ML20077K442
| ML20077K442 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/10/1983 |
| From: | Tucker H DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8301130038 | |
| Download: ML20077K442 (33) | |
Text
l .
- i, t
' DUKE Poven Goyne,c P.O. !M)x 33180 CIIARLOTTE, N.C. 28242 IIAL II. TUGKEH TEt.erisown v s. e en e.,,,,,, ganuay 10,_ 3%} ; g,3,3 37xw, NtfLEAR ,30ptt ,gog & ,,_
Mr. Harold R. Denton, Director l Of.fice of Nuclear Reactor Regulation l U. S. Nuclear Regulatory Commissic.n l Washington, D. C. 20555 i
Attention: Ms. E. G. Adensam,, Chief - l
- Licensing Branch No. '4 Re: Catawba Nuclear Station ;
Docket Nos. 50-413 and 50-414
Dear Mr. Denton:
My letter of November 30, 1982 transmitted a response to Question 100.4 which requested an evaluation to demonstrate that the Catawba Nuclear Station, Units 1 and 2, met the pertinent regulatory requirements of Title 10 of the Code of Federal Regulations. A number of typographical errors in that response have been corrected and the attached evaluation replaces the previous submittal in its entirety.
Very truly yours, f b -4 4 lg Hal B. Tucker ROS/php Attachment 1
cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. P. K. Van Doorn ,
NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild, Esq.
Attorney-at-Law .
P. O. Box 12097 Charleston, South Carolina 29412 Palmetto Alliance 21355 Devine Street Columbia, South Carolina 29205 00l 8301130038 830110 PDR ADOCK 05000413 A PDR C J
Mr. Harold R. Denton, Director January 10, 1983 Page 2 cc: Mr. Jesse L. Riley Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207 Mr. Henry A. Presler, Chairman Charlotte-Mecklenburg Environmental Coalition 943 Henley Place Charlotte, North Carolina 28207 M
N
COMPLIANCE OF CATAWBA NUCLEAR STATION WITH THE NRC REGULATIONS OF 10 CFR PARTS 20, 50, AND 100 Regulation (10 CFR) Compliance l 20.1(a) This regulation merely states the general purpose for 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 regulations 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 pre-l scribed therein. It does not impose any independent obli-gations on licensees.
20.1(c) Conformance to the ALARA principle stated in this regulation is ensured by the implementation of Duke policies and appro-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 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 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.
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. Con-formance is documented by the use of appropriate personnel monitoring devices and the maintenance of all required records.
Regulation (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 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 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. When 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 Catawba respiratory protection program implements the requirements of this regulation by ensuring the proper use of approved respiratory protection equipment. The Catawba 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) Duke Power Company uses only tested and certified respiratory equipment. Therefore, application for this authorization is not required.
20.103(f) The respiratory protection program is in full conformance with the requirements of 20.103(c).
20.103(g) Duke Power Company will make this notification within the prescribed time. Respiratory equipment is not now in use at the Catawba Nuclear Station.
20.104 Conformance with this regulation is assured by appropriate Duke Power Company policies regarding employment of indivi-duals under the age of 18 and the Duke Power Company System Health Physics Manual restricting these individuals' access to restricted areas.
~ .__ -
Regulation (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 i' are complied with through the implementation of procedures, Technical Specifications, and administrative policies which control the use and transfer of radioactive materials. Ap-propriate surveys and monitoring devices document this com-pliance.
20.106(a) Conformance with the limits specified in this regulation is assured through the implementation of procedures and applicable Technical Spedifications which provide adequate sampling and anslyses, 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.
j 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 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(e) This regulation provides criteria by which the Commission may impose further limitations on releases of radioactive i 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 7 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 Catawba Nuclear Station to determine exposure of individuals to concentrations of radio-active materials. Appropriate health physics procedures and administrative policies implement this requirement.
Regulation (10 CFR) Compliance 20.201 The surveys required by this regulation are performed at adequate frequencies and contain such detail as to be consis-tent with the radiation hazard being evaluated. When neces-sary, 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 Duke Power Company System Health Physics Manual and applicable station health physics proce-dures require these surveys and provide for their documenta-tion in such manner as to ensure compliance with the regula-tions 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 en-sure that all individuals are supplied with, and required to 4 use, appropriate personnel monitoring equipment. The Radia-l tion 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 l 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-l nating radiation hazards or radioactive materials, and using the radiation symbol, conform to the conventional design pre-scribed in this regulation.
- 20.203(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" are conformed to by the implementation of the Technical Speci-fications and appropriate station health physics procedures, as well as the System Health Physics Manual. The controls and other protective measures set forth in the regulation are maintained under the surveillance of the station Health Physics group.
l
! 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.
