ML20003G972
| ML20003G972 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 04/29/1981 |
| From: | Aswell D LOUISIANA POWER & LIGHT CO. |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| W3P81-1169, NUDOCS 8105040365 | |
| Download: ML20003G972 (63) | |
Text
I y t -
LOUISIANA ,42 eE u ncNoE srn m P O W E R & Li G H T[ p o sox sace . NEW CALEANS, IflUISIANA. (504) 70174366-2345 y70T:ES SYSTEM D. L ASWELL April 29, 1981 va*== ~ ***a=
W3P91-ll69 e 3-A1.1 M Mr. R. L. Tedesco , g Assistant Director for Licensing Division of Licensing g 'g' y
U.S. Nuclear Regulatory Coumission Washington, D.C. 20555
,[ h .
gi
- 6
SUBJECT:
WATERNRD STEAM ELECTRIC STATION - UNIT 3 NV %
COMPLIANCE WITH TITLE 10, CODE OF FEDERAL REGULATION
Dear Mr. Tedesco:
Enclosed is an itemized review of the compliance of the Waterford Steam Electric Station (SES) - Unit 3 with particularly significant rules and regulations of Title 10 of the Code of Federal Regulations. Enclosure 8 of the January 8,1981 Memorandum from William J. Dircks to the Commission (SECY-81-13) was used as guidance in determining those rules and regulations deemed to be particularly significant. It is the position of Louisiana Power
& Light Company that Waterford SES-Unit 3 meets or exceeds the requirements of all applicable regulations. Our confidence in this conclusion stems not only from the enclosed review, but also in the design process and quality assurance programs of Louisiana Power & Light Company, the NSSS vendor, and our consultants. Additional verification is provided by the independent review of the NRC staff. All of these factors, taken together, provide reasonable assurance that the public health and safety will be protected.
If we can be of further service, please advise.
Yours very truly, D. L. Aswell DLA:ver Attachment ec: Mr. E. L. Blake Mr. W. M. Stevenson po\
W 81050.403 6 A
% e STAIE OF LOUISIANA)
) SS PARISH OF ORLEANS )
D. L. Aswell, being duly sworn, states that he is Vice President - Fower Production of Louisiana Power & Light Company and that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission this document.
D. L. Aswell SL'BSCRIBED AND SWORN to before me a Notary Public in and for the Parish and State above named, this my of -
V7
/' hw Notar % 1flic "
My Commission expires:
Y h{fh 1
-,,r
l l
I l
Before the UNITED STATES NUCLEAR REGULATORY COMMISu*dN DOCKET NO. 50-382 In the Matter of Louisiana Power & Light Company J
COMPLIANCE WITH TITLE 10, CODE OF FEDERAL REGL7 ATIONS Louisiana Power & Light Company, Applicant in the above captioned proceeding, hereby files a document comparing the Waterford Steam Electric Station - Unit 3 to Applicable Title 10 Code of 'e'ederal Regulation parts.
Wherefore, Applicant requests the licenses specified under Docket No.
50-382.
Respectfully submitted, LOUISIANA POWER & LIGHT COMPANY BY D. L. Aswell Vice President - Power Production l
DATE: 4-29-81 I
1 l
. . _ . _ . _ _ . . _ _ ~ _ _
REGULATION COMPLIANCE 10CFR 19.1 This regulation states the general purposes of this part and does not impose any specific requirements.
19.2 This regulation delineates the scope of this part and does not impose any specific requirements.
19.3 This regulation delineates definitions used in this part. LP&L will adhere to those definitions in all applicable documents.
19.4 This regulation governs the interpretation of regulations by the NRC and does not impose any spe:ific requirements.
19.5 This regulation gives the address of the NRC and does not impose any specific requirements.
19.11 This regulation governs the posting of notices to workers.
Waterford compliance with this regulation is ensured through appropriate administrative procedures.
19.12 This regulation governs instructions given to workers working in or frequenting any portion of a restricted area. Waterford compliance with this regulation is ensured through Health Physics and administrative p ;cedures in addition to worker training (as described in FSAR Section 13.2).
19.13 This regulation governs the notifications and reports given to individuals concerning personal radiation exposure data. Waterford compliance with this regulation is ensured through appropriate Health Physics and administrative procedures.
19.14 When required by and la accordance with this regulation, LP&L allows workers to consult privately with an NRC inrgector, and allow a worker representative to e.: company th; '.nspection.
19.15 This regulation governs NRC consultation with workers during inspection. Waterford complies with this regulation.
19.16 This regulation allows workers who believe that a violation of the Act has occurred to request an NRC inspection. These, workers shall not be discriminated against in any way by the Licensee. Waterford complies with this regulation.
