ML20002A769
| ML20002A769 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/14/1980 |
| From: | Nichols T SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 10CFR-050.36AA, 10CFR-50.36AA, NUDOCS 8011210503 | |
| Download: ML20002A769 (25) | |
Text
- -~
s.
,s south CAROtlNA ELECTRIC & GAS COMPANY post orrier aos pe4 CotumstA. Soutu CAnouwA reats T. C. thcnots, Jn.
9
-viene.
o rne.m November 14, 1980 m.eru on~,.
1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation -
4 U. S. Nucicar Regulatory Commission-1 Washington, D. C. 20555
Subject:
Vfrgil C. Summer Nuclear Station Docket No. 50/395 Compliance with NRC Regulations
Dear Mr. Denton:
4 As requested in Mr. R. L. Tedesco's letter dated 10/27/80, South Carolina Electric and.Cas Company (SCE6G), acting for itself and agent
'for South Carolina Public Service Authority, wides forty-five (45) copies of.a listing of the. compliance of th'
.rgil C. Summer Nuclear-Station with NRC Regulations 10CFR Parts 2' 50 and 100.
If you h' ave any questions,_please let us know.
N'.
m Q
Very truly yours, i..i, 7 c.?udd, s>.
43
- m. l.$u'
?.
r do-w r
JA j
T. C. Nichols, Jr pi3
- s IE.,"
E.-,6i c
'i RBC:TCN:rh 2
o
- 2 E
S
-m ro
%'t i
1 cc:
V.<C.
Summer G. H. Fischer' in
- T. C. Nichols, Jr.
E. H. Crews, Jr.
O. W. Dixon, Jr.
-DJ A..Nauman
- 0. S. Bradham R._B.
Clary W.'A. Williams, Jr.
i B. A. Bursey
- J. B.-Knotts J. L.'Skolds A.' R. Koon NPCF/Whitaker File 4
-8011210503 e
r,1-y 3
s i.
F f
. ENCLOSURE COMPLIANCE OF VIRGIL C. SUmiER NUCLEAR STATION UNIT I WITH Tile NRC REGULAT20NS OF 10 CFR PARTS 20, 50, AND 100 Regulation (10 CFR)
Compliance 20.1(a)
This regulation merely states che general purpose for which the Part 20 regulations are established and does not impose any inde-pendent 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 t
material by any licensee, such that the total dose to an individual will not exceed the standards prescribed therein.
It does not impose any independent obligations on licensees.
20.1(c)
. Conformance to the ALARA principle stated in tLis regulation is ensured by the implementation of SCE6G policies and appropriate Technical Specifications and health physics procedures.
Chapters 11 and 12 of the FSAR describe the specific aquipment and pesign 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 plant procedures.
20.5 The units of Radioactivity specified in this regulation are accepted and conformed to in all applicable plant precedures.
20.6 This regulation governs the interpretation of regulations by the NRC and does not impose independent obligations on licensees.
20.7 This regulation gives the address of the NRC and does not impose 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.
Conformance is documented by the use of appropriate personnel monitoring devices and the maintenance of all required records.
20.102-When required by this regulation, the accumulated dose for any
. individual permitted to exceed the exposure limits specified in 20.101(a) is determined by the use of Form NRC-4.
Appropriate health physics procedures and administrative policics control this process. 4
,4 %, w 3
h..
-c' J
1
R[gulction (10 CFR)
Compliance 20.103(a)
Compliance with this regulation is ensured through the implementation of appropriate health physics procedures relating to air sampling for radioactive materials, and bioassay of individuals for internal contamination. Administrative policies and controls provide adequate margins of safety for the protection of individuals against intake of radioactive materials. The systems and equipment described in Chapters 11 and-12 of the FSAR provide the capability to minimize these 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 reasonably achievahir When necessary, as. determined by p1snt administrative guidelines, additional precautionary procedures are utilized to limit the potential for intake of radioactive materials.
20.103(c)
The plant Respiratory Protection Procedure implements the require-ments of this regulation by ensuring the proper use of approved respiratory protection equipment. The plant Respiratory Protection Procedure incorporates fully the stipulations of Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection."
20.103(d)
This regulation describes further restrictions which the Commission may impose on licensees. It does not impose any independent obligation.
on licensees.
20.103(e)
The notification specified by-this regulation will be made at least 30 days prior to fuel loading or use of respiratory equipment as stipulated in this section.
20.103(f)
The Respiratory Protection Program is in full comformance with the requirements of 20.103(c).
20.104
.Conformance with this regulation is assured by appropriate SCE&G policies regarding employment of individuals under the age of 18 and the Health Physics procedures restricting these individuals' access to restricted areas.
20.105(a)
Chapter 11 of the FSAR provides the information and related radiation dose assessments specified by this regulation.
20.105(b)
The radiation d(se rate limits specified in this regulation are complied with through the implementation of plant procedures, Technical Specifications, and administrative policies which control the use and transfer of radioactive materials. Appropriate surveys and monitoring devices document this compliance.
20.196(a)
Conformance with the limits specified in this regulation is assured through the implementation of plant procedures and applicable Technical Specifications which provide adequate sampling and analysis, and monitoring of radioactive materials in ef fluents.before and during their release. The level of radioactivity in station effluents is minimized to the extent practicable by the use of appropriate equip-ment designed for this purpose, as described in Chapter 11 of the FSAR.
20.106(b)
'SCE&G'has not included and does not intend to include in any. license 20.106(c) or amendment applications proposed limits higher than those specified in 20.106(a), as provided for in these regulations..,.
R:gulction (10 CFR)
Compliance 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 provides criteria by which the Commission may impose further limitations on releases of radioactive materials made by a licensee.
It imposes no independent obligations on licensees.
20.106(f)
This regulation merely states that the provisions of 20.106 do not apply to disposal of radioactive material into sanitary sewerage systems.
It imposes no independent obligations on licensees.
20.107' This regulation merely clarifies that the Part 20 regulations are not intended to apply to the intentional exposure of patients to radiation for the purpose of medical diagnosis or therapy.
