and Chapters 4, 7, and 15.
GDC 11 FSAR Section 3.1.2.2 indicates that prompt compensatory reactivity feedback effects are assured by unit design and operational limit considerations. The core inherent reactivity feedback characteristics and reactivity control methods are described in FSAR Section 4.3 CDc 1?
FSAR Section 3.1.2.2 describes the inherent and design features which eliminate or limit the various types of oscillations. Core stability is further described in Section 4.3
Reg' 1ation u
(10CFR)
Comoliance GDC 13 As indicated in FSAR Section 3 1.2.2, and described in more detail in Chapter 7, instrumentation and control systems have ben provided to monitor and maintain plant variables including 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 occurresces, and under, accident conditions.
GDC 14 FSAR Section 3.1.2.2 indicates that the reactor coolant pressure boundary has been designed to accommodate the system temperatures 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 1.2.2, 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 operations and anticipated transients.
The design codes used for the Reactor Coolant System are described in Chapter S.
Details concerning the protection systems are provided in Chapter 7.
GDC 16 As described in FSAR Section 31.2.2 and Chapter 6, an ice condenser containment structure, operating at a subatmospheric pressure, 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 containment and its associated engineered safety features thus meet the required functional capabiity of this criteria.
GDC 17 As described in FSAR Section 3.1.2.2, 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.
GDC 18 As described in FSAR Section 3.1.2.2 and Chapter 8, the redundant electric power systems important to safety are continuously monitored and energized during normal plant operation from redundant offsite power sources.
Redundant onsite diesel generators provide automatic backup pcwer sources. Periodic tests of the diesel generators, the transfer system and the station batteries are made, as required by Technical Specifications.
Regulaticn (10CFR),
Compliance GDC 19 FSAR Section 3.1.2.2 describes the main control room, which contains the controls and instrumentation necessary for safe operation of the unit during normal and accident 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 control roc 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 3.1.2 3 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 3.1.2.3, the protection system is designed for the high functional reliability and inservice testability commensurate with the safety functions to be performed. This section, as well as Chapter 7, describe in detail the design features provided to ensure redundancy and testability.
GDC 22 FSAR Section 3.1.2.3 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 1.2.3, 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.
GDC 24 FSAR Section 3.1.2.3 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 service of any protection system component or channel which is common to the protection and control systems, leaves intact a system satisfying l
all redundancy, reliability, and independence l
requirements of the protection system. Details conderning separation of protection and control systems are provided in Chapter 7.
Rrg*lation u
(10CFR)
Compliance GDC 25 FSAR Section 3.1.2.3 indicates that the protection system has been designed to assure that specified acceptable fuel design limits are not exceeded in the event of any single reactivity control system malfunction, including an acci*iental withdrawal of control cluster groups.
Further details are provided FSAR Sections 4.3 1.4, 7.2.2.2.3, and 7.7.2.2.
GDC 26 As indicated in FSAR Section 3.1.2.3, 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 Sections 4.3.1 5 and 7.7.2.2.
GDC 27 As described in FSAR Section 3.1.2.3, means are provided for shutdown reactivity for cooling the core under any anticipated condition and with appropriate margin for contingencies. -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 cooldown. These means are discussed in detail in FSAR Sections 4.3 and 7.2.
GDC 28 FSAR Section 3.1.2.3 indicates that core reactivity is
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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 l
and the maximum rates of reactivity insertion employing l
control rods are limited to values which prevent rupture of the coolant pressure boundary or disruption of the l
core internals to a degree which would impair core cooling capacity.
Further detalla are provided in Section 4.3 GDC 29 As indicated in FSAR Section 3.1.2.3, the protection and reactivity control systems are designed to assure extcemely high probability of performing their required safety functions in the event of anticipated operational occurrences. The protection system is further discussed in Section 7.2.
The reactivity control systems are l
discussed in Sections 4.2 3 and 7.7.
