ML19331B495
| ML19331B495 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/11/1980 |
| From: | Ferguson J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Youngblood R Office of Nuclear Reactor Regulation |
| References | |
| 701, NUDOCS 8008120343 | |
| Download: ML19331B495 (37) | |
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Vxmoz2rIA Er.ncrarc orn Pownn Coximw I R.rcuxown, VxmorwxA eneet Jaca E.Famocoont m c== v== r ,
I August 11, 1980 Mr. Harold R. Denton, Director Serial No. 701 Office of Nuclear Reactor Regulation NO,LQA/JNC,ESG,JTR Attention: Mr. B. Joe Youngblood, Chief Docket No. 50-339 Licensing Branch No. 1 License No. NPF-7 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Daar Mr. Denton:
NORTH ANNA 2 COMPLIANCE WITH NRC REGULATIONS This letter addresses the question whether the North Anna Power Station, Unit 2, complies with all applicable NRC regulations. We understand that this information has been requested by Commissioner Bradford, and that the NRC Staff intends to brief him on the subject next Wednesday, August 13, 1980.
Attachment I to this letter explains where in our license application each regulation in 10CFR Parts 20, 50 and 100 is addressed. We have limited our-selves for the present purpose to those three Parts, as we agreed with your Messrs. Tedesco and Schwencer on August 8, because of their significance to radiological health and safety.
As you will see by examining Attachment 1, Vepco believes that North Anna 2 does comply with the applicable regulations, except in those cases where specific exemptions have been justified and approved by the Staff. We base our confidence in this conclusion on the references in Attachment 1, plus the lengthy review and licensing process that North Anna 2 has undergone. The design process of our architect-engineer, the review of Vepco's own personnel, the quality assurance programs of Vepco and the architect-engineer and NSSS vendor, the independent review of the NRC Staff and ACRS, and the additional independent review of the Atomic Safety and Licensing Boards and Appeal Board, after four sessions of public hearings in the operating license proceeding alone, all together provide reasonable assurance that the public health and safety will be protected.
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vimonwa Eucmc un Powra Couruv to Mr. Harold R. Denton, Director j i
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In addition to the question of compliance with the regulations, the Staff has asked whether Vepco itself has reviewed North Anna 2 against the Staff's l Standard Review Plan (SRP) . The North Anna facility was of course designed, !
and to a large extent constructed, before the publication of the SRP, and Vepco has not systematically reviewed North Anna 2 using the SRP.
If I can be of further service, please let me know.
9 V ry tr y rs, l [lll .fhlW '
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)J.H.Ferguson Exp:utive Vice Presiden
/ Power smv/SJ2 cc: Mr. James P. O'Reilly, Director Office of Inspection and Enforcement Region II Atlanta, Georgia 30303 l
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CITY OF RICHM0h3 )
Before me, a Notary Public, in and for the City and Commonwealth aforesaid, today personally appeared Sam C. Brown, Jr. who being duly sworn, made oath and said (1) that he is Senior Vice President - Power Station Engineering and Construction of the Virginia Electric and Power Company, (2) that he is duly authorized to execute and file this letter in behalf of that Company, and (3) that the statements in the lett<:r are true to the best of his knowledge and belief.
Given under ty hand and notarial seal this M day of M 1 M Y , JW .
v My Commission expires J-n gst v 2 c' rys! .
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4l)hl)fstf Notary Public (SEAL)
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1 E AttcchmInt 1 COMPLIANCE OF NORTH ANNA 2 WITH THE NRC REGULATIONS OF 10CFR PARTS 20, 50 AND 100 Regulation (10CFR) Compliance 20.1(a) his 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 vill 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 Com-pany 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 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 procedures, 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 h e 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 regula-tions by the NRC and does not impose independent obligations on licensees.
20.7 his regulation gives the address of the NRC and does not impose independent obligations on licensees.
20.101 he radiation dose limits specified in this regulation l are complied with through the implementation of and ,
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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.
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1 0 Regulation (10CFR) Compliance 20.102 Wen 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. He 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 equip-ment, as described in Chapters 11 and 12 of the FSAR, are installed and operated to maintain levels of airborne radioactivity as low as practicable. Een necessary, as determined by station administrative guidelines, additional precautionary procedures are utilized to linit the potential for intake of radio-active materials.
20.103(c) The station Respiratory Protection Manual implements the requirements of this regulation by ensuring the proper use of approved respiratory protection equip-ment. He station Respiratory Protection Program incorporates fully the stipulations of Regulatory Guide 8.15, " Acceptable Programs for Respiratory Pro-tection".
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 September 1,1977.
20.103(f) The Respiratory Protection Program is in full confor-mance with'the requirements of 20.103(c).
20.104 Conformance with this regulation is assured by appro-priate Company policies regarding employment of individuals under the age of 18 and the station Health Physics Manual restricting these individuals' access to the station restricted areas.
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I Regulation (10CFR) 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 regu-lation are complied with through the implementation of station 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 regula-tion is assured through the implementation of station procedures and applicable Technical Specifications which provide adequate sampling and analyses, and monitoring of radioactive materials in effluents prior to and during their release. The level of radioactivity in station effluents is minimized to the extent practi-cable by the use of appropriate equipment designed for this purpose, as described in Chapter 11 of the FSAR.
20.106(b) Vepco has not and does not currently intend to include 20.106(c) in any license or amendment applications proposed limits higher than those specified in 20.106(a), as provided far 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(e) This regulation prwides criteria by which the Commis-sion may impose further limitations on releases of radioactive materials made by a licensee. It imposes no independent obligations on licensees.
20.106(f) This regulation merely states that the provisions of 20.106 do not apply to disposal of radioactive material into sanitary sewerage systems. It imposes no indepen-dent obligations on licensees.
20.107 This regulation merely clarifies that the Part 20 regulations are not intended to apply to the inten-l tional exposure of patients to radiation for the purpose of medical diagnosis or therapy. It does not
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impose any independent obligations on licensees.
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Regulation (10CFR) Compliance 20.108 Neceesary bioassay equipment and procedures, including ,
Wole Eody Counting, are utilized at the station to determine exposure of individuals to concentrations of radioactive materials.
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Appropriate health physics procedures and administrative policies implement this requirement.
20.201 The surveys required by this regulation are performed at adequate frequencies and contain such detail as to be consistent with the radiation hazard being evaluated.
Wen necessary, the Radiation Work Permit system estab-lished at the station provides for detailed physical surveys of equipment, structures and work sites to determine appropriate levels of radiation protection.
The station Health Physics knual and applicable health physics procedures require these surveys and provide for their documentation in such manner as to ensure compliance with the regulations of 10CFR Part 20.