4 .
i i Regulation (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 3 Manual.
20.203(f) The container labeling requirements set forth in this regula-tion are complied with through the implementation of appropriate 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 Catawba Nuc-lear Station. Adequate controls are provided within the sta-tion health physics procedures to assure safe and proper ap-plication of these exceptions.
20.205 All of the requirements of this regulation pertaining to pro-j cedures for picking up, receiving, and opening packages of
- radioactive materials are implemented by the System Health l Physics Manual and appropriate station health physics pro-i cedures. These procedures also provide for the necessary documentation to ensure an auditable record of compliance.
l 20.206 The requirements of 10 CFR 19.12 referred to by this regula-tion are satisfied by the orientation training conducted at Catawba Nuclear Station. Appropriate departmental procedures set fo=th requirements for all employees who frequent or work at Catawba Nuclear Station to receive this instruction on a periodic basis.
20.207 The storage and contro 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 i regulation are complied with through station operating pro-J cedures, Technical Specifications, and the provisions of the station license.
1 20.302 No such application for proposed disposal procedures, as de-scribed in this regulation, has been made or is contemplated 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.
- - -,, --,,my.. _
Regulation (10 CFR) Compliance 20.305 Specific authorization, as described in this regulation, is not currently being sought by Duke Power Company for treatment or disposal of wastes by incineration.
20.306 Duke Power rad waste procedures and Catawba Nuclear Station Technical Specifications will assure compliance with this regulation.
20.401 All of the requirements of this regulation are complied with through the implementation of appropriate Technical Specifica-tions and station procedures pertaining to records of surveys, radiation monitoring and waste disposal. The retention periods specified for such records are also provided for in these specifications and in station and departmental procedures.
20.402 Catawba Nuclear Station has established an appropriate inven-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.
20.403 Notifications of incidents, as described in this regulation, are assured by the requirements of the Technical Specifica-tions, the System Health Physics Manual and appropriate sta-tion procedures, which also provide for the necessary assess-l ments to determine the occurrence of such incidents.
20.405 Reports of 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.
20.407 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 i upon termination of an individual's employment or work assign-ment is generated through the provisions of Duke Power Company procedures.
L 20.409 The notification and reporting requirements of this regulation, and those referred to by it, are satisfied by the provisions l
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 obligations on licensees.
i l
l
Regulation (10 CFR) Compliance 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 impose 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 viola-tions of its rules. It does not impose any independent obliga-tions on licensees.
l f
l i
l k
l 1
, Regulation (10 CFR) Compliance i
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 in-
- dependent obligations on licensees.
50.3 This regulation governs the interpretation of the regulations by the NRC and does not impose independent obligations on licensees.
1 50.4 This regulation gives the address of the NRC and does not im-l pose independent obligations on licensees.
50.7 This regulation describes the rights of employees of a licensee.
Notice of these rights is posted at the Catawba Nuclear Station and appropriate General Office areas.
50.8 This regulation states that the NRC has approval from the OMB t for the collection of information specified in this regulation.
{ This regulation does not impose any independent obligations on the licensee.
l 50.10 These 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 activities at Catawba 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 l on licensees.
50.2'1 These regulations merely describe the types of licenses that 50.21 the NRC issues. They do not address the substantive require-50.22 ments that an applicant must satisfy to qualify for such 50.23 licenses.
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 require-ments in effect at the time when filing its license applica-tion and the amendments to it. In particular, 10 CFR 50.30(f) requires that a license application must be accompanied by l . _- -
Regulation (10 CFR) Compliance
, any Environmental Report required pursuant to 10 CFR Part 51, and Duke Power Company has submitted an Environmental Report covering Catawba Nuclear Station.
50.31 These regulations merely permit more efficient organization i
50.32 of the license application and impose no independent obliga-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 quali-fications, and a list of regulatory agencies with jurisdica-tion over the applicant's rates and services. This infor-mation was provided in the Catawba Nuclear Station operating license application.
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, as amended, was performed at the construction permit stage.
50.34(a) This regulation governs the contents of the Preliminary Safety Analysis Report and is relevant to the construction permit stage rather than the operating license stage.