REGULATION COMPLL\NCE 10CFR 19.17 This regulation merely states the procedures which may be followed in the event an inspection is not deemed warranted pursuant to a request in accordance with paragraph 19.16. It does not impose any specific requirements.
19.30 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 specific requirements.
19.31 This regulation provides for the granting of exemptions from 10CFR19 regulations, provided such exemptions are authorized by law and will not result in undue hazard to life or property. It does not impose any specific requirements.
19.32 This regulation prohibits the licensee from any sexual discrimination under any program or activity licensed by the NRC. Waterford compliance with this regulation is ensured through administrative procedures.
I I
- 1
~
l L REGULATION COMPLIANCE
! IUCFR i
- 20.101 The radiation dose limits specified in this regulation are complied with through:
a) Conservative design considerations b) Health Physics procedures, and c) Administrative policies and controls Ccuformance is documented in FSAR Sections 12.1 and 12.3 and will be further documented by the use of appropriate personnel monitoring devices and the maintenance of all required records.
20.102 (a) When required by and in accordance with this regulation, individuals will submit an appro-priate written, signed statement. Appropriate health physics procedures and administrative procedures control this process.
(b) When required by this regulation, the accumu-lated 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 procedures control this process. (FSAR Chapter 13.3.5) 20.103 (a) Compliance with this regulation is ensured through the implementation of appropriate health physics procedures relating torair sampling for radioactive materials, and bioassay of individuals for internal conta-mination pursuant to the requirements of paragraph (a)3 of this section.
20.103 (b) Appropriate process and engineering controls and equipment, as described in Chapters 9, 11, and 12 of the FSAR, are installed and operated to maintain levels of airborne radioactivity as low as reasonably achievable.
When necessary, as determined by Waterfor 4 administrative and Health Physics guidelines, I additional precautionac1 procedures are utilized to lLait the potential for intake of radioactive materials. ,
20.103 (c) Therequirementsofthisregulationareensured) by the proper use of approved respiratory i protection equipment. Waterford administra-I l
l i
l REGULATION COMPLIANCE tive procedures incorporate 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 specific obligations on licensees.
20.103(e) Waterford will make the appropriate noti-fication in accordance with this regulation.
20.103(f) The Waterford Respiratory Protection Program will be in full accordance with the require-ments of 20.103 (c) prior to core loading.
20.104 Conformance with this regulation is assured by appropriate LP&L policies regarding employment of individuals under the age of 18 and the Waterford Health Physics Manual restricting these individuals access to restricted areas.
20.105(a) Chapter 12.2 of the FSAR provides the information and related radiation dose assessments specified by this regulation.
20.105(b) The radiation dose rate limits specified in this regulation are complied with through the implementation of Waterford procedures, Technical Specifications, and administrative policie5 which control the use and transfer of radioactive materials. Appropriate surveys and monitoring devices document this compliance.
20.106 (a) Conformance with the limits specified in this regulation is assured through the implementation of Waterford procedures and applicable Technical Specifications which i provide adequate sampling and analyses, and '
monitoring of radioactive materials in effluents before and during their release.
Monitoring e%Cluent.. releases is carried out
-in.accordance with Regulatory Guide 1.21 and is discussed in Chapter 11 of the FSAR.
Appropriate surveys and monitoring devices document this compliance.
~ -
k,,
m REGULATION COMPLIANCE 1
j 20.106 (b) LP&L has not used and does not intend to 20.106 (c) include in any license or amendment appli- J cations proposed limits higher than those specified in 20.106 (a) , as provided for in these regulations.
20.106 (d) Appropriate allowances for dilution and dispersion of radioactive effluents are made in conformance with this regulation, and are described in detail in Chapter 11 of the FSAR.
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 specific obligations on licensees.
20.106 (f) This regulation merely states that the pro-visions of 20.106 do not apply to disposal of radioactive material into sanitary sewerage systems. It imposes no specific 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 specific obligations on licensees.
20.108 Necessary bioassay equipment and procedures including Whole Body Counting, are utilized at Waterford to determine exposure of in-dividuals to concentrations of radioactive materials. Appropriate health physics and .'.dministrative procedures implement this requirentent.
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.
Waterford Health Physics Manual and appliccble 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 Waterford Health Physics Manual and applicable health physics procedures set forth policies and practices which ensure that all individuals are supplied with, and required to use, appropriate personnel
REGULATION COMPLIANCE monitoring _ equipment.
20.202(b) The terminology set forth in this regulation is accepted and conformed to in all appli-cable Waterford procedures, Technical Specifications, and those portions of the Waterford Health Physics Manual in which its use is made.