20.202(a) The station Health Physics knual and applicable health physics procedures set forth policies and practices which ensure that all individuals are supplied with, and required to use, appropriate personnel conitoring equipnent. The Radiation Vork Pernit syste= is estab-lished to provide additional control of personnel working in radiation areas and to ensure that the level of protection afforded to these individuals is consis-tent with the radiological hazards in the work place.
20.202(b) The terminology set forth in this regulation is accepted and conformed to in all applicable station procedures, Technical Specifications, and those portions of the station Health Physics knual 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 conven-tional design prescribed in this regulation.
20.203(b) This regulation is conformed to through the implemen-tation of appropriate health physics procedures and portions of the Health Physics knual relating to posting of radiation areas, as defined in 10CFR 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 health physics procedures, as well as the station Health Physics knual.
The controls and other protective acasures set forth in the regulation are maintained under the surveillance of the station Health Physics group.
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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 signe: ' re- ;
quired by paragraph 20.203(c)(2) of 100FR20", which '
will prevent unauthorized entry into a high radiation l area. i l
20.203(d) Each Airborne Radioactivity Area, as defined in this l regulation, is required to be posted by provisions of the Health Physics Manual and appropriate health physics procedures. These procedures also provide for the surveillance requirements necessary to determine airborne radioactivity levels.
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 health physics procedures, and portions of the station Health Physics Manual.
20.203(f) The container labeling requirements set forth in this regulation are complied with through the implementa-tion of appropriate health physics procedures, and portions of the station Health Physics Manual.
20.204 The posting requirement exceptions described in this regulation are used where appropriate and necessary at the station. Adequate controls are provided within the station health physics procedures to ensure safe and proper application of these exceptions.
20.205 All of the requirements of this regulation pertaining to procedures for picking up, receiving, and opening packages of radioactive materials are implemented by l the station Health Physics Manual and appropriate health physics procedures. These procedures also provide for the necessary documentation to ensure an auditable record of compliance.
1 i 20.206 The requirements of 10CFR19.12 referred to by this l regulation are satisfied by the General Employee Train-l ing Program conducted at the station. Appropriate l health physics procedures set forth requirements for
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all radiation workers to receive this instruction on a periodic basis.
20.207 The storage and control requirements for licensed I
materials in unrestricted areas are conformed to and documented through the implementation of station health physics procedures and applicable portions of the station Health Physics Manual.
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Regulation (10CFR) Compliance 20.301 The general requirements for waste disposal set forth in this regulation are complied with through station health physics procedures, the Technical Specifications, and the provisions of the station license. Chapter 11 of the FSAR describes the Solid Waste Disposal System installed at the station.
20.302 No such application for proposed disposal procedures, as described in this regulation, has been made or is currently contemplated by Vepco.
20.303 No plans for waste disposal by release into sanitary sewerage systems, as provided for in this regulation, are contemplated by the station, nor is this practice currently utilized.
20.304 Disposal of wastes by burial in soil (i.e., onsite burial), as provided for in this regulation, is not performed currently or being contemplated by the station.
20.305 Specific authorization, as described in this regulation, is not currently being sought by Vepco 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, radiation monitoring and waste disposal. The retention periods specified for such records are also provided for in these specifications and procedures.
20.402 The station has established an appropriate inventory and control program to ensure strict accountability for all licensed radioactive materials. Reports of thef t or loss of licensed material are required by reference to the regulations of 10CFR in the Technical Specifica-
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20.403 Notifications of incidents, as described in this regula-tion, are assured by the requirements of the Technical Specifications, the station Health Physics Manual and appropriate health physics procedures, which also provide for the necessary assessments to determine the i occurrence of such incidents.
20.404 This regulation was deleted effective September 17, 1973 (38 Fed. Reg. 22220).
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Regulation (10CFR) Compliance 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.
20.406 his regulation was deleted August 17, 1973, effective September 17, 1973 (38 Fed. Reg. 22220).
20.407 he personnel uonitoring report required by this regulation la espressly 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 todiation exposure required by this regulation upon termination of an individual's employ-ment or work assignment is generated through the provisions of a station health physics procedure.
20.409 he notification and reporting requirements of this regulation, and those referred to by it, are satisfied by the provisions of a station health physics procedure.
20.501 his regulation provides for the granting of exemptions from 10CFR Part 20 regulations, provided such exemp-tions are authorized by law and will not result in undue hazard to life or property. It does not impose independent obligations on licensees.
20.502 This regulation describes the means by which the Com-mission 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 Com-mission may obtain in order to enforce its regulations, and sets forth those penalties or punishnents which may be imposed for violations of its rules. It does not impose any independent obligations on licensees.
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Regulation (10CFR) Compliance P
50.1 This regulation merely states the purpose of the Part 50 regulations and does not impose any independent obliga-tions 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 obligations 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 50.11 may not be imdertaken without a license from the NRC.
Vepco does not propose to conduct any such activities at North Anna 2 without an NRC license.
50.12 This regulation provides for the granting of exemptions from 10CFR Part 50 regulations, provided such exemp-tions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. It does not impose independent obligations on licensees.
50.13 This regulation says that a license applicant need not design against acts of war. It imposes no independent obligations on licenses.
50.20 These regulations merely describe the types of licenses 30.21 that the NRC issues. They do not address the substan-50.23 tive 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. Vepco has substantial'y complied with the proce-dural requirements in effc.ct at the time when filing its license application and the amendments to it. In particular, 10CFR50.30(f) requires that a license application must be accompanied by any Environmental Report required pursuant to 100FR Part 51, and Vepco has submitted an Environmental Report covering North Anna 2.
50.31 These regulations merely permit more efficient organi-50.32 zation of the license application and impose no independent obligations on licensees.
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l Regulation (10CFR) Compliance 50.33 This regulation requires the license application to contain certain general information, such as an identi-fication of the applicant, information about the applicant's financial qualifications, and a list of regulatory agencies with jurisdiction over the appli-cant's rates and services. This information is provided in Part A of Vepco's operating license application for North Anna 1 and 2.
50.33a This regulation requires applicants for construction permits to submit information required for antitrust review. As is noted in Part A of Vepco's application for an operating license for North Anna 1 and 2 at page NIF-A 8-1, the antitrust review required under Section 105c of 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 construc-tion 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 information required:
(1) site evaluation factors - Chapter 2 (2) structures, e,ystems and components - Chapters 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 15 (3) radioactive effluents and radiation protection -
Chapters 11, 12 (4) design and performance evaluation - ECCS perfor-mance is discussed and shown to meet the require-ments of 10CFR50.46 in Chapters 6 and 15.
(5) results of research programs - Chapter 1 (6) (i) organizational structure - Chapter 13 (ii) managerial and administrative controls - Chap-ters 13 and 17. Chapter 17 discusses compliance with the quality assurance requirements of Appendix B.