GDC 30 As described in FSAR Section 3.1.2.4, reactor coolant pressure boundary components are designed, fabricated, inspected, and tested in conformance with ASME Nuclear Power Plant Components Code Section III. Major components are classified as seismic Class 1 and are l
Regulation (10CFR)
Cornliance accorded the quality neasures appropriate to this classification.
The evaluations of reactor coolant pressure boundary components are discussed in Section 5.2.
GDC 31 As indicated in FSAR Section 3.1.2.4, close centrol is maintained over material selection and fabrication for the reactor coolant system to ensure that the boundary behaves in a nonbrittle nanner. 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 Technical Specifications ensure operation within the pressure-temperature limit relative to this criterion.
GDC 32 FSAR Section 3.1.2.4 describes how the design of the reactor vessel and its arrangement in the system provides the capability 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 containment provides sufficient space for insepetion 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 3.1.2.4, the chemical and volume control system provides a means of reactor coolant makeup and adjustment of the boric acid concentration. A high degree of functional reliability and safe response to probable modes of failure is assured by provision of standby components. 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
.m GDC 34 FSAR Section 3.1.2.4 indicates that the residual heat removal 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 linitg. Suitable redundancy is accomplished below 350 F with the two residual heat removal pumps with means available for draining and monitoring of leakage, two heat exchangers, and the associated piping and cabling. The residual heat removal system is able to operate on either onsite or gffsite electrical power. Suitable redundancy above 350 F is provided by the steam generators, auxiliary reed pumps, and attendant piping. Details of the residual heat removal systen design are in FSAR Section 5 3.2..
R;gulation (10CFR)
Compliance GDC 35 FSAR Section 3.1.2.4 describes the use of passive accumulators with two certrifugal charging pumps and two low head safety injection pumps to provide redundancy for failure of any component in any system. The primary
.Jnction of the emergency core cooling systen is to deliver borated cooling uater to the reactor core in the event of a loss-of-coolant acci, dent. 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 preserved.
Further details are provided in Chapters 6 and 7.
GDC 36 As described in FSAR Section 3 1.2h, design provisions are nade for inspection to the extent practical of all components of the emergency core cooling system. An inspection is perforaed periodically to demonstrate system readiness. To the extent possible, the critical parts of the reactor vessel internals, injection nozzles, pipes, valves, and pumps are inspected visually or by boroscopic examination for erosion, corrosion, and vibration wear evidence. Nondestructive inspection is performed where such techniques are dcsirable and approprite. Technical Specifications require inservice inspection in accordance with applicable ASME Codes.
Details of the inspection programs are provided in Chapters 5 and 6.
GDC 37 FSAR Section 3.1.2.4 indicates that the components of the energency core cooling system located outside the contain=ent will be accessible for leaktightness inspection during appropriate periodic tests.
Each active component of the system may be individually actuated on the normal power source at any time during plant operation to demonstrate operability. The centrifugal charging pumps are part of the charging system, and this system is in continuous operation during plant operatons.
Actuation circuits are tested 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 requirements.
GDC 38 As indicated in FSAR Section 3 1.2.4, the containment spray system, ice condenser, and RHR spray systen are provided to remove heat fron the containment following.
B;gulation (10CFR)
Cerplienco
- a loss-of-coolant accident. An air return syste= is used to circulate air and steas through the centainnent after the initial blevdewn. This caintains preper nixinr.
of the centain=ent air and stean with the heat renoval
=edia for the necessary heat re= oval. The less of a single active co=penent was assumed in the design of these systems. E=ergency power system arrangenents ensure the proper functioning of these systems. Two electrical buses, eacn connected to both ensite and offsite power, feed the pu=p motors and the necessary valves. Further details are ;rovided in Sectiens 6.2 and 8.3 GDC 39 As indicated in Secticn 31.2.4, the ice condenser design includes previsions for visual inspections of the ice bed flev char.nels, doers, and cooling equip =ent.