- 20. 203 ( A) All materials used for labeling, posting, or otherwise designating radiation hazards or radioactive materials, and using the radiation symbol, conform to the conventional design prescribed in this regulation.
20.203(b) This regulation is conformed to through the implementation of appropriate health physics procedures and portions of the Health Physics Manual relating to posting of radiation areas, as defined in 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 Technical Specification 6.12 and appropriate plant health physics pro-cedures, as well as the Waterford Health Physics Manual. The control and other protective measures set forth in the regu-lation are maintained under the surveillance of the Health Physics group.
20.203(d) Each ..irborne Radioactivity Area, as defined in this regulation, is required to be posted by provisions of the Health Physics Manual and appropriate health physics procedures.
These procedures also provide for the sur-veillance 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 Waterford Health Physics Manual.
20.203(f) The container labeling requirements set forth in this regulation are complied with through the implementation of appropriate health physics procedures, and portions of the Waterford Health Physics Manual.
l l
l
i REGULATION COMPLIANCE 10CFR 20.204 The posting requirement exceptions described in this regulation are used where appropriate and necessary at Waterford Adequate controls are provided within the Waterford 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 Waterford Health Physics Manual and appropriath health physics procedures. These procedures also provide for the necessary documentation to ensure an auditable record of compliance.
20.206 The requirements of 10CFR19.12 referred to by this regulation are satisfied. Appropriate health physics procedures set forth require-ments for all radiation workers to receive this instruction on a periodic basis.
20.207 The storage and control requirements for licensed materials in unrestricted areas are conformed to and documented through the implementation of Waterford health physics procedures and applicable portions of the Waterford Health Physics Manual.
20.301 The general requirements for waste disposal set forth in this regulation are complied with through appropriate administrative procedures. Chapter 11 of the FSAR describes the Radioactive Waste Management System installed at Waterford.
20.302 No such application for proposed disposal procedures, as described in this regulation, has been made or is contemplated by LP&L.
20.303 No plans for waste disposal by release into sanitary sewerage systems, as provided for in this regulation, are contemplated by Waterford.
20.304 Disposal of wastes by burial in soil (i.e.,
onsite burial) , as provided for in this ;
regulation, is not being contemplated by )
Waterford.
1 l
i
REGULATION . COMPLIANCE 10CFR 20.305 Specific authorization, as described in tlis regulation, is not currently being sought by LP&L. for treatment or disposal of wastes by incineration.
20.401 All of the requirements of this regulation are complied with through the implementation of Technical Specification 6.10 and health physics procedures pertaining to records of surveys, radiation monitoring and waste dis-
- posal. The retention periods specified for such records are also provided for in these specifications and procedures.
20.402 Waterford 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 l in Technical Specification 6.9.1.8.
20.403 Notifications of incidents, as described in this regulation, are assured by the require-ments of Technical Specification 6.5.1.6, the 4
Waterford Health Physics Manual and appro-priate plant procedures, which also provide for the necessary assessments to determine 4
the occurrence of such incidents.
20.405 Reports of overexposures to radiation and the occurrence of excessive levels and con-centrations, as required by this regulation, are provided for by reference in Technical Specification 6.9.1.9 and in appropriate health physics procedures.
20.406 This regulation was deleted August 17, 1973, effective September 17, 1971(38 Fed. Reg. 22220).
20.407 The personnel monitoring report required by this regulation is expressly provided for l by Technical Specification 6.9.1.3. Appro-priate health physics procedures establish j the data base from which this report is generated.
~
._- - . _ . _ _ . - .. _ _____s.,_---
_7-REGULATION COMPLIANCE 20.408 The report of radiation exposure required by this regulation upon termination.of an individual's employment or work assignment is generated through the provisions of LP&L procedures.
20.409 The notification and reporting requirements of this regulation, and those referred to by it, are satisfied by the provisions of LP&L procedures..
l I
v v
o .
REGULATION COMPLIA' ICE 10CFR 21.1 This regulation states the general purposes of this part and does not bapose any specific requirements.
21.2 This regulation merely establishes the applicability of this part and imposes no specific requirements.
21.3 The definitions contained in this regulation will be adhered to in all applicable documents.
21.4 This regulation governs the interpretation of regulations by the NRC and does not impose any specific requirements.
21.5 This regulation gives the address of the NRC and does not impose any specific requirements.
21.6 This regulation governs the posting of certain notices.
Waterford compliance is ensured through appropriate administrative procedures.
21.7 This regulation provides for the granting of exemptions from 10CFR21 regulations, provided such exemptions are authorized by law and will not endanger lif e or property or the conson defense and security, and are otherwise in the public interest. It Laposes no specific r equirements.