(iii) plans for preoperational testing and initial operations - Chapter 14
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l Regulation (10CFR) Compliance (iv) plans for conduc't of'ncemal operations -
Chapters 13 nad 17. Surveillance and periodic testing is spe2ified in the Technical Speci-fications.
(v) plans for coping with emergencies - Faergency Plan (Chapter 13)
(vi) Technical Specifications - prepared in con-junction with the Staff (Chapter 16)
(vii) not applicable, since the operating license application was filed prior to February 5, 1979 (7) technical qualifications - Chapter 13 (8) operator requalification program - Chapter 13 and letters of 6-27-80 and 7-15-80 (incorporated into the application by Amendment 69). 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 cur-rently implemented.
50.34(d) A safeguards contingency plan is included as Chapter 8 of the physical security plan.
50.35 This regulation is relevant to the construction permit stage rather than the operating stage.
50.36 Technical Specifications have been prepared and imple-mented, including items in each of the categories specified, including (1) safety limits and limiting safety settings, (2) limtting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.
50.36a The Environnental Technical Specifications, Part I, include specifications which require compliance with 10CFR50.34a (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 10CFR20.106. The reporting requirements of 50.36a(a)(2) are also included in these specidications.
SL' 37 Tnis regulation requires the applicant to agree to limit access to Restricted Data. Vepco's agreement to do so is in Part A of the operating license application for North Anna 1 and 2 at page N1F-A 7-1.
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Regulation (10CFR) Compliance ;
50.38 This regulation prohibits the NRC from issuing a license I to foreign-controlled entities. Vepco's statement that )
it is not owned, controlled, or dominated by an alien, foreign corporation, or foreign government is in Part A of the operating license application for North Anna 1 and 2 at page NIF-A 3.1-5.
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 granting licenses, as follows:
50.40(a) The design and operation of the facility is to provide reasonable assurance that the applicant will comply with NRC regulations, including those in 10CFR Part 20, and that the health and safety of the public will not be endangered. The basis for Vepco's assurance that the regulations will be met and the public protected is contained in this Attachment and in the license appli-cation and the related correspondence over the years.
Moreover, the lengthy process by which the plant is designed, constructed, and reviewed, including reviews by the architect-engineer, Vepco'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, af ter an extensive review, concluded that Vepco had the commitment and technical qualifications necessary to operate North Anna 1 and 2 safety and in compliance with all appli-l cable radiological health and safety requirements (see below).
l 50.40(b) Another consideration is that the applicant be tech-nically and financially qualified. Both Vepco's i technical qualifications and its financial qualifica-tions were examined in great detail at hearings before the Atomic Safety and Licensing Board. The Board's findings on those matters are contained in its decision.
l Virginia Electric and Power (North Anna Nuclear Power Station, Units 1 and 2), LBP-77-68, 6 NRC 1127, 1150, 1177 (Dec. 13, 1977), aff'd., ALAB-491, 8 NRC 245 (Aug. 25, 1978).
50.40(c) Another consideration is that the issuance of the l li:ense is not to be inimical to the common defense and se'urity or to the health and safety of the public. The
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(10CFR) Compliance individual showings of compliance with particular regu-lations contained in this Attachment, 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 Vepco, its architect-engineer, its NSSS vendor, and the government, provide Vepco 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 Vepco has a qualifying security plan for the power station, that Vepco is not controlled by agents of foreign countries, and that Vepco has agreed to limit access to Restricted Data (see above).
Moreover, the Atomic Safety and Licensing Board at the construction permit stage expressly found that "the issuance of permits for the construction of the facilities will not be inimical to the common defense and security nor to the health and safety of the public",
Virginia Electric and Power Company, (North Anna Nuclear Power Station, Units 1 and 2), 3 CCH Atomic Energy L.
Rep. 111,593 (Feb. 9, 1971). To Vepco's knowledge, this conclusion of the board has never been challenged, and no evidence contrary to it has ever been presented.
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. Vepco has submitted an Environ-mental Report, the NRC Staff has issued an Environ-mental Impact Statement, and Environmental Technical i
Specifications have been issued for North Anna 2.
I l 50.41 This regulation applies to class 104 licensees, such as those for devices used in medical therapy. Vepco 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 considerations are (a) that the proposed l activties 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
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be taken of the antitrust device 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 construc-I tion permit stage, Virginia Electric and Power Company (North Anna Power Station Units 1 and 2), LBP-74-81,
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Regulation (10CFR) Compliance 8 AEC 773, 799-801 (Oct. 30,1974) . Although conditions affecting the need for power are constantly changing, Vepco periodically makes load projections, and in Vepco's judgement the need for North Anna 2 is still substantial.
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As for the amount of special nuclear material or source .
material used, there is no reason to believe that their I proportion in relation to the power produced is sub-stantially greater than that of other commercial power reactors in this country. As for the antitrust advice of the Attorney General, as noted above, the antitrust review was done at the construction permit stage.
50.43 This regulation imposes certain duties on the NRC and addresses the applicability of the Federal Power Act and the right of government agencies to obtain NRC licenses. It imposes no direct obligations on licensees.
50.44 The North Anna combustible gas control system is de-scribed 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 10CFR50.44(e).
The system is also designed to purge the containment at a controlled rate, through the gaseous waste dis-posal system. The system, shared between Unit 1 and 2, consists of two identical portable skid mounted hydrogen recombiners, two hydrogen analyzers, two purge blowers, and associated piping systems. In accordance with the requirements of NUREG-0660 and NUREG-0694, the hydrogen analyzers are being replaced with new units qualified to IEEE-323,1974 and IEEE-344,1974.
Section 6.2.5.3 of the FSAR indicates that the recom-biners can be placed into operation in sufficient time to maintain the hydrogen concentration following a LOCA at a level below the 4 percent lower hydrogen fla.,ma-bility limit. In accordance with 50.44 (a) (2) , the
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hydrogen contribution of the core metal-water reaction I is assumed to be that resulting from reaction of 5 percent of the fuel cladding. FSAR Section 15.4.1 demonstrates that the North Anna ECCS system complies with the core metal-water reaction limits of 10CFR50.
46(a)(6), and thus does not exceed 1% of the total amount of zircaloy in the reactor. Consequently, the requirements of 50.44 are satisfied.
50.45 This regulation provides standards for construction permits rather than operating licenses and is therefore not material to this operating license proceeding.
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(10CFR) Compliance 50.45 FSAR Sections 6.3 and 15.4.1 describe the Energency Core Cooling System and the methods used to analyze '
ECCS performance following a postulated loss of coolant l accident.