50.34(b) A Final Safety Analysis Report (FSAR) has been prepared and submitted, which addresses in the chapters indicated the i 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 (4) design and performance evaluation - ECCS performance is discussed and shown to reet 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 (ii) managerial and administrative controls -
Chapters 13 and 17. Chapter 17 discusses compliance with the quality assurance require-ments of Appendix B
Regulation (10 CFR) Compliance (iii) plans for preoperational testing and initial operations - Chapter 14 (iv) plans for conduct of normal operations -
Chapters 13 and 17. Surveillance and periodic testing is specified in the Technical Specifica-tions (v) plans for coping with emergencies - Emergency Plan (Chapter 13)
(vi) Technical Specifications prepared in conjunc-tion with the Staff (Chapter 16)
(vii) as discussed in various FSAR sections, the major safety-related structures, systems, and components are not shared between units. Section 1.2.2.13 identifies those structures, systems, and com-ponents that are shared. Station operating pro-cedures provide for separation were necessary.
Control of access is discussed in the Security Plan (7) technical qualifications - Chapter 13 (8) operator requalification program - Chapter 13 50.34(c) A physical security plan was prepared and submitted as re-t quired by this regulation for Catawba Nuclear Station.
50.34(d) A safeguards contingency plan has been prepared and submitted as required by this regulation for Catawba Nuclear Station.
50.34(e) Duke Power administrative and document control procedures pro-tect plans and other related Safeguards Information against unauthorized disclosure in accordance with 10 CFR 73.21.
50.34(f) This regulation applies only to applicants for construction permits as of February 16, 1982. However, the response to TMI-related items is described in FSAR Secticn 1.9.
50.34(g) This regulation applies only to applicants for operating licenses or construction permits docketed after May 17, 1982.
Catawba Nuclear Station was docketed prior to May 17, 1982.
50.34a This regulation requires the inclusion of the design objec-tives for equipment used to control relases of radioactive material in effluents of nuclear power reactors into the ap-plication for construction permit and operating license.
i Chapter 11 of the Catawba FSAR provides the required infor-l mation concerning the radioactive waste management system at the Catawba Nuclear Station.
t .
Regulation (10 CFR) Compliance 50.35 This regulation is relevant to the construction permit stage l rather than the operating licenst 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.
50.36a The Catawba Nuclear Station Technical Specifications include 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 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 Catawba Nuclear Station.
50.38 This regulation prohibits the NRC from issuing a license to foreign-controlled entities. Duke Power Company's statement that is is not owned, controlled, or dominated by an alien, l foreign corporation, or foreign government is contained in the operating license application for Catawba Nuclear Station.
50.39 This regulation provides that applications and related docu-ments may be made available for public inspection. This im-poses no direct obligations on applicants and licensees.
50.40 This regulation provides considerations to " guide" the Com-mission in granting licenses as follows:
50.40(a) The design and operation of the facility is to provide reason-
- able assurance that the applicant will comply with NRC regula-l tions, including those in 10 CFR Part 20, and that the health and safety of the public will not be endangered. The basis for Duke Power Company's assurance 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, the lengthy process by which the plant is designed, constructed, and reviewed, including re-views by Duke Power Company'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 af-fected.
50.40(b) This regulation provides that no consideration of financial qualifications is necessary for an electric utility applicant l
(
i
Regulation (10 CFR) Compliance at the operating license stage. The technical qualifications of Duke Power Company are disc 1ssed in FSAR Section 13.1.
50.40(c) Another consideration is that le 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
' l enclosure as well as the contents of the entire FSAR and re-lated correspondence over the years, 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 assur-ance that the license will not be inimical in that Duke Power Company has a viable security plan for Catawba Nuclear Station that Duke Power Company is not controlled by agents of foreign l countries, and that Duke Power 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 submitted an Environmental Report as part of the operating license application.
50.41 This regulation applies to class 104 licensees, such as those for devices used in medical therapy. Catawba Nuclear Station 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 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 chang-l ing, Duke Power Company periodically makes load projections, and in Duke Power Company's judgment the need for Catawba Nuc-lear 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 pro-duced 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.
~
Regulation (10 CFR) Compliance 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 Catawba Nuclear Station 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 atmo-sphere, as specified in 30 CFR 50.44(e). The system consists of internal recombiners, a hydrogen analyzer, and a hydrogen skimmer system. The containment recirculation system and hydrogen purge system complement the recombiner system.
Catawba Nuclear Station meets the requirements of NUREG-0660 l and NUREG-0694. The requirements of 10 CFR 50.44 are satisfied.
50.45 This regulation provides standards for construction permits rather than operating licenses and is therfore not material to this operating license proceeding.
, 50.46 FSAR Sections 6.3 and 15.6.5 describe the Emergency Core Cooling System and the methods used to analyze ECCS perfor-mance following a postulated loss of coolant accident.