It does not impose any independent obligations on licensees.
20.103 Necessary bioassay equipment and procedures, including Whole Body Counting, are utilized to determine exposure of individuals to concentrations of radioactive materials. 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. When necessary, the Radiation Work Permit system established at the station provides for detailed phy-sical surveys of equipment, structures, and work sites to determine appropriate levels of radiation protection.
The plant administrative and applicable health physics procedures require these surveys and provide for their documentation in such manner as to ensure compliance with the regulations of 10 CFR Part 20.
20.202(a)
The plant administrative procedures and applicable health physics procedures set forth policies and practices which ensure that all individuals are supplied with, and required to use, appropriate personnel monitoring equipment. The Radiation Work Permit system is established to provide additional control of personnel working in radiation areas and to ensure that the level of protection afforded to these individuals is consistent with the radiological hazards in the work place.
20.202(b)
The terminology set forth in this regulation is accepted and conformed to in all applicable procedures and Technical Specifications.
20.203(a)
All materials used for labeling, posting, or otherwise designating radiation hazards or radioactive materials, and using the radiation symbol, conform to the conventional design prescribed in this regulation.
20.303(b)
This regulation is conformed to through the implementation of appro-priate' health physics procedures and portions of the plant adminis-trative procedures relating to posting of radiation areas, as defined in 10 CFR Part 20.202(b)(2)..
R:gulction (10 CFR)
Compliance 20.203(c)
The requirements of this regulation for "High Radiation Areas" are conformed to by the implementation sf the Technical Specifications and appropriate plant health pbveics procedures, as well as the plant administrative procedures. The controls and other protective measures set torth in the regulation are maintained under the surveillance of the Virgil C. Summer Nuclear Station Health Physics Group.
It should be noted.that Technical Specifications provide alternate access control methods to be applied "in lieu of the ' control device' or ' alarm ' signal' required by paragraph 20.203(c)(2) of 10 CFR 20."
which will prevent unauthorized entry into a high-radiation area.
20.203(d)
Each Airborne Radioactivity Area, as defined in this regulation, is required to be posted by provisions of the plant administrative procedures and appropriate health physics procedures. These pro-cedures 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 plant administrative procedures.
20.203(f)
The container labeling requirements set forth in this regulation are complied with through the implementation of appropriate health physics procedares and portions of the plant administrative procedures.
20.204 The posting requirement exceptions described in this regulation are used where appropriate and necessary. Adequate controls are provided within the plant 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 the plant administrative procedures and appropriate health physics procedures. These procedures also provide for the necessary documentation to ensure an auditable record of compliance.
20.206 The requirements of 10 CFR 19.12 referred to by this regulation are satisfied by the station orientation training conducted at the Virgil C. Summer Nuclear Station. Appropriate procedures cet forth require-ments for radiation workers to receive instruction on a periodic basis.
20.207 The storage and. control requirements for licensed materials in unres-tricted areas are conformed to and documented through the implementatio of plant health physics procedures and applicable portions of the plant administrative procedures.
20.301 The general requirements for waste disposal set forth in this regu-lation are complied with through surveillance instructions, the
. Technical Specifications, and the provisions of the station license.
Chapter'11 of the FSAR describes the Solid Waste Disposal System installed at the plant.
-4_
a n
- R2gulation (10 CFR)'
' Compliance 20.302-No'such application for proposed ^ disposal procedures, as described in this regulation, has been:made or is contemplated.by SCE&G.
20.303 No plans for waste disposal by release into sanitary' sewerage systems, as provided:for in this regulation, are contemplated by the Virgil t
C. Summer Nuclear Station.
20.304 Dispo' sal of wastes by burial in soil (i.e., onsite burial), as pro-vided for in this regulation, is not performed or being contemplated.
20.305 Specific' authorization, as described in this regulation, is not l
currently being sought by SCE&G 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 Virgil C. Summer Nuclear Station has established an appropriate inventory and control program to ensure strict accountability for all licensed radioactive materials.
Reports of theft or loss of licensed material are required by reference to the regulations of 10 CFR'in the Technical Specifications.
20.403 Notifications of incidents, as described in this regulation, are assured by the requirements of the Technical Specifications.
Plant i
administrative procedures and appropriate plant procedures, which also provide for the necessary assessments to determine the occurrence of such incidents.
~
20.404:
This regulation was deleted effective September 17, 1973 (38 Fed.
Reg. 22220).
g
-20.405 Reports of overexposures to ' radiation and the occurrence of excessive j
levels and concentrations, as required by this regulation, are pro-vided for by reference in the-Technical Specifications and in appro-priate health physics procedures.
20.406 This regulation.was deleted August 17, 1973, effective September 17, 1973 (38 Fed. Reg. 22220).
20.407 The personnel monitoring report required by this regulation is r
expressly provided.for by the Technical Specifications. Appropriate health physics procedures establish the data base from which this F
treport is generated.
I 20.408 Thd report of radiation exposure required by.this regulation upon-termination of an individual's-amployment or work assignment is j
generated through~the' provisions of plant ~ procedures.
i.
_ l' o.
. t
+
-+<
y e--y e-g
,e
+-eg
+
g 4--
r**--g-----g
--=*N--n-W t'
9-+
y4
Rrgulttion (10 CFR)
Compliance 20.409 The notification and reporting requirements of this regulation, and.those referred to by it, are satisfied by the provisions of plant 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.
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 violations of its rules.
It does not impose any independent obligations on licensees.
Appendix A
-(Reserved) - Not used by the NRC.
Appendix B This appendix only provides information and does not impose indenen-dent obligations on licenstas.
Appendix C This appendix only provides information and does not impose indepen-dent obligations on licensees.
Appendix D This appendix only provides information and does not impose indepen-dent obligations on licensees.
c 1 '
\\
R6gul; tion (10 CFR)
Compliance 50.1 This regulation otates 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 inde-pendent 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 independent obligations on licensees.
impose 50.10 These regulations specify the types of activities that may not be 50.11 undertaken without a license from the NRC.
to conduct any such activities without an NRC license.SCE&G does not propos 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.
impose independent obligations on licensees.