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By letter dated November 29, 1979, Vepco provided the results of a LOCA-ECCS analysis for North Anna Units 1 and 2, using an NRC approved evaluation model which is in compliance with Appendix K to 10CFR50. He analysis, based on an overall peaking factor (Fq) of 2.10, pro-vided results in compliance with the criteria of 10CFR50.46(b) . The Fq limit is reflected in the Tech-nical Specifications.
The Staff has proposed new standards for fuel cladding swelling and rupture models in draft NUREC-0630. If these standards were to be adopted for the currently approved evaluation models, a decrease in the Fq Technical Specification limit would be required. How-ever, by letter dated December 19, 1979, Vepco provided an assessment of this reduction as well as the benefits of modeling improvements described to the NRC by Westinghouse Electric Corporation. It was concluded that application of these improvements would more than offset any Fq reduction resulting from the revised staff cladding model; therefore, the existing Technical Specification limit is acceptable. The Staff in SER Supplement 10 has agreed with this conclusion.
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 number 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 North Anna 2 will be within the United States and sub-t ject to the jurisdiction of the United States, as is
! evident from the description of the facility in Part A of the operating license application.
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Regulation (10CFR) Compliance 50.54 This regulation specifies certcin conditions that are incorporated in every license issued. Compliance is effected simply by including these conditions in the license when it is is sued , and in fact the low-power license already in effect for North Anna 2 expressly says that it is subject to the conditions in Section 50.54 of 10CFR Part 50 (see 12.D of the North Anna 2 license for fuel loading and low-power testing). In-deed, much of 50.54 merely provides that other provisions of the law apply, which would be the case even without 50.54.
50.55 This regulation addresses conditions of construction permits, not operating licenses, and so it is not material 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.
50.55a(a) (2) This paragraph is a general paragraph leading into paragraphs (c) through (i) of the regulation.
50.55a(b) (1) These paragraphs provide guidance concerning the 50.55a(b) (2) approved 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 fo;ad in Chapter 5, Sections 5.2 and 5.4 of the FSAR.
50.55a(d) Reactor coolant system piping meets the requirements of USAS B31.7 (1969). Information can be found in Chapter 5, Section 5.2 of the FSAR.
50.55a(e) Reactor Coolant pumps meet the requirements of ASME III (1968). Information can be found in Chapter 5, Section 5.2 of the FSAR.
50.55a(f) Based upon the original projected construction permit date of October,1969, the valves within the Reactor Coolant System pressure boundary were ordered and supplied in accordance with the requirements of USAS B31.1-1967, plus addendum USAS B16.5, and MSS-SP-66.
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Regulation (10CFR) Compliance 50.55a(g) Inservice Inspection (ISI) requirements are delineated in this part and are specified in the Unit 2 Technical Specifications, paragraph 4.0.5. As permitted by this i part and the Technical Specifications, certain exemp- l tions have been requested and granted for the inservice inspection of various systems and the inservice testing !
of various pumps and valves. The following letters from the licensee provide further information.
Preservice Inspection Letter Serial No. 651, 12-4-78 Inservice Testing of Pumps and Valves Letter Serial No. 046, 1-31-79 Inservice Inspection Letter Serial No. 112, 3-12-79 In Supplement 10 of the Staff's SER for North Anna, Sections 5.2.11 and 5.2.12 provide a discussion of preservice inspection and inservice testing of pumps and valves. In particular, the pump and valve program has been accepted pending certain further review.
Approval of the inservice inspection program is not required at this time. Additional information on ISI can be found in FSAR Chapter 5. Section S.2.5.
50.55a(h) As discussed in Chapter 7, Sections 7.1.2.2.3, 7.2.2.2.1, 7.1.2.3.3, and 7.3.1.2, the protection systems meet IEEE 279-1971.
50.55a(i) Fracture toughness requirements are set forth in Appendices G and H of 10CFR50. Technical Specifications require the use of reactor vessel material irradiation surve 111ance specimens and updating of the "heatup" and "cooldown" curves given in the Technical Specifica-l tions. Further information is given in FSAR Chapter 5, l 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 Fed. Reg. 26297, and in any event it would apply to fuel reprocessing plants, not to facilities such as North Anna 2.
50.56 This regulation provides that the Commission will, in the abse ce of good cause shown to the contrary, issue an operating license upon completion of the ccastruc-tion of 3 facility in compliance with the terms and conditions of the construction permit. This imposes no independent oblige ions on the applicant.
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Regulation (10CFR) Compliance 50.57(a) This regulation requires the Commission to make certain findings prior to the issuance of an operating license.
These findings have already been made for North Anna 2 in connection with the issuance of license NPF-7 for fuel loading and low power testing, and they can also be made for full power operation for the reasons given in this Attachment generally. Specifically:
(1) Construction of the facility has been substan-tially completed in conformity with the construction permit and the application as amended. Conformance of the facility to the NRC rules and regulations and the Act, as imple-mented by the regulations, has been demonstrated by the application.
(2) The Technical Specifications and resulting operating procedures provide assurance that the unit will operate in conformity with the appli-cation as amended and with the rules and regula-tions, with one exception (to Appendix J of 10CFR50), as documented in License NPF-7.
(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 Virginia Electric and Power Company is technically and financially qualified to operate the unit, as confirmed by the initial decision of the Atomic Safety and Licensing Board.
(5) The applicable provisions of 10 CFR 140 have been satisfied. The indemnity agreement B-80 for the North Anna Power Station has been amended to include Unit No. 2.
(6) The approved Security Plan assures that special nuclear material is being appropriately safe-guarded. The application demonstrates that the operation of the unit will not be inilical 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 Commission.
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Regulation (10CFR) Compliance 50.57(c) This regulation provides for a low-power testing license.
Such a license has already been issued for North Anna 2.
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 North Anna 2 in accordance with its usual practice.
50.59 his regulation provides for the licensing of certain changes, tests, and experiments at a licensed facility.
Technical Specifications and procedures provide imple-mentation of this regulation.
50.60 h is regulation has been deleted, 40 Fed. Reg. 8790.
50.65 his regulation has been deleted, 43 Fed. Reg. 6915.
50.70 The Commission has assigned resident inspectors to the North Anna Power Station. Vepco has provided office space in accordance with the requirements of this section. Vepco pennits access to the station to NRC inspectors in accordance with 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 regu-lation.
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 North Anna 2 proceeding.
50.81 This regulation permits the creation of mortgages, pledges, and liens on licensed facilities, subject to certain provisions. The regulation prohibits secured creditors from violating the Atomic Energy Act and the Commission's regulations. To the best of our knowledge, none of Vepco's creditora has attempted to violate the applicable provisions of law.
50.82 h is regulation provides for the termination of licenses.
It does not apply to North Anna 2 because Vepco has not requested the termination of a license.