In FSAR Section 15.6.5, Duke Power Company provided the re-I sults of a LOCA-ECCS analysis for Catawba Nuclear Station using an NRC approved evaluation model, which is in compliance with Appendix K to 10 CFR 50. The analysis, based on an over- _
all peaking factor (Fq) of 2.32, provided results in compliance i with the criteria of 10 CFR 50.46(b). The Fq limit will be reflected in the Technical Specifications.
50.47 The Radiological Emergency Plan was submitted to the NRC as a separate document at the same time the FSAR was filed.
50.48 The Fire Protection System is described in Section 9.5 of the FSAR. An evaluation of the fire protection plan which demon-strates compliance with 10 CFR 50 Appendix R was included as Appendix D to the Catawba Nuclear Station Response to Appendix A to BTP-APCSB 9.5-1, dated August 1981.
I 50.50 This regulation provides that the NRC will issue a license l 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.
=. _ _ _
Regulation (10 CFR) Compliance 50.52 This regulation provides for the combining in a single license of a number of activities. It imposes no independent obliga-tion 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 Catawba Nuclear Sta-tion will be within the United States and subject to the jurisdiction of the United States, as is evident from the des-cription of the facility in the operating license application.
50.54 This regulation specifies certain conditions that are incor-porated in every license issued. Compliance is effected 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, l 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 provides information concerning the Quality Assurance Pro- ,
gram that was utilized. As a further example of a specific system, FSAR 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 paragraph leading into paragraphs (c) through (i) of the regulation.
l 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) Reactor coolant system components were designed and fabricated in accordance with these regulations except for the Unit 1 1 reactor vessel which was designed and fabricated to ASME Sec-i tion III, 1971 Edition through Winter 1c71 Addenda as dis-cussed in FSAR Section 5.2.1.1. Additio'al information on reactor coolant system components can be found in Chapters 3 and 5 of the FSAR.
50.55a(g) Inservice Inspection (ISI) requirements are delineated in this part and are specified in the Technical Specifications.
The Catawba inservice inspection program is delineated in Section 6.6 of the FSAR.
Regulation (10 CFR) Compliance 50.55a(h) As discussed in Chapter 7, the protection systems meet IEEE 279-1971.
50.55a(i) Fracture toughness requirements ere set forth in Appendices G and H of 10 CFR 50. Technical Specifications require the use of reactor vessel material irradiation surveillance speci-mens and updating of the "heatup" and "cooldown" curves given in the Technical Specifications. Further information is given in FSAR Section 5.3 concerning the irradiation surveillance program.
50.55a(j) This regulation applies only to the construction permit stage.
50.55e This regulation is only proposed, 39 Federal Regulation 26297, j and applies to fuel reprocessing plants.
50.56 This regulation provides that the Commission will, in the absence of good cause shown to the contrary, issue an operat-ing license upon completion of the construction of a facility in compliance with the terms and conditions of the construc-tion permit. This imposes no independent obligations on the applicant.
50.57(a) This regulation requires the Commission to make certain find-ings before the issuance of an operating license. These findings for Catawba Nuclear Station can be made for the reasons given in this enclosure generally. Specifically:
(1) Construction of the facility is being completed in con-formity 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 re-gulations, has been demonstrated by the application.
i l (2) The Technical Specifications and resulting operating pro-l cedures provide assurance that the facility will operate in conformity with the application as amended and with the rules and regulations, with the noted exceptions to 10 CFR 50.
(3) The application demonstrates that the facility can be operated without endangering the health and safety of the i public and in compliance with the regulations, as noted above.
, (4) The application demonstrates that Duke Power Company is l technically and financially qualified to operate the unit.
(
(5) The applicable provisions of 10 CFR 140 have been satisfied.
- - . .- . .-. - .-- ._~
Regulation (10 CFR) Compliance (6) The Catawba Security Plan assures that special nuclear material will be appropriately safeguarded. The applica-tion demonstrates that the operation of the facility 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-i pleted on a schedule acceptable to the Commission.
50.57(c) This regulation provides for a low power testing license.
Such a license has not been requested for Catawba Nuclear Station.
50.58 This regulation provides for the review and report of the Advisory Committee on Reactor Safeguards. The ACRS will review the operating license application for Catawba Nuclear Station in accordance with its usual practice.
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.
l 50.70 The Commission has assigned a resident inspector to Catawba Nuclear Station. Duke Power Company has provided office space in accordance with the requirements of this section. Duke Power Company oermits 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 re-quirements of sections (a) through (e) of this regulation and the license. Section (e) requires that the FSAR be updated i within 24 months after date of issuance of operating license j 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.
I l 50.78 This regulation concerns the transmittal of installation in-formation and verification to implement the US/IAEA Safeguards Agreement upon request of the NRC. The.NRC has not requested this information.
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 Catawba Nuclear Station proceeding.