The air return fan syste= provides for visual inspection of the fans and the associated backflev da:pers and for duct systens that are not embedded in concrete. The centainment spray syste= and the residual heat re=cval systen (RHR) sprays are designed such that active and passive cecpenents can te readily inspected to desenstrate syste= readiness.
Pressure centained syste=s are inspected for leaks free pu=p seals, valve packing, flange joints, and relief valves. During cperatienal testing of the centainnent spray punps and RH3 pu=ps, the portiens of the systers subjected to pressure are inspected for leaks. Systen design details are given in Section 6.2.
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Regulation (10CFR)
Compliance GDC 40 As described in FSAR Section 3.1.2.4, the containment heat removal systems described in Section 6.2 are designed to permit periodic testing so that proper operation can be assured. In some cases whole syste=s can be operated for test purposes.
In others, indivi-dual components are operated for functional tests so that plant operations are not disrupted. The ice con-denser contains no active components required to function during an accident condition. Sa=ples of the ice are taken periodically and tested for boron concen-tration. 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 intermediate deck doors are tested for operability during the shutdown condition.
All active components of the containment 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 functioning. Further testing of the active components is conducted after component maintenance. Air test lines, located upstream of 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 operability while the reactor is shut-down for refueling.
GDC 41 As indicated by FSAR Section 3.1.2.4, the, shield building, surrounding the primary containment, serves as a secondary containment. The emergency gas treatment system (Section 6.2) maintains this secondary contain-ment at a negative pressure during the entire post-accident period. The emergency gas treatment system also collects and processes the secondary contain=ent atmosphere. After processing, the portion of this processed air necessary to assure a negative pressure is exhausted through the plant vent. The remainder is recirculated and distributed in the secondary contain-ment. The auxiliary building serves to collect any equipment leakage during the recirculation of contain-ment sump water. The auxiliary building ventilation system (Section 9.4.2) is isolated by an accident signal. The auxiliary building gas treatment system (Section 9.4.2) then maintains the building at a negative pressure and processes any in leakage prior to release to the environment. Postaccident hydrogen control within the containment is provided by an electrical hydrogen recombiner (Section 6.2).
Dis-tribution 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 compartment to prevent stagnation and excessive accumulation of hydrogen.
. 1
Regulation
' (10CFR)
Compliance GDC 42 FSAR Section 3.1.2.4 indicates that the emergency gas treatment system, the auxiliary building gas treatment system,.and the hydrogen recombiners are designed to permit appropriate periodic inspection of the important components. Additional discussion is provided in FSAR Sections 9.4 and 6.2.
GDC 43 FSAR Section 3.1.2.4 indicates that the emergency gas treatment system, the auxiliary building gas treatment system, and the hydrogen recombiner system are designed to permit periodic pressure testing id functional testing of their c tponents'. Furthi details are pro-vided in Sections 6.2 and 9.4.
CDC 44 FSAR Section 3.1.2.4 describes how a Seismic Category I Component Cooling System (CCS) (Section 9.2) is pro-vided to transfer heat from the Reactor Coolant System, reactor support equipment and engineered safety equip-ment to a Seismic Category I Essential Raw Cooling Water System (ERCW) (Section 9.2).
The CCS serves as an intermediate system and thus a barrier between potentially or normally radioactive fluids and the river water which flows in the ERCW System. The CCS con-sists 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 ERCW System is also used as heat sink for the containment and engineered safety equipment through use of compartment and space coolers. The ERCW System consists of two independent loops, each of which is capable of pro-
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viding all necessary heat sink requirements. The ERCW System transfers heat to the ultimate heat sink (Section 9.2).
Electric power is discussed in Chapter 8.
GDC 45 As indicated in FSAR Section 3.1.2.4, the integrity and capability of the component cooling water system l
(Section 9.2) and essential raw water system (Section l
9.2) are monitored during normal operation by alterna-l ting operation of the systems between the redundant system components. Nonsafety related systems may be l
isolated temporarily for inspection. All major com-ponents will be visually inspected on a periodic basis.