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Regulation (10CFR) - Compliance 50.90 This regulation governs applications for amendments to licenses. Future requests for license admendments will be made in accordance with these requirements.
50.91 This regulation provides guidance to the NRC in issuing license amendments.
50.100 These regulations govern the revocation, suspension, and 50.101 modification of licenses by the Commission under unusual 50.102 circumstances. No such circumstances are present in the 50.103 North Anna 2 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 require-ment, 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 North Anna 2 proceeding, and so this regulation is not applicable.
Appendix A GDC 1 Section 3.11 of the FSAR describes the design provi-sions - made to ensure that these requirements are met.
i Codes and standards utilized for the unit are specified
,
throughout the FSAR. Chapter 17 describes the quality l 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 nonnal and accident conditions with the effects of natural phenomena, and the importance of the safety functions to be performed.
GDC 3 FSAR Section 3.1.3 describes in general the measures which have been taken to minimize the probability and effects of fires and explosions. Section 9.5.1 de-scribes the fire detectior 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.
l These modifications will be completed by November 1, l 1980, with the exception of an alternate shutdown
! systen, which will be completed by April,1981. This
! schedule is in accor I nce with the Commission's Order l- dated May 23, 1980.
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Regulation (10CFR) Compliance GDC 4 FSAR Section 3.1.4 describes the design features used to accommodate the effects of and be compatible with the environmental conditions associated with all modes of operation and postulated accidents. Chapter 3 provides information concerning the specific design features for protection against missiles, jet impinge-ment and pipe rupture. Provisions for qualification of equipment for all postulated environments is described in several sections of the FSAR, particularly Appendix 3C and S7.17. In addition, the results of a review of all electrical equipment against the new Staff criteria in NUREG-0588 have been incorporated into the application in Amendment 69. This review has confinned that most electrical equipment has been adequately demonstrated to be qualified for its expected service environments. Forty-eight of the remaining 57 items have been classified by the manufacturers as being qualified, and confirmatory test data will be obtained by November 1,1980. Justification in the form of additional evaluations and documentation has been pro-vided to the staff for operation of the unit pending completion of the environ:nental qualification review.
This will be completed on a c hedule in accordance with the Commission's Memorandum and Order of May 23, 1980, and applicable license conditions.
GDC 5 As described in FSAR Section 3.1.5, those structures, systems and components which are shared with Unit 1 are tabulated in FSAR Section 1.2.11. It is concluded that safety functions are not significantly impaired by such sharing.
GDC 10 FSAR Section 3.1.6 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 3.1.6, and Chapters 4, 7, and 15.
GDC 11 FSAR Section 3.1.7 indicates that promit compensatory reactivity feedback effects are assured by unit design and operational limit considerations. ile core inherent reactivity feedback characteristics and reactivity con-trol methods are described in FSAR Sect;o-J., 4.3, 4.2.3, and 9.3.4.
- GDC 12 FS/R Section 3,1.8 describes the inherent and design features which eliminate or limit the various types of oscillations. Core stability is further described in Section 4.3.
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Regulation (10CFR) Compliance GDC 13 As indicated in FSAR Section 3.1.9, and described in more detail in Chapter 7, instrumentation and control systems have been provided to monitor and maintain plant variables including those variables which affect the fission process, integrity of the reactor core, the reactor coolant pre 7sure boundary, and the containment, over their prescribed ranges for normal operation, anticipated occcurrences, and under accident conditions.
GDC 14 FSAR Section 3.1.10 indicates that the reactor coolant pressure boundary has been designed to accommodata the system temperatures and pressures attained un ~ ' all expected operational modes and anticipated tria tents, and to maintain stresses within applicabl limits.
Design criteria and methods are described in Sections 3.1.10, 5.2.1, 3.6, 3.7, 5.2.2, 5.2.4, and 5.2.5.
GDC 15 As indicated in FSAR Section 3.1.11, 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 5. Details concerning the protection systems are provided in Section 7.2.2.
GDC 16 As described in FSAR Section 3.1.12 and Chapter 6, a reinforced concrete, steel-lined containment structu .,
operating at a subatmospheric pressure, is prc.ided.
It is designed to sustain, without loss of tequired 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 capability of this criteria.
GDC 17 As described in FSAR Section 3.1.13, 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 Sections 3.1.14, 8.1.4, 8.3.1.2.2, and 8.3.2.2, 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 gene-rators provide automatic backup power sources. Periodic tests of the diesel generators, the transfer system and the station batteries are made, as required by Technical Specifications.
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Eagulatica (10CFR) Compliance i
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GDC 19 FSAR Section 3.1.15 describes the Main Control Room, which contains the controls and instrumentation neces-
- sary for safe operation of the unit during normal and i accident conditions. Sufficient shielding, distance, structural integrity, and ventilation systems are 4
I provided to ensure that control room personnel will not receive radiation exposures in excess of the criterion i
for the duration of the accident. In the event that access to the main control room is restricted, auxiliary shutdown panels have been provided,' within the protected envelope, which may be used to maintain the reactor in a hot shutdown condition. Subsequent cold shutdown may be achieved using suitable procedures.
GDC 20 FSAR Section 3.1.16 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 Sections 7.2.2.1.1, 7.2.2.1.2, and 7.3.1.
GDC 21 As indicated in FSAR Section 3.1.17, the protection system is designed for the high functional reliability and inservice testability commensurate with the safety functiont to be performed. This section, as well as Sections 7.2.2.1.2, 7.2.2.2.1, 7.3.1 and 7.3.2, describe in detail the design features provided to ensure redundancy and testability.
. CDC 22 FSAR Section 3.1.18 indicates that the protection system
! has been designed to provide sufficient resistance to a
, broad clara of accident conditions or postulated events.
4 Sections 7.2.2.2.1 and 7.3.1 provide further design l details concerning this resistance such that indepen-dence is maintained.
l GDC 23 As indiceted in FSAR Section 3.1.19, the protection
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system is designed with due consideration of the most probable lailure modes of the components under various perturbations of energy sources and the enviromnent.
Further details are supplied in Sections 7.2.2.2.1, 7.3.1, and 7.3.2.
I GDC 24 FSAR Section 3.1.20 discusses separation of the protec-tion and control systems, such that the failure of any single 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 all redundancy, reliability, and independence requirements of the protection system. Details concern-ing separation of protection and control systems are provided in Sections 7.2.2.2.1 and 7.3.1.
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l Regulation (10CFR) Compliance GDC 25 FSAR Section 3.1.21 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 accidental withdrawal of control cluster groups.
Further details are provided in FSAR Sections 4.3.14 and 7.7.2.2.