50.81 This regulation permits the creation of mortgages, pledges, and liens on licensed facilities, subject to certain provisions.
These provisions concern the requirements and restrictions on creditors and do not impose independent obligations on licensees.
s-_m-w%.a4 =---+. e- a- -A - =r.- - - - eu. u. m u a. -- d.24=-a ma-- a A h-A--. w a u-.,.__+,- - - .Au--s e aan-aa _.mm.1 u~- 4 .6.m. 4 _.u_.., a 4
1 4 e
?
5 I
I 4
l.
(
l i
4
.1 1
5
.1 4
I i
g -e ng*
.l 1 2 A
i -
t e
I I
r 0
l t
i.
I b
,- i i
a l
l n
t h
6 I
I I
i m -
)
Regulation (10 CFR) Compliance i
50.82 This regulation provides for the termination of licenses. It does not apply to Catawba 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 accord-ance 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 modi-50.101 fications of licenses by the Commission under unusual circum-50.102 stances. No such circumstances are present in the Catawba
! 50.103 Nuclear Station proceeding, and these regulations are not ap-l plicable.
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.
50.110 This regulation governs enforcement of the Atomic Energy Act, the Energy Reorganization Act of 1974, and the NRC's regula-tions and orders. No enforcement action is at issue in the Catawba Nuclear Station proceeding, and so this regulation is not applicable.
l I
- , - - - - - . - - - - - - , - - - - - - . , , - - , , , ---n--. -- - -- - -- - - --- - - -, - - - ~ - - - ----
Regulation (10 CFR) Compliance Appendix A GDC 1 Section 3.1 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 pro <isions 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, and the importance of the safety functions to be performed.
GDC 3 FSAR Section 3.1 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.
GDC 4 FSAR Section 3.1 describes the design features used to accom-modate the effects of and compatibility with the environmental conditions associated with all modes of operation and postu-lated accidents. Chapter 3 provides information concerning the specific design features for protection against missles, jet impingement and pipe rupture. Provisions for qualifica-tion of equipment for all postulated environments is described in FSAR Section 3.11.
GDC 5 As described in FSAR Section 3.1, those structures, systems, and components which are shared by both units are tabulated in FSAR Section 1.2.2.13. It is concluded that safety func-l tions are not significantly impaired by such sharing.
GDC 10 FSAR Section 3.1 indicates that the reactor core and asso-ciated systems are designed to function throughout the de-sign lifetime without exceeding fuel damage limits, using protection criteria specified in Section 3.1 and Chapters 4, 7, and 15.
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 FSAR Chapter 4.
FSAR Section 3.1 describes the inherent and design features l GDC 12 which eliminate or limit the various types of oscillations.
Core stability is further described in Section 4.3.
Regulation (10 CFR) Compliance 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 FSAR 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 I and associated auxiliary, control and protection systems are designed to ensure the integrity of the reactor coolant pres-sure boundary with adequate margins during normal operations and anticipated transients. The design codes used for the Re-actor Coolant System are described in Chapter 5. Details con-cerning 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 l
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 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 power sources. Periodic tests of the diesel generators, the transfer system and the station batteries are made, as required by Technical Specifica-tions.
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 accident conditions.
Regulation (10 CFR) Compliance 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 room is restricted, an auxi-liary 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 pro-tection 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 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 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.
GDC 24 FSAR Section 3.1 discusses separation of the protection and control systems, such that the failures of any signal control l system component or channel or the failure or removal from l service of any protection system component or channel which is common to the protection ana 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 been designed to assure that specified acceptable fuel de-sign limits are not exceeded in the event of any single reactivity control malfunctions, including an accidental withdrawal of control cluster groups. Further details are provided FSAR Sections 4.3.1.4 and 7.7.2.2.
GDC 26 As indicated in FSAR Section 3.1, two independent reactivity I control systems of different design principles are provided.
l l l 1
b Regulation (10 CFR) Compliance One of the systems uses control rods; the second system em-ploys dissolved boron as a chemical shim. Reactivity con-trol 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, means are provided for shutdown reactivity for cooling the core under any antici-
, pated condition and with appropriate margin for contingencies, i Combined use of rod cluster control and chemical shim control I permit the necessary shutdown margin to be maintained during the long term xenon decay and plant cooldown. These means l are discussed in detail in FSAR Sections 4.3 and 7.2 GDC 28 FSAR Section 3.1 indicates that core reactivity is controlled l by a chemical poison dissolved in the coolant, rod cluster i
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 re-activity 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 protection system is further discussed in Section 7.2. The '
reactivity control systems are discussed in Sections 4.6 and 7.2.