It does not 50.13 This regulation says that a license applicant need not design against acts of war.
It imposes no independent obligations on licensees 50.20
.These regulations merely describe the types of licences that the NRC 50.21 issues.
They do not address the substantive requirements that an 50.22 applicant must satisfy to qualify for such licenses.
50.24 This regulation haa beca deleted,, 35 Fed. Reg. 19655.
50.30 This regulation sets down procedural requirements for the filing of lice.nse applications, such as the number of copies of the application that must be provided the NRC.
with the procedural requirements in effect atSCESC has substantially complie p
its license application and the amendments to it.the time when filing 10 CFR 50.30(f) requires ' Eat a license application must be accom-In particular, panied by any Environmental Report required pursuant to 10 CFR Part 51, and SCE&C has submitted u Final Environmental Report covering the Virgil C. Summer Nuclear Station.
50.31 These regulations merely permit more efficienc organization of the 50.32 license application and impose no independent obligations on licensees 50.33 This regulatica requires the license application to contain certain
~~
general. Information, such as an identification of the applicant, information about the applicant's financial qualifications, and a list of regulatory agencies with jurisdiction over the applicant's rates and services.
Nuclear Station operating license application.This information was 50.33(a)
This regulation requires applicants for construction permits to submit information required for antitrust review.
The antitrust review re-quired by the Atomic Energy Act of 1954, as amended, was pt rformed the construction permit stage.
<d at fby
_7,.
( men g ::a ds"-
W
- R2gulacion (10 CFR)
Compliance 50.34 (a)
_ This regulation governs the contents of the ?reliminary 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 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 1 (4)
Jesign and performance evaluation - ECCS performance is discussed and shown to meet the requirements of 10 CFR 50.46 in Chapters 6 and 15.
(5) results of research programs - There were no research and deve-lopment programs required to demonstrate that a safety question identified at the construction permit stage has been resolved.
(6)
(i) organizational structure - Chapter 13 (ii) managerial and administrative' controls - Chapters 13 and 17.
Chapter 17 discusses compliance with the quality assurance requirements of Appendix B.
(iii) plans for pre-operational 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 Specifications.
(v) plans for coping with emergencies - Emergency Plan (Chapter 13)
(vi)
Technical Specifications - prepared in conjunction with the Staff (Chapter 16 - withdrawn from FSAR; to be issued by NRC as part of operating license)
(vii) not applicabic, since the op(rating license application was filed before February 5, 1979 (7) technical qualifications - Chapter 13-(8) operator requalification program - Chapter 13 50.34(c)
A physical security plan has been prepared and is implemented for the Virgil C. Summer Nuclear Station.
50.34(d)
An emergency response plan for SCE&G has been prepared and is imple-mented for the Virgil C. Summer Nuclear Station.
_8_
e.
.c
,em. we a
ww-
.m.m.-
P
-en-
~.
i 1R gulstion
-(10 CFR)~
Compliance
+
50.35 This' regulation is-relevan't to the construction permit stage rather than the operating stage.
'I 50.36
' Technical Specifications have been prepared incorporating 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) adminis-trative controls.
50.36(a)
.The Environmental Technical Specifications, Part 1, include specifi-
- cations which require compliance with 10 CFR 50.34(a) (releases as low as is reasonably achievable).and that ensure that concentrations 1
of radioactive effluents released to unrestricted areas are within the limits specified in 10 CFR 20.106. The' reporting requirements of 10 CFR 50.36(a)(2) are also included in these specifications.
50.3'7
~ This regulation requires the applicant to agree to limit access to I
Restricted Data. SCE6G's agreement to do so is the operating license 4
application for the Virgil C. Summer Nuclear Station.
50.38 This regulation prohibits the NRC frou issuing a license to foreign-l' controlled entities.
SCE6G's statement that it is not owned, controlled or. dominated by an alien, a foreign corporation, or a foreign govern-ment is in the operating license application for the Virgil C. Summer
.Nubicar Station.
4 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 considsratio'ns to " guide":the Commission in granting licenses as follows:
s i
The design and operation of the facility is to provide reasonable
'50.40(a) assurance that the applicant will comply with NRC regulations, including those in 10 CFR Part 20, and that the health and safety of the public will not be endangered. The basis for SCE&G'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 reviews by SCE&G's own staff, the NRC staff, the ACRS, and NRC licensing boards, provides a great deal of assurance that the public h"+1th' and safety will not be endangered.
50.40(b)
Another consideration is that the.a'pplicant-be technically and finan-cially qualified. Both SCE&G's technical qualifications and.its financial qualifications have been submitted and will be reviewed in hearings.before the Atomic Safety and -Licensing Board.
50.40(c)
Another' consideration is that the issuance of the license is not to be
~ inimical to the common defense and security or to the health and safety of the public. The individual. showings of compliance with particular
. regulations contained in this enclosure, as well as the _ contents of the entire FSAR and related correspondence over the years, plus the.
. lengthy' process of design, construction, and review by SCE&G, its NSSS 3
~
Regulation (10'CFR)
Compliance vendor, and the government, provide SCE&G 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 consider-able assurance that the license will not be inimical in that SCE&G has a security plan for the Virgil C. Summer Nuclear Station, that SCE&G
. is not controlled by agents of foreign countries, and that SCE&G has agreed to limit access to. Restricted Data (see above).
50.40(d)
The' final 50.40 " consideration" is that.the applicable requirements of Part.51 have been satisfied. Part 51 concerns compliance with the National Environmental Policy Act of 1969. SCE&G has submitted a Final Environmental Report pursuant to 10 CFR 50, Appendix D.
Envi-ronmental Technical Specifications are being generated for the Virgil C. Summer Nuclear Station.
50.41 This regulation applies to class 104 licensees, such as those for devices used in medical therapy.
SCE&G 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 Commiselon in issuing Class 103 licenses.