GDC 26 As indicated in FSAR Section 3.1.22, two independent reactivity control systems of different design princi-ples are provided. One of the systems uses control rods; the second system employs dissolved boron as a chemical shim. Reactivity control cystem redundancy and capability are described further in Sections 4.3.1.5 and 7.7.2.2.
GDC 27 As described in FSAR Section 3.1.23, 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 4.2.3 and 9.3.4.
GDC 28 FSAR Section 3.1.24 indicates that core reactivity is controlled by a chemical poison dissolved in the coolant, rod cluster assem111es and burnable poison rods. The maximum reactivit, insertion rates due to withdrawal of a bank of rod cluster control assemblies or by boron dilution are limited. The maximum worth of control i rods and the maximum rates cf reactivity insertion l employing control rods are limited to values which j prevent rupture of the coolant pressure boundary or i disruption of the core internals to a degree which l
would impair core cooling capacity. Further details are provided in Section 4.3.1.4.
GDC 29 As indicated in FSAR Section 3.1.25, the protection and reactivity 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 4.2.3 and 7.7.
l l GDC 30 As described in FSAR Section 3.1.26, reactor coolant pressure boundary components are designed, fabricated, inspected and tested in conformance with ASME Nuclear Power Plant Components Code Section III. Major com-
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ponents are classified as seismic Class 1 and are
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Regulation (10CFR) Compliance 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 3.1.27, close control is maintained . over material selection and fabrication for the Reactor Coolant System to assure that the boundary
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behaves in a nonbrittle manner. The materials testing is consistent with 10CFR50, Appendices 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 3.1.28 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 inspection of the external surfaces of the reactor coolant piping, except for the area of pipe within the primary shielding concrete. Additional details can be found in Section 5.2.
GDC 33 As indicated in FSAR Section 3.1.29, the Chemical and ,
Volume Control System provides a means of reactor i coolant makeup and adjustment of the boric acid con- )
centration. 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 9.3.4 and details of the electrical power systems are given in Sections 8.2 and 8.3.
GDC 34 FSAR Section 3.1.30 indicates that the Residual Heat
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Removal System, in conjunction with the steam and power
- conversion systen, is designed to traasfer the fission product decay heat and other residua:1 heat from the reactor core within acceptable limits. Suitable redun-dancy 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 Systen is able to operate 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 Best Removal System design are in FSAR Section 5.5.4.
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Regulation (10CFR) Compliance GDC 35 FSAR Section 3.1.31 describes the use of passive accumu-lators 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 Faergency 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 Sections 7.3.1.3.5 and 6.3.
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GDC 36 As described in FSAR Section 3.1.32, 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 visually or by boroscopic examination for erosion, corrosion, and vibration wear evidence. Non-destructive inspection is performed where such techniques are desirable and appropriate. Technical Specifications require inservice inspection in accordance with applicable ASME Codes.
, Details of the inspection programs are provided in l Sections 5.4 and 6.3.4.
GDC 37 FSAR Section 3.1.33 indicates that the components of the Emergency Core Cooling System located outside the contaianent will be accessible for leak-tightness 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 cen-trifugal charging pumps are part of the charging syste=,
and this system is in continuous operation during plant operation. Actuation circuits are tested and remote cperated valves are exercised periodically. The test-ing is described in detail in FSAR Section 6.3.4 and as per Technical Specification surveillance requirements.
GDC 38 As indicated in FSAR Section 3.1.34, two Quench Spray Subsystems and four separate Recirculation Spray Sub-systems are provided to remove heat from the containment following a loss-of-coolant-accident. Each subsystem contains a separate pump and spray header, and each Recirculation Spray Subsystem contains a separate
, cooler. 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.
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Regulation (10CFR) Compliance GDC 40 As described in FSAR Section 3.1.36, provision has been made to allow testing the Quench Spray Subsystem and the Recirculation Spray Subsystem throughout the life of the unit to ensure that the systems are operational.
In order to verify that the Containment Depressuriza-tion System will respond promptly and perform its design function, the Quench Spray and Outside Recirculation Spray pumps are flow tested, the motor operated valves in the Containment Depressurization System are tested, and the Quench Spray Subsystem is tested for the presence of particulate matter which could clog the spray nozzles. These tests are performed periodically during the life of the plant. Additional discussion is provided in Sections 6.2.2.4 and 7.3.2.
GDC 41 As indicated by FSAR Section 3.1.37, systems are pro-vided to control hydrogen and fission products generated by a design basis accident as described in Chapter 6 of the North Anna FSAR. These systems are sufficiently redundant to meet the single failure criterion and are operable vith either onsite or offsite power. The Containment Atmosphere Cleanup System maintains hydrogen below the flam=able limit of 4 volume percent while the caustic sprays from the Quench Spray Subsystem remove radioactive iodine and particulate fission products by absorption and washing action. Sections 6.2.3 and 6.2.5 discuss containment atmosphere cleanup in more detail.
GDC 42 FSAR Section 3.1.38 indicates that the Containment Atmosphere Cleanup System and the Containment Depres-surization System are designed to permit appropriate periodic inspection of the important components.
Additional discussion is provided in FSAR Sections 5.4 and 6.2.
CDC 43 FSAR Section 3.1.39 indicates that the Containment Atmosphere Systan and the Containment Depressurization Systen are designed to permit periodic pressure testite and functional testing of their components. Further details are provided in Sections 7.3.2, 6.2.2.4, and 6.2.5.4.
I GDC 44 FSAR Section 3.1.40 describes how all safety related items requiring cooling during an accident are cooled by the Service Water System. Beat exchangers requiring cooling during normal operation and cooldown are cooled by either the Component Cooling Systen or the Service Water System. The Service Water System has sufficient redundancy to enable the systen to meet the single failure criterion, including failure of an emergency
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d Regulation (10CFR) Complience
- generator. The Component Cooling System ic provided with redundant pumping and heat transfer equipment and will operate with anergency onsite power. Discussion
! of these systems is provided in FSAR Chapter 9.
GDC 45 As indicated in FSAR Section 3.1.41, the Service Water
,
System and the Component Cooling Systen transfer heat 1
from structures, systans, and components important to safety to an ultimate ' heat sink. Inspection of the majority of header piping in the Service Water System is not anticipated since it is buried and encased in concrete. All remaining piping, valves, equipment, and
,
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associated electrical gear from the Service Water System can be readily inspected. All piping, valves, equipment,
- and associated electrical gear for the Component Cooling System can be readily inspected. Further discussion of these systems is provided in FSAR Chapter 9.
CDC 46 As described in FSAR Section 3.1.42, the cooling water
! system referred to in this criterion encompasses the
! Service Water System and the Component Cooling System.
l The service water supply to the recirculation spray heat exchangers is tested periodically to ensure that the automatic valves function as required and the leak tight and structural integrity of the pressure containing components is retained. The Component Cooling System is in continuous use, thereby ensuring that the struc-tural integrity, operability of active components, and operability of the system in its entirety is monitored continuously. Details of these systems are found in Chapter 9.