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 accorded the quality measures 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, close control is maintain-
. ed over material selection and fabrication for the reactor l coolant system to ensure that the boundary behaves in a non-t brittle 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 Technical Specifications ensure operation within the pressure-temperature limit re-lative to this criterion.
Regulation (10 CFR) Compliance l GDC 32 FSAR Section 3.1 describes how the design of the reactor 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 l of the vessel. The reactor arrangement within the contain-ment provides sufficient space for inspection of the external surfaces of the reactor coolant piping. - cept for the area of pipe within the primary shielding .oncrete. Additional details can be found in Section 5.2.
l GDC 33 As indicated in FSAR Section 3.1, the chemical and volume con-trol system provides a means of reactor coolant makeup and ad-justment 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 de tails 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 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 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 redundancy above 350 F is provided by the steam generators, auxiliary feed pumps, and attendant piping. Details of the residual heat removal system design are in FSAR Section 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 preserved. 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 components of the emergency core cooling system. An inspection is performed 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 Regulation (10 CFR) Compliance visually or by boroscopic examination for erosion, corrosion, and vibration wear evidence. Nondestructive inspection is performed where such techniques are desirable and appropriate.
Technical Specifications require inservice inspection in ac-cordance with applicable ASME Codes. Details of the inspec-tion programs are provided in Chapters 5 and 6.
CDC 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 i 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 cen-trifugal charging pumps are part of the charging system, and
, this system is in continuous operation during plant operations.
Actuation circuits are tested and remote-operated valves are exercised periodically. The testing is described in detail 1 in FSAR Sections 6.3.4, 7.3, and per Technical Specification surveillance requirements.
GDC 38 As indicated in FSAR Section 3.1, the containment spray sys-i tem, ice condenser, and RHR spray system are provic'ed to re-move heat from the containment following a loss-of-coolant l accident. An air return system is used to circulate air and steam through the containment after the initial blowdown.
This maintains proper mixing of the containment air and steam with the heat removal media for the necessary heat removal.
The loss of a single active component was assumed in the de-sign of these systems. Emergency power system arrangements ensure the proper functioning of these systems. Two electrical buses, each connected to both onsite and offsite power, feed
- the pump motors and the necessary valves. Further details are provided in Sections 6.2 and 8.3.
GDC 39 As indicated in Section 3.1, the ice condenser design includes i provisions for visual inspections of the ice bed flow channels, doors, and cooling equipment. The air return fan system pro-vides for visual inspection of the fans and the associated backflow dampers and for duct systems that are not embedded in concrete. The containment spray system and the residual
- heat removal system (ND) sprays are designed such that active i and passive components can be readily inspected to demonstrate system readiness. Pressure containing systems are inspected l for leaks from pump seals, valve packing, flange joints, and relief valves. During operational testing of the containment spray pumps and ND pumps, the portions of the systems subjected to pressure are inspected for leaks. System design details are given in Section 6.2.
t
. _ - _ - . . . . . - - _ _ - -.,_,-- -~ _- . _ . , - . . - - - , _ _ _
Regulation (10 CFR) Compliance 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 1
periodically and tested 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. The top deck doors 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 i 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 shutdown for re-fueling.
GDC 41 As indicated by FSAR Section 3.1, the shield building, sur-
- rounding the primary containment, serves as a secondary con-tainment. The annulus ventilation system (Section 6.2) main-tains this secondary containment at a negative pressure dur-ing the entire post-accident period. The annulus ventilation system also collects and processes the secondary containment 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 containment. The auxiliary build-ing serves to collect any equipment leakage during the recir-culation 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 9.4.3) processes any inleakage prior to release to
! the environment. Postaccident hydrogen control within the containment is provided by electrical hydrogen recombiners (Section 6.2). Distribution of the atmosphere within the containment is provided by the air return fan system (Section
- 7.4.6). The system also takes a suction in each compartment l to prevent stagnation and excessive accumulation of hydrogen.
GDC 42 FSAR Section 1.3 indicates that the annulus ventilation sys-tem and the hydrogen recombiners are designed to permit ap-propriate periodic inspection of the important components.
l l -
Regulation (10 CFR) Compliance Additional discussion is provided in FSAR Sections 9.4 and 6.2.