The two considerations.are (a)~that the proposed activities will serve a useful purpose propor-tionate 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 "us'eful 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 r.ced for power are constantly changing, SCE&G periodically makes load projections, 'and in SCE&G'c judgment, the need for.the Virgil C. Summer Nuclear Station is.still substantial.
As for the amodnt of special nuclear material or source material used, there is no reason to believe that their proportion in relation to the power produced is substantially greater than that of other commercial power reactors in this country.. As for the antitrust advice of the Attorney General, as noted above, the antitrust review was done at the construction permit stage.
50.43 This regulation imposes certain duties on the NRC and addresses the applicability of the Federal Power Act and the right of government agencies to obtain NRC licenses.
It imposes no direct obligations on licensees.
50.44 The Virgil C. !humner Nuclear Station combustible gas control system is described in FSAR Section 6.2.5.
The system is designed to maintain the hydrogen concentration in ccntainment at a safe level following a LOCA, without purging the containment atmosphere, as specified in 10 CFR SG.44(e). The system consists of internal recombiners and two hydrogen analyzers. -This system meets the requirements of NUREG-0660 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 therefore not material to this operatint license proceeding. -
m,,a y.%.*,.~4
.w
-<o-e-
_p
R:gulction (10 CFR)
Compliance 50.4L FSAR Sections 6.3 and 15.4.1 describe the Emergency Core Cooling System and the methods used to analyze ECCS performance following a postulated loss-of-coolant accident.
50.47 FSAR Chapter 13 and the Emergency Response Plan for the Virgil C.
Summer Nuclear Station provide information regarding emergency plans.
50.50 This regulation provides that the NRC will issue a license upon deter-mining 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 ao 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 to licensec 50.53 This regulation provides that licenses are not to be issued for acti-vities that are not u'nder or within the jurisdiction of the United States. The operation of the Virgil C. Sumner Nuclear Station will be within the United States and subject to the jurisdiction of the United States, as is evident from the description of the facility in the operating license. application.
50.54 This regulation specifies certain conditions that are incorporated in every license issued. Compliance is effected simply by including thest 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 permics, not operating licenses.
It will not be relevent when the operating licenst is issued. During the construction phase, SCE&G procedures nddress the conditions of this regulation.
- 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 providts information concerning the Qualit Assurance Program'that was utilized. As a further example of a speci-fic 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 approved Edition and 50.55a(b)(2)
Addenda of Section III and XI of the ASME Code.
50.55a(c)
Design and fabrication of the reactor vessel was carried out in accor-dance with ASME Section III, 1971 Edition.
Information can be found in Chapter 5.
- e
~.
Reguiction (10 CFR)
Compliance 50.55a(d)
Reactor coolant system piping meets the requirements of ASME Section III, 1971 Edition through the Winter 1971 Addendum.
Information can be found in Chapter 5.
50.55a(e)
Reactor coolant pumps meet the requirements of ASME Section III, 1971 Edition through the Summer 1972 Addendum.
Information can be found in Cha'pter 5.
50.55a(f)
Noting the construction permit date'of March 1973, the valves within the reactor coolant system pressure boundary were designed and fab-ricated in accordance with the requirements of ASME Section III, 1971 Edition through the Summer 1971 Addendum, except the pressurizer safety valves, which meet the requirements of ASME Section III, 1971 Edition through the Winter 1972 Addendum.
50.55a(g)
Inservice Inspection (ISI) requirements are delineated in this part and are specified in the Technical Specifications, paragraph 4.0.5.
As permitted by this part and the Technical Specifications, certain exemptions are requestcd'and granted for the inservice inspection of various systems and the inservice testing of various pumps and valves.
By letters dated 4/11/80, 5/16/80,.and 9/17/80, the Virgil C. Summer Nuclear Station Pre-service Inspection Program was docketed.
50.55a(h)
As discussed in FSAR Chapter 7, the protection systems meet IEEE 279-1971.
50.55a(i)
Fracture toughness requirements are set forth in Appendices G and R of 10 CFR 50.
Technical Specifications require the use of reactor vessel material irradiation surveillance specimens and updating of the "heatup" and "cooldown" curves given in the Technical Specifications.
Further information is given in FSAR Section 5.4.3.6 concerning the irradiation surveillance program.
50.55b This regulation has been revoked. 43 Fed. Reg. 49775.
50.55e This regulation is only proposed, 39 Fed. Reg. 26297, 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 operating license upon completion of the construction of a facility in compliance with the terms and conditions of the construction permit. This imposes no independent obligations on the applicant.
50.57(a)
This regulation requires the Commission to make certain findings before the issuance of an operating license. These findings can be made for full power operation for the reasons given in this enclosure generally.
Specifically:
,1).
Construction of the facility has been substantially completed in conformity with the construction permit and the application as amended. Co'nformance of the facility to the NRC rules and regula-tions and the Act, as implemented by the regulations, has been demonstrated by the application. -
R:guletion (10 CFR)
Compliance (2) The Technical Specifications and resulting operating procedures provide assurance that the unit 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 public and in compliance with the regulations, as noted above.
(4) The application demonstr' st SCE6G is technically and finan-cially qualified to ope unit.
(5) The applicable provisiot
'FR 140 have been satisfied.
(6) The Security Plan assure, ecial nuclear material is being appropriately safeguarded application demonstrates that the operation of the unit will not be inimical to the health and safety of the public.
50.57(b)
The license, as issued, will contain appropriate conditions to assure that items of construction or modification are completed on a schedule acceptable to the Commission.
50.57(c)
This regulation provides for a low-power testing license.
50.58 This regulation provides for the review and report of the Advisory Committee on Reactar Safeguards.
- The 'ACRS :will review the operating license application in accordance with its usual practice.
50.59 This regulation provides for the licensing of cer.tain changes, tests, and experiments at a licensed facility. Technical Specifications and procedures provide implementation of this regulation.
50.60 This regulation has been deleted, 40 Fed. Reg. 8790.
50.65 This regulation has been deleted, 43 Fed. Reg. 6915, 50.70 The Commission has assigned resident inspectors to the Virgil C.
Summer Nuclear Station.