The component cooling, essential raw cooling, and auxiliary essential raw cooling pumps are arranged such that any pump may be isolated for inspection and maintenance while maintaining full plant operational
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capabilities.
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Regulation
- (10CFR)
Compliance GDC 46 As described in FSAR Section 3.1.2.4, the cooling water systems are pressurized during plant operations; thus, the structural and leaktight integrity of each system and the operability and performance of their active components are continuously demonstrated. In addition, normally idle portions of the piping system and idle components are tested during plant shutdown. The emergency functions of the systems are periodically tested out to the final actuated device.
For details, see the writeups on Electric Power (Chapter 8), Component Cooling Water (Section 9.2),
Essential Raw Cooling Water (Section 9.2), and Instrumentation and Controls (Chapter 7).
GDC 50 FSAR Section 3.1.2.5 indicates that the containment structure, including access openings and penetrations, is designed with sufficient conservatism to accommodate, without exceeding the design leakage rate, the transient peak pressure and temperature associated with a postu-i lated reactor coolant piping break up to and including a double-ended rupture of the largest reactor coolant pipe. Containment design basis is discussed further in Sections 3.8 and 6.2.
GDC 51 As discussed in FSAR Section 3.1.2.5, the containment vessel and its penetration sleeves meet the material, design, and technical process requirements of ASME Boiler and Pressure Vessel Code,Section III, Class B.
Charpy V-notch impact tests were made of the contain-ment vessel material (ASTM A 516, Grade 60) 5/8 inch' and greater, weld deposit, and the base metal weld heat effected zone employing a test temperature at least 30 F below minimum service temperature in accordance with ASME Code, Paragraph N-330. This test measures the ductile to brittle transition with I
allowable values for energy absorption given in Tables N-421 and N-422.
It insures that the material used will not behave in a brittle manner and that rapidly propagating fracture is minimized. The con-tainment boundary design considered uncertainties in material properties, residual, steady-state and transient stresses, and material flaws along with l
conservation allowable stress levels for all stressed elements of the containment boundary. All material was examined for flaws that would adversely affect the performance of the material in its intended purpose.
See Section 6.2, Containment Functional Design, for l
further details.
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'Refulation (10CFR)
Compliance CDC 52 As indicated in FSAR Section 3.1.2.5, the contain=ent design permits overpressure strength testing during construction and permits preoperational integrated leakage rate testing at contain=ent design pressure and at reduced pressure, in accordance to Appendix J.
10CFR50. The contain=ent and other equipment which may be subjected to contain=ent test conditions are designed so that periodic integrated leakage rate testing can be conducted at contain=ent design pressure.
However, reduced pressure tests are planned for the surveillance program, in accordance to the Type A test guidelines in Appendix J of 10CTR50. The pre-operational integrated leak tests at peak pressure and at reduced pressure verify that the contain=ent, including the isolation valves and the resilient penetration seals, leaks less than the allowable value of 0.25 weight percent per day at peak pressure.
Details concerning the conduct'of periodic integrated leakage rate tests are in Section 6.2.1.4.
CDC 53 FSAR Section 3.1.2.5 indicates that the containment and the containment isolation system (Section 6.2) are designed so that: (1) integrated leak tests can be run during plant lifetime (see compliance to Criterion 52), (2) visual inspections can be made of all important areas, such as penetrations, (3) an appropriate surveillance program can be maintained (Section 6.2), (4) periodic testing at containment design pressure of the leaktightness of isolation valves and penetrations which have resilient seals and expansion bellows is possible, and (5) the opera-bility of the containment isolation system can be demonstrated periodically. In testing locally the resilient seals and expansion bellows leakages, the guidelines'for Type B tests in Appendix J of 10CFR50 will be followed.
CDC 54 As described in FSAR Section 3.1.2.5, the contain=ent isolation features are classified as Seismic Category:I.