GDC 50 FSAR Section 3.1.43 indicates that the containment structure is designed to leak less than 0.1 percent by volume of its contents per day under post design basis accident conditions and to tithstand pressures and temperatures above those that are conservatively calcu-lated to result from a design basis accident by a sufficient margin to ensure that design conditions are not exceeded. Containment design basis is discussed further in Sections 3.8.2, 6.2.1, and 15.4.
I GDC 51 As discussed in FSAR Section 3.1.44, the design condi-tion of the containment pressure boundary is based on a
the parameters derived after the design basis accident.
i For this design condition, the steel liner material behaves in a nonbrittle manner and has the capability l to minimize the propagation of any undetected flaw.
- Detailed information on the steel liner material is
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found in Section 3.8.2.
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Regulation (10CFR) Compliance GDC 52 As indicated in FSAR Section 3.1.45, the containment
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has been subjected to an "one time only" air pressure test at 115 percent design pressure. Periodic inte-grated leakage rate tests will be performed as required by 10CFR Appendix J, as implemented by the Technical Specifications.
GDC 53 FSAR Section 3.1.46 indicates that the contaitunent design includes provisions for testing the leak tightness of all penetrations. Penetrations with resilient seals will be visually inspected and pressure
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tested. Penetrations with expansion bellows will be pressure tested. Test channels for checking the weld between penetrations and the containment liner have been provided, thereby allowing surveillance of the conditions inside the containment. Additional details are provided in Sections 3.8.2 and 6.2.1.
GDC 54 As described in FSAR Section 3.1.47, the Containment Isolation System provides at least two barriers between the atmosphere outside the containment structure and either the atmosphere inside the containment structure i or the fluid inside the reactor coolant pressure bound-ary during accident conditions. Operation of the Containment Isolation Systen is automatic. The failure of one barrier or valve does not prevent isolation.
Means are provided to periodically test the sensor set-points, the speed of response, the operability of f ail-safe features, and the leakge rates of all valvas used for containment isolation. Further discussion is i provided in Section 6.2.4.
GDC 55 As indicated in FSAR Section 3.1.48, all pipe penetra-tions throt.6h the contairunent structure have at least
! two barriers between the atmosphere outside the contain-ment and either the atmosphere inside the contairunent structure or the fluid inside the reactor coolant '
pressure boundary during accident conditions. The piping is designed to Class I or II of the USA Standard Code for Pressure Piping - ANSI 2 H ,7-1969, Nuclear i Power Piping. The isolation valves outside the contaitument are located as close as practical to the penetration. All valves used to ef fect ' contaitunent isolation may be readily inspected at any time. Addi-tional details of the Containment Isolation System are
. provided in Section 6.2.4.
I GDC 56 As indicated in FSAR Section 3.1.49, all pipt penetra-tions thro'agh the containment structure have at least
,
two barriers between the atmosphere outside the contain-
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ment and either the atmosphere inside the containment
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..
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Regulation (10CFR) Compliance structure or the fluid inside the reactor coolant pressure boundary during structural accident conditions.
The piping is designed to Class I or II of the USA Standard Code for Pressure Piping - ANSI B31.7-1969, Nuclear Power Piping. The isolation valves outside the containment are lo.:ated as close as practical to the penetration. All valves used to effect containment isolation may be readily inspected at any time. Addi-tional details of the Containment Isolation System are provided ia Section 6.2.4.
GDC 57 FSAR Section 3.1.50 indicates that each pipe penetration 1 through the containment structure has at least two barriers between the atmosphere outside the containment and either the atmosphere inside the containment struc-ture or the fluid inside the reactor coolant pressure boundary during atecident conditions. The isolation valves outside the containment are located as close as practical to the penetration. Further details regard-ing the Containment Isolation System are provided in FSAR Section 6.2.4.
In reference to GDC 55, 56, and 57, the Atomic Safety and Licensing Board has concluded that the departures from current criteria for pipe penetrations have been described and justified by Vepco and that the NRC Staff has described its bases for accepting the departures.
Virginia Electric and Power Company (North Anna Nuclear Power Station, Units 1 and 2), LBP-77-68, 6 NRC 1127, 1133 (Dec. 13, 1977).
- GDC 60 As described in FSAR Section 3.1.51, the control of waste gas effluents is accomplished by holdup of waste '
gases in decay tanks until the activity of the tank contents and the existing environmental conditions permit discharges within the requirements of 10CFR20 and 10CFR50. Sufficient holdup of waste gases is l
'
provided to cope with all anticipated operational occurrences and all site environmental conditions.
, Control of liquid waste effluents is accomplished by holdup of waste liquids in storage tanks. Liquid effluents are monitored for radioactivity and rate of flow. Sufficient holdup of liquid wastes is provded to cope with all anticipated operation occurrences and all site environmental conditions. Sufficient handling capacity is provided for solid wastes, which will be prepared batch-wise for offsite disposal, to cope with all anticipated operational occurrences. Further details are provided in FSAR Qiapter 11.
GDC 61 FSAR Section 3.1.52 indicates that systems which may contain radioactivity are - designed to ensure adequate safety under normal and postulated accident conditions.
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l
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Regulation l (10CFR) Compliance The systems are designed to permit inspection and test-ing as described in FSAR Chapters 5, 6, 9, and 11. The systems and components are provided with suitable i shielding for radiation protection. To preclude gross i mechanical failures which could lead to significant {
radioactivity release, appropriate containment, con- ;
finement, and treatment facilities and procedures are !
provided. Residual heat removal capability which has I reliability and testability is provided. The fuel pit storage, fuel pit cooling, and fuel pit water makeup
- systems are designed to prevent significant reduction in fuel pit water inventory under accident conditions.
GDC 62 This criterion requires that criticality in the fuel storage and handling system be prevented by physical
, systems and processes, preferably by the use of geomet -
4 rically safe configurations. As noted in FSAR Sections 3.1.53 and 9.1, the new and spent fuel handling, transfer and storage equipment and facilities are designed and arranged to provide sufficient center-to-center spacing to preclude criticality. The fuel tonsfer equipment is designed to handle only one fuel assembly at a time. The wate.: used in the reactor cavity when the reactor vessat head is removed is maintained with a sufficient 'oorot. concentration to ensure a k effective less than 0.99 even if all control rods are withdrawn, as required by Technical Specifica-tions.