SCE&G has provided office space in accordance with the requirements of this section. SCE&G permits access to the station to NRC inspectors in accordance with 10 CFR 50.70(b)(3).
50.71 Records are and will he maintained in accordarce with the requirements of sections (a) through (e) of this regulailon nr.d the license.
Sec-tion (c) requires that the FSAR be updated by July 22, 1982, and annually thereafter.
Such updates will be made.
Notification of significant even s to.the NRC will be made in accor-50.72 e
dance with the requirements in tnis regulation.
50.80 This regulation provides that licenses may not be transferred without NRC consent. No application for transfer of a license is involved in the Virgil C. Summer Nuclear Station's proceeding. -.
'Regulcti n
_(10 CFR)
Compliance 50.81 This regul.ation permits the creation of mortgages, pledges, and liens on licensed facilities, subject to certain provisions. SCE6G licensed j
facilities are not mortgaged, pledged, or otherwise encumbered as those terms are used in this regulation.
50.82 This regulation provides for the termination of licenses.
It does not apply to the Virgil C. Summer Nuclear Station because SCE6G has not requested the termination of a license.
1 50.90 This regulation governs applications for amendments to licenses.
Future requests for license amendments will be made in accordance with these requirements.
I 50.91 This regulation provides guidance to the NRC in issuing license amendments.
50.100 These regulations govern the revocation., suspension, and modification 50.101 of licenses by the Commission under unusual circumstances.
No such 50.102 circumstances are present in the Virgil C. Summer Nuclear Station pro-50.103 ceeding, and these regulations are not applicable.
50.109 This regulation specifies the conditions under which t'ae NRC may re -
quire the backfitting of a facility. This regulation imposes no inde-pendent obligations on a licensee unless the NRC propose.s a backfitting requirement, and so this regulation is not applicable.
50.110 This regulation; governs enforcement of the Atomic Energy Act, the Energy Reorganization Act of 1974, and the NRC's regulations and orders.
No enforcement action is at issue.in.the Virgil C. Summer Nuclear Station proceeding, and so.this.. regulation is nqt applicable.
Appendix A CDC 1 Section 3.1.2.1 of the FSAR describes the design provisions made to ensure that these requirements are met.
Codes and standards utilized for the unit are specified throughout the FSAR.
Chapter 17 describes the quality assurance program and the provisions for maint snance of records.
GDC 2 FSAR Section 3.1.2.1 addresses the design considerations for natural phenomena, which are described in detail in Chapters 2 and 3.
Appro-priate 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 tha safety func-tions to be performed.
GDC 3 FSAR Section 3.1.2.1 describes in general the measurcs 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.
Related material is found in Sections 7.6, 8.3, and 13.0.
GDC 4 FSAR Section 3.1.2.1 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...
m n w e r.~ m & -m:c ~
w
4
~.
ET i Regulstioni (10 CFR)
Compliance
. Chapter 3 provides information concerning the specific design features for protection against missiles, jet impingement, and pipe rupture. Pro-visions for qualification of equipment for all' postulated' environments is described in several sections of the FSAR. A NUREG-0588 review is
'in~ progress to confirm that electrical equipment is adequately demon-strated to be qualified for its expected service environments.
~GDC 5'
.Since.the Virgil C. Summer Nuclear Station is a or,e-unit plant, this criterion ~does not apply.
GDC 10 FSAR Section 3.1.2.2 indicates that the reactor core and associated systems.are designed to. function throughout the design lifetime with-out exceeding. fuel damage limits, using protection criteria specified
.in Section 3.1.2.2 and Chapters 4, 7, and 15.
3 GDC 11 FSAR Section 3.1.2.2 indicates that prompt compensatory reactivity feedback effects are assured by unit design and operational limit i '
considerations. The core inherent reactivity feedback characteristics
.and reactivity control methods are described in FSAR Section 4.3.
GDC 12 FSAR Section 3.1.2.2 describes the inherent and design features which eliminate or limit the various types of oscillations. Core stability is further described in Section 4.3.
GDC 13' As indicated in FSAR.Section 3.1.2.2, and described in more detail in
.Che ter 7,. instrumentation and. control systems have been provided to itor and maintain plant variables including those variables which n.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.
CDC 14 FSAR Sectien 3.1.2.2 indicates that the reactor coolant pressure i
boundary has been designed to accomnodate the. system temperatures and pressures attained under all expected operational modes and anticipated transients, and to maintain stresses within applicable limits.
CDC 15 As indic'ated in FSAR Section 3.1.2.2,.the. reactor coolant system and associated auxiliary, control, and protection systems are designed
-to ensure the' integrity of the reactor coolant pressure boundary with adequate margins during normal operations and anticipated transients.
The design codes used for the Reactor Coolant. System are described in Chapter 5.
Details concerning the protection systems are provided in-Chapter 7.
GDC 16-As' described in FSAR Sections 3.1.2.2 and 3.8'and Chapter 6, a steel-lined concrete 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 containment and its asso-ciated. engineered safety features thus meet the required functional capability of this criteria.
CDC 17 As described in FSAR Section 3.1.2.2, onsite and offsite power systems
. arc provided'which can independently supply the electric power requiret i
for the operation of safety-related systems. This capability is
Regulation (10 CFR)
Compliance maintained even with the failure of any single active component in
-either system. Chapter 8 provides the design details of the power systems and their compliance with this criterion.
GDC 18 As described in FSAR Section 3.1.2.2 and Chapter 8, the redundant electric power systems important to safety are continuously moni-tored 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 Specifications.
GDC 19 FS AR Section 3.1.2.2 describes the main control room, which contains the controls and instrumentation necessary for safe operation of the unit during normal and accident conditions.
Sufficient shielding, distance, structural integrity, and ventilation r.ystems 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, a local control-room evacuation panel is provided, within the protected envelope, which may be uscu to bring the reactor to cold shutdown.
CDC 20 FSAR Section s.l.243 discusses the desig.. criteria for the protection system and engineered safety features actuation, to ensure that the requirements of this criterion are met.
Further details are supplied
.in Chapter 7.