GDC 43 FSAR Section 3.1 indicates that the annulus ventilation sys-tem and the hydrogen recombiner system are designed to permit periodic pressure testing and functional testing of their components. 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 (KC) (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 (RN) (Section 9.2). The KC Sys-tem serves as an intermediate system and thus a barrier be-tween potentially or normally radioactive fluids and the lake /
pond water which flows in the RN System. The KC System con-sists of two independent engineered safety subsystems, each of which is capable of serving all necessary loads under nor-mal or accident conditions. In addition to serving as the heat sink for the KC System, the RN System is also used as heat sink for the containment and engineered safety equipment through use of compartment and space coolers. The RN System consists of two independent loops, each of which is capable of providing all necessary heat sink requirements. The RN 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, the integrity and capabi-lity of the component cooling water system (Secticn 9.2) and nuclear service water system (Section 9.2) are monitored during normal operation by alternating operation of the sys-i tems between the redundant system components. Nonsafety related systems may be isolated temporarily for inspection.
All major components will be visually inspected on a periodic basis. The component cooling and nuclear service water pumps i are arranged such that any pump may be isolated for inspection and maintenance while maintaining full plant operational ca-pabilities.
GDC 46 As described in FSAR Section 3.1, the cooling n ter systems i are pressurized during plant operations; thus, the structural i 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 period-dically tested out to the final actuated device.
l For details, see the write-ups on Electric Power (Chapter 8),
I Component Cooling Water (Section 9.2), Nuclear Service Water (Section 9.2), and Instrumentation and Controls (Chapter 7).
l l _ _ _ _ -_
Regulation (10 CFR) Compliance 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 design leakage rate, the transient peak pressure and temperature associated with a postulated reactor coolant l 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, the design condition for the containment pressure boundary is based on the parameters derived from the design basis accident. For this design con-dition, the steel liner material behaves in a nonbrittle manner and has the capability to minimize the propagation of any undetected flaw. Detailed information on the steel liner material is found in Section 3.8.2.
GDC 52 As indicated in FSAR 3.1, the containment design permits overpressure strength testing during construction and per-mits preoperational integrated leakage rate testing at cal-culated peak 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 conducted at calculated peak accident pressure. The preoperational 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 integrated leakage rate tests are in Section 6.2.6.
GDC 53 FSAR Section 3.1 indicates that the containment and the con-tainment 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 peak accident pres-sure of the leaktightness of isolation valves and penetrations which have resilient seals and expansion bellows is possible, and (5) the operability 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 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 N t
\ 'g Regulation 'x '
(10 CFR) Compliance 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 6.2.6 has further details on testing. See Section 6.2.4 for general containment isolation details.
GDC 55 As indicated in FSAR Section 3.1, the reactor coolant pres-sure boundary is defined as those piping systems and compo-nents 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 i '
structure. All sampling lines are provided with remotely '
operated valves for isolation in the event of a failure.~ (i These valves also close automatically on a containment iso- \ ', q lation signal. All other piping and components which may ',d contain reactor coolant are low pressure, low temperature" -
systems which would yield mininial environmental doses in the ,1 ->
event of failure. The sampling system and low pressure sys- ,'
tems are described in Section 9.3. An , analysis of malfunc- h g S '
tions in these systems is included in Chapter 15. ,
GDC 56 As indicated in FSAR Section 3.1, at least two barriers are; 3
~
provided between the atmosphere outside the containment and '
' ,s -
the containment atmosphere, the reactor co9 1a'nt system, er- %t i .\ 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 communicate with either the reactor coolant pressure boundary or the containment atmosphere, and s , ,
are not affected by loss-of-coolant accident forces are dp i fincd as closed systems. All lines penetrating the coritain-' 3 (
ment are designed to meet GDC Criterion 57. ,
GDC 60 As described in FSAR Section 3.1, provisions for liquid, 's gaseous, and solid radioactive waste processing is provided, The principles of filtration, demineralization, evaporation,' -
~
solidification and storage for decay are utilized as de- '
t s -
scribed in Sections 11.2, 11.3, and 11.4. Process monitoring .
is provided to control this equipment and regulate releases -
to the environment as described in Section 11.5. ,' .
GDC 61 FSAR Section 6.1 indicates that systems which may contain 4 radioactivity are designed to ensure adequate 'safeti under normal and postulated accident conditions. Components sre designed and located such that appropriate periodic inspec-tion and testing may be performed. All areas of the plant ,
s
~
Regulation (10 CFR) -
Con:pliance are designed with suitable shielding for radiation protec-tion 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 is design-ed such that no postulated accident could cause excessive I 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 charcoal filtratian which minimizes radioactive material release as-sociated with a postulated spent fuel handling accident.
Foal storage and handling is discussed in Section 9.1 and radioactive wa;te management in Chapter 11.
, 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 ciscussed in Section 9.1. The center-to-center distance be-1 tween the adjacent spent fuel assemblies is sufficient to
' ensure subcriticality, 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
> locat'icns, thereby preventing any possibility of accidental
'. , criticality- Layout of the fuel handling area is such that L' ,the spent fuel ca'sks will never be required to traverse the s
< spent fuel' storage pool during removal of the spent fuel
,Essemblies.