These components require quality assurance measures which enhance reliability. The containment isolation design provides for a double barrier at the contain=ent penetration in those fluid systems that are not required to function following a design basis event. All piping systems penetrating the containment, in so far as practical, have been provided with tests vents and test connections or have other provisions to allow periodic leak testing as required. Section 6.2.4.4 has further details on testing. See Section 6.2.4 for general contain=ent isolation details.
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Regulation (10CFR)
Compliance CDC 55 As indicated in FSAR Section 3.1.2.5, the reactor coolant pressure boundary is defined as those piping systems and components which contain reactor coolant at design pressure and 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 isolation signal. All other piping and components which may contain reactor coolant are low pressure, low temperature systems which would yield minimal environmental doses in the event of failure. The sampling system and low-pressure systems are described in Section 9.3. An analysis of malfunctions in these systems is included in Chapter 15.
GDC 56 As indicated in FSAR Section 3.1.2.5, at least two barriers are provided between the atmosphere outside the containment and the containment atmosphere, the reactor coolant system, or 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 3.1.2.5 indicates that those lines that penetrate the containment, do not communicate with -
either the reactor coolant pressure boundary or the containment atmosphere, and are not affected by loss-of-coolant accident forces are defined as closed systems.
All lines penetrating the containment are designed to meet CDC Criterion 57, with the exception of the feedwater system. Each of the feedwater penetrations utilize a check valve and a remotely controlled auto-matic valve external to the containment as the means for feedwater isolation. This system is further dis-cussed in Section 6.2.4.3.
CDC 60 As described in FSAR Section 3.1.2.5, liquid, gaseous, and solid radioactive waste processing equipment is provided. The principles of filtration, demineralization, evaporation, solidification and storage for decay are utilized as described in Sections 11.2, 11.3, and 11.5.
Process monitoring is provided to control this equipment and regulate releases to the environment as described in Section 11.4 CDC 61 FSAR Section 3.1.2.6 indicates that systems which may contain radioactivity are designed to ensure adequate safety under normal and postulated accident conditions.
Neguistion (10CFR)
Compliance Components are designed and located such that appropriate i
periodic inspection and testing may.be performed. All areas of the plant are designed with suitable shielding for radiation protection based on anticipated radiation dose rates and occupancy as discussed in Section 12.1.
Individual components which 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 i
is designed such that no postulated accident could cause excessive loss of coolant inventory.
Radioactive waste treatment systems are located in the auxiliary building, which contains or confines leakage under 1
normal and accident conditions.
The auxiliary building gas treatment system includes charcoal 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 radio-active waste management in Chapter 11.-
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Regulation LD OFR Compliance GDC 62 As noted in Section 31.2.6, the restraints and interlocks provided for safe handling and storage of new cr spent fuel are discussed in Section 91. The center-to-center distance between the adjacent spent fuel assemblies is sufficient to ensure suberiticality, even if unborated water is used to fill the spent fuel storage pool. The design of the spent fuel storage rack assembly is such that it is impossible to insert the spent fuel assemblies in other than prescribed locations, thereby preventing any possibility of accidental criticality. Layout of the fuel handling area is such that the spent fuel casks will never be required to traverse the spent fuel storage pool during removal of the spent fuel assemblies.
GDC 63 FSAR Section 3.1.2.6 and Chapters 9,11, and 12 descrite the monitoring capability in the fuel storage and waste handling areas and indicates that the operator will take appropriate actions if an alarm from any of these monitors is received.
GDC 64 FSAR Section 31.2.6 indicates the facility contains means for monitoring the containment atmosphere and all other important areas during both normal and accident conditions to detect and measure radioactivity which could be released under any conditions. The monitoring system includes area gamma monitors, atmospheric monitors and liquid monitors with full indication in the centrol room. Alarms are provided to warn of high activity.
Sections 11.4 and 12.h discuss 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 detall 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 qualification. TVA's financial qualifications were fully litigated before the Atomic Safety and Licensing Board, and the Board expressly found that TVA had satisfied the burden of proving that it has reasonable 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 requirements of Part 51 have been satisfied..