GDC 21 As indicated in FSAR Section 3.1.2.3, the protection system is designed for the high functional reliability and inservice testability commensurate with the safety functions to be performed..This section, as well.as Chapter 7, describe in detail the design features provided to ensure redundancy and testability.
GDC 22 FSAR Section 3.1.2.3 indicates that the protection system has been designed to provide sufficient resistance to a broad class of acci-dent conditions or postulated events. Chapter 7 provides further design details concerning this resistance such that independence is maintained.
GDC 23 As indicated in FSAR Section 3.1.2.3, the protection system is designed with due consideration of the most probable failure modes of the com-ponents under various perturbations of energy sources and the environment.
Further details are supplied in Chapter 7.
GDC 24 FSAR Section 3.1.2.3 discusses separation of the. protection and control systems, such that the failure of any signal control system component or channel or the failure or removal from service of any protection system component cr channel which is common to the pro-tection and control systems leaves intact 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. -
' R2gulctio'n (10 CFR)
Compliance GDC 25 FSAR Section'3.1.2.3 indicates that the protection system has been.
designed to assure that specified acceptable fuel-design limits are not exceeded in the' event of any single reactivity control system malfunction, including ~an accidental withdrawal of control cluster groups. Further. details are provided in'FSAR Sections 4.3.1.4,
-7.2.2.,2.3,.and 7.7.2.2.
GDC 26 As indicated.in FSAR Section 3.1.2.3, two independent reactivity control systems of different design principics.are provided. One of
-the systems-uses control rods; the second system employs dissolved boron as'a chemical shim. Reactivity control system redundancy and capability are described further in. Sections 4.3.1.5'and 7.7.2.2.
.GDC 27 As described in FSAR Section 3.1.2.3, means are provided for shutdown reactivity for cooling the core under any anticipated condition and with appropriate margin for contingencies.
Combined use of rod 7
cluster control and chemical shim control permit the necessary shutdown i:
margin to be maintained.during the long term xenon decay ~and plant.
~cooldown. These means are discussed in detail in FSAR Sections 4,3 and 7.2.
CDC 28 FSAR Section 3.1.2.3. indicates that core reactivity. is controlled '
by a chemical' poison dissolved in the coolant, rod cluster assemblies,.
and-burnable poison rods. The maximum reactivity insertion rates due to withdrawal of a-bank or rod cluster control assemblies or by boron dilution are limited. The' maximum worth of control rods and the maximum rates of reactivity insertion employing control rods are
~
' limited to values which prevent rupture of the coolant pressure boundary lor disruption of the core intervals to a degree which would impair. core cooling capacity., Further details are providedlin Section 4.3.
GDC 29 As indicated in FSAR Section 3.1.2.3, the. protection and reactivity control systems are. designed to assure extremely high probability of t
performing their required safety functions in the event of antici-pated 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.
GDC 30 lb described in FSAR Section 3.1.2.4, reactor coolant pressure boundary
- components are designed, fabricated,Jinspected, and tested in'confor-mance with ASME Nuclear' Power Plant Components Code Section III.
}bjor components are classified as seismic Class 1 and are accorded
'the-quality measures appropriate to'this. classification, The evalua-tions of reactor coolant pressure boundary components are discussed
.in Section 5.2.
CDC.31 As' indicated in FSAR Section 3.1.2.4, close control is maintained over mat'erial selection-and fabrication for the reactor coolant 1 system to' ensure.that the boundary behaves in a nonbrittle manner
~
~The materials testing is consistent with 10 CFR 50, Appendices G ud 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 Speci-fications ensure' operation within the pressure-temperature limit McMa Chi: critari....
s
R:gulction (10 CFR)
Compliance GDC 32 FSAR Section 3.1.2.4 describes how the design of the reactor vessel and its arrangement in the system provide the capability for accessi-bility during service life to the entire internal surfaces of the vessel and certain external zones of the vessel.
The reactor arrange-ment 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 3.1.2.4, the chemical and volume control system provides a means of reactor coolant makeup and adjustment of the boric acid concentration. A high degree of functional reliability and safe response to probable modes of failure is assured by provision of standby components.
Details of system design are included in Chapters 6, 8, and 9.
GDC 34 FSAR Section 3.1.2.4 indicates that the residual heat removal system, in conjunction with the steam and power conversion system, is designed to transfer the fission product decay heat and other residual heat
.from the reactor core within acceptable limits. Suitable redundancy is accomplished below 3500F with the two residual heat removal pumps with means available for draining and monitoring of leakage, two heat exchangers, and the associated piping and cabling. The residual heat removal system is able to operate on either onsite or offsite electrical power. Sultable redundancy above 3500F is provided by the steam generators, emergency feed pumps, and attendant piping. De-tails of the residual heat removal system design are in FSAR Section 5.5.7.
CDC 35 FSAR Sections 3.1.2.4 and 6.3 describe the use of passive accumulators with three high-head centrifugal charging pumps and two low-head resi-dual heat removal pumps to provide redundance for failure of any com-ponent in any system. The primary function of the emergency 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 sub-stantially intact and in place, with its essential heat transfer geometry preserved.
GDC 36 As deecribed in FSAR.Section 3.1.2.45 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. Nondestructive
' inspection is performed per ASME Code Section XI.
Technical Specifi-cations require ~ inservice inspection in accordance with applicable ASME codes. Details of the inspection programs are provided in Chapters 5 and 6.
CDC 37 FSAR Section 3.1.2.4 indicates that the components of the emergency core cooling system located outside the containment will be accessible for leaktightness inspection during appropriate periodic tests. Each active component of the system may be individually actuated.on the
_io_
RQgulation (10 CFR)
Compliance normal power source at any time during plant operation to demonstrate operability. The centrifugal charging pumps are part of the charging system, and this system is in continuous operation during plant operations. Actuation circuits are tested, and remote-operated valves are exercised periodically. The testing is described in detail in FSAR Sections 6.3.4 and 7.3.2.2.5 and per Technical Specification surveillance requirements.