3 GDC 63 This criterion specifies that fuel storage and radio-active waste areas be monitored for loss of residual heat renoval capability and for excessive radiation levels, and that appropriate safety actions.be taken if such conditions arise. FSAR Section 3.1.54 and Chapters 9 and 11 describes 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 )f these monitors is received.
l
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GDC 64 This criterion requires means be provided for monitoring radioactive releases. FSAR Section 3.1.55 indicates that monitoring is provided for the containment atmos-phere, effluent discharge paths, and the safeguards areas. Further details are provided in Chapters 11 and 12.
1
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Regulation (10CFR) Compliance Appendix B Chapter 17 of the FSAR describes in detail the provi-sions of the quality assurance program which has been implemented to meet all applicable requirements of Appendix B. Evidence on the quality assurance program is in the record of the operating license hearings.
See Virginia Electric and Power Company (North Anna Nuclear Power Station, Units 1 and 2), LBP-77-68, 6 NRC 1127, 1136-37 (Dec. 13, 1977).
Appendix C his Appendix provides a guide for establishing the applicant's financial qualifications. Vepco's finan-cial qualifications were fully litigated before the Atomic Safety and Licensing Board, and the Board expressly found that Vepco had satisfied the burden of proving that it has reasonable assurance of having the funds that it needs to operate the facility in compli-ance with the Commission's regulations. Virginia Electric and Power Company (North Anna Nuclear Power Station, Units 1 and 2), LBP-77-68, 6 NRC 1127, 1164, 1178 (Dec.13,1977), his was affirmed by the Appeal Board, Virginia Electric and Power Company (North Anna Power Station, Units 1 and 2), ALAB-491, 8 NRC 245 (Aug. 25, 1978).
Appendix D h is Appendix has been superseded by 10CFR Part 51. As noted in the discussion for 10CFR50.40(d), the require-ments of Part 51 have been satisfied.
Appendix E his Appendix specifies requirements for emergency
-
plans. Ati emergency plan was prepared for North Anna Units 1 and 2 prior to the granting of an operating license for Unit 1. This Emergency Plan was litigated before the Atomic Safety and Licensing Board, and the l
'
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.
Virginia Electric and Power Company (North Anna Nuclear Power Station, Units 1 and 2), LBP-77-68, 6 NRC 1127, 1178 (December 13, 1977), affirmed, ALAB-491, 8 NRC 245 (August 25, 1978).
l 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 modi-fled and improved. This revised plan, which meets the criteria in NUREG-0654 and has been accepted by the NRC Staf , has been implemented.
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Regulation (10CFR) Compliance ;
Appendix F This Appendix applies to fuel reprocessing plants and ;
'
related waste management facilities, not to power reactors such as North Anna 2, and it is therefore not applicable to this proceeding.
Appendix G Fracture coughness requirements of this Appendix and program requirements given in Appendix H form the basis for Technical Sepcification surveillance requirements dealing with the use of surveillance specimens. Addi-tional information to demonstrate compliance can be found in FSAR Chapter 5, Section 5.4.3.6, concerning the irradiation surveillance program. Heatup and cooldown limits consistent with the requirements of this Appendix are established in the Technical Specifi-cations.
Appendix H Rea: tor vessel material surveillance progran require-me ts are delineated 11 this part. Technical Specifi-cation 4.4.9.1.2 and operating procedures have been established to implement these requirements with a further requirement to update the "heatup" and " cool-down" curves provided in the Technical Specifications.
Further information is provided in FSAR Chapter 5, Section 5.4.3.6.
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. The compliance of North Anna 1 and 2 with this requirement was litigated in great detail before the Atomic Safety and Licensing Board, which expressly found that there was reasonable assurance that release of radioactive materials in effluents from North Anna Units 1 and 2 to unrestricted
,
areas would be in compliance with applicable NRC regu-
'
lations. Virginia Electric and Power Company (North Anna Nuclear Power Station, Units 1 and 2), LBP-77-68, 6 NRC 1127, 1167, 1178 (Dec. 13, 1977), affirmed, ALAB-491, 8 NRC 245 (Aug. 25,1978).
l l
Appendix J Reactor containment leakage testing for water cooled power reactors is delineated in this Appendix. These requirements are given in Technical Specifications 3/4.6.1.2, 3/4.6.1.3, and 3/4.6.1.6. Additional infor-mation concerning compliance can be found in FSAR t
Chapter 3, Sections 3.1.45, 3.1.46, 3.1.47, 3.8.2.8 and FSAR Chapter 6. Section 6.2.1.4.
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Regulation. !
(10CFR) Compliance It should be noted that a specific exemption to this Appendix has been granted in Technical Specifica-tion 3/4.6.1.3 regarding testing of each containment air lock.
Appendix K This Appendix specifies features of acceptable ECCS evaluation models. As noted above for 50.46, the analysis for North Anna has been conducted using a model which has been accepted by the Commission staff as meeting the requirements of 1.his Appendix.
Appendix L his Appendix covers information requested by the Attorney General for anti-trust review of license applications. As noted above, the anti-trust review for North Anna 1 and 2 took place at the construction permit stage.
Appendix M This Appendix covers standardization of design and it not applicable to the North Anna 2 proceeding.
Appendix N This Appendix covers standardization of nuclear power plant designs and is not applicable to North Anna 2.
Appendix 0 This Appendix covers stw iardization of design and is not applicable to North .-.na 2.
Appendix P This Appendix is proposed, 39 Fed. Reg. 26293, and it applies to fuel reprocessing plants, not power reactors such as North Anna 2. Accordingly, it is inapplicable.
Appendix Q This Appendix governs preapplication early review of site suitability issues and is not applicable to North Anna 2.
Appendix Q This Appendix is only proposed, 39 Fed. Reg. 26297, and (Proposed) it would apply to fuel reprocessing plants, not power reactors such as North Anna 2.
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Regulation .
(10CFR) Compliance )
l 100.1 This regulation is explanatory and does not impose l independent obligations on licensees. )
l 100.2 This regulation is explanatory. North Anna 2 is not l novel in design and is not unproven as a prototype or ,
pilot plant. l l
100.3 This regulation is explanatory and does not impose independent 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, 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 2.12. 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 six (6) miles from the centerline of Unit No. I containment. 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 Charlottesville, Virginia, which is 38 miles west of the site. Frede-ricksburg, Virginia is expected to become the nearest population center during the operating life of the station. The distance to Fredericksburg is 23.5 miles, well in excess of the requirements.
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 North Anna site with this Appendix was thoroughly litigated some years ago in the North Anna geologic fault "Show Cause" proceeding. The Commission determined that the North Anna site meets the NRC seismic and geologic siting criteria, and this decision was affirmed by the United States Court of Appeals for the District of Columbia Circuit. North Anna Envirotunental Coalition v. U. S. Nuclear Regulatory _
Commission, 533 F.2d 655 (D.C. Cir. 1976).
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