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Appendix E Thin AppIndix cpecifica requirements for emergency pitnn.
An emergency plan was prepared for SQNP before the granting of an operating license for unit 1.
This a
Emergency Plan was litigated before the Atomic Safety and Licensing Board, and the Board found that the emergency plan provided reasonable assurance that 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 and has.been accepted by the NRC Staff, has been implemented.
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 TVA meets the requirements of this appendi'., except that the requirement stated in Paragraph IV.b of Appendix G has not been met for the unit 1 reactor vessel. The specific areas of noncompliance, the evaluation and recommendation for an exemption to the requirements of Paragraph IV.b of Appendix G for the Sequoyah Nuclear Plant unit 1 are described in NUREG-0011, Supplement 1.
Appendix H Technical Specification 4.4.9.1.2 and operating procedures have been established to implement these requirements. Further information is provided in FSAR Section 5.4 3 7 Appendix I This Appendix provides numerical guides for design objectives and limiting conditions for operation to meet the criteria "as low as is reasonably achievable" for radioactive material in light-water-cooled nuclear power reactor effluents. TVA filed with the Commission on June 4, July 14, and September 10, 1976, the necessary information to permit an evaluation of the SQNP with respect to the requirements of Sections II. A, II.B, and II.C of Appendix I.
In this submittal TVA provided the necessary information to show conformance with the Commission' September 4, 1975 amendment to Appendix I rather than perform a detailed cost-benefit analysis required by Section II.D of Appendix I.
Appendix J TVA meets the requirements of this Appendix, except for airlock testing (Section III.D.2).
TVA has described in the FSAR and associated Technical Specifications 4
R:gulation (10CFR)
Compliance its proposed leak testing procedure for the containment airlocks, and proposes an exemption from the associated requirement of Appendix J to 10 CFR Part 50.
The proposed leak testing procedures adn the proposed exemption to Appendix J are described in NUREG-0011, Supplement 1.
It should be noted that a specific exemption to this Appendix has been granted in Technical Specification 3/4.6.1 3 regarding testing of each containment airlock.
Appendix K This Appendix specifies fet. ares of acceptable ECCS evaluation models. As noted above for 50.46, the analysis for SONP 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 Gen-ra! for anti-trust review of license applicat '.uns.
As noted above, the anti-trust review for S0' / 1 and 2 took place at the construction permii staaa.
Appendix M This Appendix covers standardization of design and is not applicable to SONP.
Appendix N This Appendix covers standardization of nuclear power plant designs and is not applicable to SQNP.
Appendix 0 This Appendix covers standardization of design and is not applicable to SONP.
I Appendix P This Appendix is proposed, 39 Fed. Reg. 26293, and it applies to fuel reprocessing plants.
Accordingly, it is not applicable to SQNP.
Appendix 0 This Appendix governs preapplication early review of site suitability issues and is not applicable to SQNP.
Appendix 0 This Appendix is proposed, 39 Fed. Reg. 26297, and (Proposec) it would apply to fuel reprocessing plants, not power reactors..
d e
Regulation (10CFR) __
Corplinnee 100.1 This regulation is explenatory and does not impose independent obligations on licensees.
100.2 This regulation is explanatory. SONP is not novel in design and is not unproven as a prototype or pilot plant.
100 3 This regulation is explanatory and does not iepose independent obligations on licensee.s.
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 of the FSAR.
The exclusion area, low population zone, and population center distance are provided and described. The FSAR also describes the characteristics of reactor design and operation.
100.11 An 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 three (3) miles from the centerline of Unit No.1 contain=ent. As indicated in Section 2.1 3 5, the nearest population center, as defined by 10 CFR 100 3(c), based on the 1970 census, is Chattanooga, Tennessee, which is 9 5 miles southwest of the site.
The FSAR accident analyses, particularly those in Chapters 6 and 15, demonstrate that offsite doses 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 SQNP site with this Appendix is discussed in the Sequoyah Nuclear Plant Safety Evaluation Report, NUREG-0011.