GDC 38 As indicated in FSAR Section 3.1.2.4, the reactor building spray system and the reactor building cooling system are provided to remove heat from the containment following a losc-of-coolant 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 design 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 3.1.2.4, the reactor building spray system and the reactor building cooling system are designed such that components can be readily inspected to demonstrate system readiness.
System design details are given in Section 6.2.
GDC 40 As described in FSAR~Section 3.1.2.4, the reactor building heat removal systems described in Section 6.2 are. designed to permit periodic testing no 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. All active components of the reactor building spray system, the. reactor building cooling units, and the related portions of the service water system are tested in place af te:
Installation.
GDC 41 As indicated by FSAR'Section 3.1.2.4, atmospheric cleanup in the reactor building is provided when (a) the reactor building spray
~
systen sprays sodium hydroxide (NaOH) solution into the reactor building to remove iodine; (b) the post-accident hydrogen recombines to remove hydrogen from the reactor building after an accident; and (c) the recirculation system HEPA filter removes particlates from the reactor building. Detailed discussions of these systems are found in Sections 6.2, 6.5, and 15.4.
GDC 42 FS AR Section 3.1.2.4 indicates that the reactor building. spray system, the post-accident hydrogen removal system, and the reactor building cooling unit HEPA filters are designed to. permit appropriate periodic inspection of the important components. Additional diecussion is provided in FSAR Sections 9.4 and 6.2.
GDC 43 FSAR Section 3.1.2.4 indicates that the reactor building spray system, the post-accident hydrogen removal system, and the reactor building cooling unit HEPA filters are designed to pernf,t periodic pressure-testing and functional testing of their cor'orents. Further details are provided in Sections 6.2 and 9.4.,
L
' Regulation 1
'(10.CFR)
Compliance
' GDC,44'
~FSAR'Section 3.1.2.4 describes how a Seismic Category I component
~
1 Cooling Water : System (CCWS) (Section 9.2) is provided to transfer heat from.the Reactor Coolant System,_ reactor support equipment, and
' engineered safety equipment to a Seismic Category I Service Water
- Cooling System '(SWS) (Section 9.2). :The CCWS serves as an inter-rediate system and thus a barrier between potentially or normally radioactive fluids and the SWS. ThelCCWS consists of two independent
.engineeredJsafety subsystems, each of which is capable of serving 4
all necessary loads under normal or accident conditions.
In addition to serving as'the heat sink for the CCWS, the SWS is also used as heat sink-for thc containment.and engineered safety ' equipment'through
.use'of compartment and space coolers.--The SWS consists of two:inde-4 pendent;loopsy each of which is capable of providing all necessary-sheat~ sink. requirements. The SWS transfers heat to the ultimate heat
~~
sink (Section 9.2).
Electric power is discussed.in Chapter 8.
-As indicated in FSAR Section 3.1.2.4, the important components of-the component cooling water system (Section 9.2) and service water system
~
(Section 9.2) are accessible for required periodic inspections.
CDC 46' As described in FSAR Section 3.1.2.4, the cooling water-systems are provided with redundancy and isolation to allow periodic pressure and functional. testing.
s For _ details, see the write-ups on Electric. Power (Chapter 8), Com- -
ponent Cooling Water -System (Section 9.2),. Service Water System (Section.9.2),Jand Instrumentation and Controls (Chapter 7).
CDC.5'0 FSAR Section 3.1.2.5 indicates that the containment structu're, including accese' openings and penetrations, is designed with suffi-
'cient. conservatism to accommodate, without exceeding the design
. leakage; rate', the transient peak pressure and temperature associated
-with a postulated main steam line break. Containment design basis.
is discussed'further in Sections 3.8 and 6.2.
CDC-51 As discussed in FSAR Section 3.1.2.5, the reactor building liner has-a maximum nil ductility transition temperature of at least 300F below-l minimum service temperature in accordance with the ASME Code. -This
. test measures'the ductile-to-brittle transition-with allowable values for energy absorption.
It. insures that the material used will not
' behave in a brittle manner and that' rapidly pro'pagating fracture is-minimized. The containment boundary design considered uncertainties in material-properties; residual,-steady-state, and transient' stresses;.
and material flaws along with conservation allowable stress levels for all Lstressed elements of the containment boundary.. All material 1 was examined for flaws that would--adversely affect the performance of the' material in its-intended purpose. See Section 6.2, Contain--
ment Functional. Design, for'further details. All ferritic materials with thicknesses greater.than. 5/g" in containment penetration process pipe and' system piping considered part of the containment pressure boundary were impact tested.
Refer to FSAR Section110.3. Additional
~
impact testing was1 performed on other penetration assembly components
'with ferritic material thicknesses' greater than 54 1<
_ - er 7
f
i s.
Reguletion (10 CFR)
Compliance CDC 52 As indicated in FSAR Section 3.1'.2.5, the containment design permits periodic integrated leakage rate testir.g during the plant lifetime, in accordance to Appendix J, 10 CFR 50.
Details concerning the conduct of. periodic integrated leakage rate tests are in Section 6.2.
CDC 53 FSAR Section 3.1.2.5 discusses the program for performing individual leak rate tests on applicable penetrations in accordance with Appendix J to 10 CFR 50.
Section 6.2.6 provides further information on this subject. The inservice tendon surveillanca' program is discussed in the Technical Specifications.
GDC 54 As described in FSAR Section 3.1.2.5, the containment isolation fea.ures are provided in all piping systems penetrating containment.
The containment isolation design provides for a double barrier at the containment penetration in those fluid systems that are not required to function following a design basis event. All piping systems penetrating the containment have been provided with test connections to allow periodic leak testing as required.
Sections 6.2.1, 6.2.4, and 6.2.6 provide detailed information on this subject.
GDC 55 As indicated in FSAR Section 3.1.2.5, the reactor coolant pressure bcundary is defined in accordance with Section 4 of ANS-N18.2.
The entire reactor coolant pressure boundary, as defined above, is located entirely within the containment structure.