GDC 53 FSAR Section 3.1 and Chapters 9, 11, and 12 describe the monitori~ng. capability in the fuel storage and waste handling areat and indicates that the operator will take appropriate actions if an alarm from any of these monitors is received.
' ~
GDC E4/' FSAR Section 3.1 indicates the facility contains means for monitoring the containment atmosphere and all other impor-tant areas dur.ing both normal and accident conditions to de-
. tect and mea's'ure' radioactivity which could be released under
>any conditions. The monitoring system includes area gamma y
' monitors,' atmospheric monitors and liquid monitors with full
,, indication in the control room. Alarms are provided to warn effluent and area radiological monitoring systems of high Section.11.5 discusses the process and 11.3 des-activity.
-cribes the offsite monitoring program.
Appendix 8 Chap'ter 17 of the FSAR describes in detail the provisions of
-the quality assurance program which has been implemented to e meet all applicable requirements of Appendix B.
3[ , --
- r ,
Regulation (10 CFR) Compliance Appendix E This Appendix specifies requirements for emergency plans.
An emergency plan was prepared for Catawba Nuclear Station in accordance with the provisions of this Appendix.
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. AdC tional information to demon-strate compliance can be found in FSAR Section 5.2.4 concern-ing the irradiation surveillance program. Heatup and cooldown limits consistent with the requirements of this Appendix are established in the Technical Specifications.
A Reactor vessel material surveillance program requirements l ppendix H are delineated in this part. Technical Specifications and op-erating procedures have been established to implement these requirements with a further requirement to update the "heatup" and "cooldown" curves provided in the Technical Specifications.
Further information is provided in FSAR Section 5.2.4.
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. Further information is provided in FSAR Chapters 11 and 12 and Catawba Nulear Station Technical Specifications.
Appendix J Reactor containment leakage testing for water cooled power reactors is delineated in this Appendix. These requirements are given in the Technical Specificaticns. Additional infor-mation concerning compliance can be found in FSAR Section 6.2.6 and FSAR Chapter 14.
Appendix K This Appendix specifies features of acceptable ECCS evaluation models. As noted above for 50.46, the analysis for Catawba Nuclear Station has been conducted using a model which has been accepted by the Commission staff as meeting the require-ments 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 Catawba Nuclear Sta-tion took place at the construction permit stage.
Appendix M This Appendix covers standarization of design and is not applicable to Catawba Nuclear Station.
Regulation (10 CFR) Compliance Appendix N This Appendix covers standardization of nuclear power plant designs and is not applicable to Catawba Nuclear Station.
Appendix 0 This Appendix covers standardization of design and is not applicable to Catawba Nuclear Station.
Appendix P This Appendix is proposed, 39 Federal Regulation 26293, and it applies to fuel reprocessing plants. Accordingly, it is not applicable to Catawba Nuclear Station.
Appendix Q This Appendix governs preapplication early review of site suitability issues and is not applicable to Catawba Nuclear Station.
Appendix Q This Appendix is proposed, 39 Federal Regulation 26297, and (Proposed) it would apply to fuel reprocessing plants, not power reactors.
Appendix R This Appendix provides the guidelines for the Fire Protection Program. The Fire Protection System is described in Chapter 9 of the FSAR. An evaluation of the Catawba Nuclear Station Fire Protection Program which demonstrates compliance with 10CFR50 Appendix R was included as Appendix D to the Catawba Nuclear Station Response to Appendix A to BTP APCSB 9.5-1, dated August 1981.
100.1 This regulation is explanatory and does not impose indepen-dent obligations on licensees.
100.2 This regulation is explanatory. Catawba Nuclear Station is l 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.
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, l low population zone, and population center distance are pro-l vided and described. The FSAR also describes the character-l istics of reactor design and operation.
l 100.11 An exclusion area has been established, as described in FSAR Section 2.1.2. The low population zone required by 100.11 (a)(2) has been established, as described in FSAR Section 2.1.3.4, as the area within a radial distance of 20,000 feet (6096 meters) from the centerline of the reactors. As indi-cated in Section 2.1.3.5, the nearest population center, is defined by 10 CFR 100.3(c), based on the 1970 census, is Rock l Hill, South Carolina, which is 5.8 miles south-southeast of l the site.
1 Regulation (10 CFR) Compliance The FSAR accident analyses, particularly those in Chapters l 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 Catawba Nuclear Station site with this Appendix is dis-cussed in FSAR Section 2.5.