GDC 56 As. indicated in FSAR Section 3.1.2.5,- each line that connects directly to the reactor building atmosphere and senet rates containment is pro-vided with containment isolation valves, except where,it can be demonstrated that containment isolation provisions for a specific class of lines are acceptable. Further details are provided in Sections 5.5 and 6.2 and Chapter 9.
CDC 57 FS AR Section 3.1.2.5 indicates that those lines that penetrate the containment do not communicate with either the reactor coolant pressure boundary or the containment atmosphere, have at least one isolation valve located outside containment near the penetration, or is a closed system outside containment. This system is further discussed in Chapters 6 and 9.
GDC 60 As described in FSAR Section 3.1.2.6, liquid, gaseous, and solid radioactive waste processing equipment is provided. The prfnciples of filtration, demineralization, evaporation, solidification, and storage for decay and process monitoring to control this cquipment and regu-late releases to the environment are described in Chapters 6,11, and 12 GDC 61 FSAR Section 3.1.2.6 indicates that systems which may contain radio-
~
activity are designed to ensure adequate safety under normal and
-postulated accident conditions.
Components are designed and located such that appropriate periodic inspection and testing may be per-formed. All areas of the plant are designed with suitable shielding for radiation protection based on anticipated radiation dose rates and occupancy as discussed in Section 12.1.
Individual coeponents 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 J
s
Regulation-(10 CFR)
Compliance from the fuel stored in the spent fuel pool. The syst-m is designed for testability to permit continued heat removal. The spent fuel pool is designed such that no postulated accident could cause exces-sive loss of coolant inventory.
Radioactive waste treatment systems are located in the auxiliary building, which contains or confines leakage under nornal and accident conditions. The fuel handling building ventilation system includes charcoal filtration which minimizes radioactive material release associated with a postulated spent fuel handling accident.
Fuel storage and handling is discussed in Section 9.1, and radioactive waste management in Chapter 11.
CDC 62 As noted in Section 3.1.2.6, the restraints and interlocks provided for safe handling and storage of new or spent fuel are dis:ussed in Section 9.1.
The center-to-center distance between the adjacent spent fuel assemblies is sufficient to ensure suberiticality, even if unborated water is used to fill the spent fuel storage pool.
See Section 4.3 for further details. The design of the spent fuel storage rack assembly is such that it is impossible to insert the spent fuel assemblies in other than prescribed locations, thereby preventing any possibility of accidental criticality.
GDC 63-FSAR Section 3.1.2.6 and Chapters 9, 11, and 12 describe the moni-toring capability in the fuel storage and waste handling areas and indicate that the operator will take appropriate actions if an alarm from any of these monitors is received.
>FSAR Section 3il.2.6 indicates the facility contains means for monitoring the containment atmosphere and all other important areas during_both normal and accident conditions to detect and measure radioactivity which could be released under any conditions. The monitoring system includes area ga=ma monitors, atmospheric monitors, and liquid monitors with full indication in the control room. Alarms are provided to warn of high activity. Sections 11.4 and 12.2.4 discuss the process and effluent and area radiological monitoring systems. Section 12.3 describes the associated health physics program.
Appendix B Chapter 17 of the FSAR describes in detail the provisions of the quality assurance program which has been implemented to meet all applicable requirements of Appendix B.
Appendix C This Appendix provides a guide for establishing the applicant's financial qualification. SCE6G's financial qualifications are pre-sented in the application for license. Reasonable assurance is provided that SCE&G has the funds that it needs to operate the
-facility in compliance with the Commission's regulations.
Appendix D This Appendix has been superseded by 10 CFR Part 51.
As noted in the discussion for 10 CFR 50.40(d), the requirements of Part 51 have been satisfied.
Appendix E This Appendix specifies requirements for emergency plans. An emer-gency plan was prepared and has been submitted to the NRC staff for review.and approval.
It provides reasonable assurance that appro-priate measures can and will be taken in the event of an emergency to protect public health and safety and prevent damage to property.
This plan addressed the criteria in NUREG-0654. -
Ragulation (10 CFR)
Compliance 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 C SCE&G meets the requirements of this Appendix as described in Chapter 5.
Appendix 11 Technical Specifications and operating procedures have been estab-lished to implement these requirements.
Further information is provided in FSAR Chapter 5.
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. Conformance to' Appendix I design criteria was established pursuant to applicable Commission criteria, in the Virgil C. Summer Operating License Environmental Report.
Appendix J SCE6G meets the requirements of this Appendix, with exceptions noted and described in FSAR Chapter 6 and associated Technical Specifications Appendix K This Appendix specifies features of acceptable ECCS evaluation models.
As noted above for 50.46, the analysis for the Virgil C. Summer Nuclear Station has been conducted using a model which has been accepted by the Commission staff as meeting the requirements of this Appendix.
Appendix L This Appendix covers information requested by the Attorney Genert1 for antitrust review of license applications. An antitrust revicw
.took place at the construction permit stage, and a review is cur-rently in progress for the operations phase.
Appendix M This Appendix covers standardization of design and is not applicable to the Virgil C. Summer Nuclear Station.
Appendix N This Appendix covers standardization of nuclear power plant designs and is not applicable to the Virgil C. Summer Nuclear Station.
Appendix 0 This Appendix covers standardization of design and is not applicable to the Virgil C. Summer Nuclear Station.
Appendix'P This Appendix applies to fuel reprocessing plants. Accordingly, it is not applicable to the Virgil C. Summer Nuclear Station.
Appendix Q
.This Appendir governs pre-application early - review of site suitability issues and is not applicable to the Virgil C. Summer Nuclear Station.
R;gul: tion (10 CFR)
Compliance 100.1 This regulation is explanatory and does not impose independent obligations on licensees.
100.2 This regulation'is explanatory. The Virgil C. Summer Nuclear Station is not novel in design and is not unproven as a prototype or pilot plant.
100.3 This regulation is explanatory and does not impose 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 2.1.
The low population zone required by 100,11(a)(2) has been established, as described in FSAR Section 2.1.
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 Virgil C. Summer Nuclear Station is discussed in FSAR Chapter 2.
da e