21.21(a) This paragraph governs the evaluation of a deviation and the reporting of same to a responsible officer.
Waterford compliance is assured through appropriate Quality procedures.
21.21(b) This paragraph governs the reporting of a deviation to the NRC. Waterford compliance is ensured through appropriate administrative procedures.
21.21(c) As required by this paragraph, individuals subject to 21.21(b) of this part will supply the Commission with-available additional information.
21.31 As required by this regulation, each procurement document (as applicable) shall specify that the provisions of 10CFR Part 21 apply.
21.41 As required by this regulation, the licensee shall permit
, authorized representatives of the Commission to make l inspections as allowed by this part.
~. . :- - - --.- - . .. .:. . .
A REGULATION COMPLIANCE 10CFR 21.51 As required by this regulation, LP&L will maintain appropriate records as ide. ified in this procedure.
Compliance is ensured through appropriate administrative procedures.
21.61 This regulation describes the renedies which the Commission may obtain in order to enf orce its regulations, and sets forth those penalties or punishnents which may be imposed for violations of its rules. It does not impose any specific requirementa.
i 1
4
- _,, . ,1-.. ..-,, _ _ . -.. -- -,,_ -,
~ . _ . . _ _ . , .. .
. o REGULATION COMPLIANCE 10CFR 50.36 Tecnnical Specifications have been prepared which include items in each of the categories specified, including: (1) safety limits and lLaiting safety settings, (2) limiting con-ditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls. These Technical Specifications will be revised prior to commercial operations to reflect the latest Standard Technical Specifications (FSAR Chapter 16).
50.36 (a) The Radiation Technical Specifications will include specifications which require compliance with 10CFR50.34a(releases as low as is rea-sonably achievable), and that ensure that concentrations of radioactive effluents released to unrestricted areas are within the limits specified in 10CFR20.016. The reporting requirements of 10CFR50.36a (a) (2) will also be included in these spccifications.
The Radiation Technical Specifications will be submitted prior to commercial operation to reflect the latest Standard Technical Specifications.
50.46 The Waterford Emergency Core Cooling System meets the requirements of this regulation (FSAR Sections 6.3 and 15.6) .
50.54 This regulation specifies certain conditions that are incorporated in every license issued.
Compliance is effected simply by including these conditions in the license when it is issued.
- 50. 55a (a) (1) In accordance with this paragraph, structures, systems, and cr.mponents are designed, fab-ricated, erected, constructed, Pested, and inspected to quality standards cemmensurate with their safety importance. Applicable FSAR discussions are in accordance to Regu-latory Guide 1.70 Revision 2.
50.55a (a) (2) This paragraph is a general statement leading into paragraphs (c) through (i) of the regulation.
i
l REGULATION COMPLIANCE 50.55a(b) This paragraph merely provides guidance concerning the approved Edition and Ad-denda of Section III and XI of the ASME Boiler and Pressure Vessel Code.
50.55a (c) These paragraphs delineate the codes and 50.55a (d) standards to which various components must 50.55a (e) adhere. All Waterford components subject 50.55a (f) to 50.55a meet or exceed the design and construction requirements as discussed in FSAR Section 5.2.1 and in the letter LP&L 8254 dated February 24, 1978.
50.55a (g) This paragraph delineates the ASME Code preservice inspection requirements to which applicable components must adhere.
As discussed in FSAR Section 5.2.1, Waterford meets or exceeds these require-ments.
50.55a (h) In accordance with this regulation, Waterford protection systems meet IEEE 279-1971.
50.55a (i) In accordance with this regulation, Fracture Toughness requirements of Appendices G and H have been satisfied at Waterford.
50.55a (j) This paragraph specifies exemptions for certain facilities. It is not ap-plicable to Waterford.
50.70 This regulation discusses requirements for allowing plant access and providing adequate office facilities for an NRC inspector. LP&L will comply with this regulation.
50.71 Records are and will be maintained in accordance with the requirements of sections (a) through (e) of this regulation and the license.
50.80 This regulation provides that license r.ay not be transferred without NRC consent. No application for a transfer of a license is involved in the Waterford proceeding. l 3
. , ~. . _ . . ~ .
1 REGULATION COMPLIANCE 50.81 This regulation permits the creation of !
mortgages, pledges, and liens on licensed facilities, subject to certain provisions.
LP&L licensed facilities are not mortgaged, pledged, or otherwise encumbered as those terms are used in this. reg @lation.
50.82 This requirement provides f or the termi-nation of licenses. Shoulu LP&L request a license terminatien, LP&L will comply with this regulation.
50.109 This regulation specifies the conditions under which the NRC may require the back-fitting of a facility. LP&L will comply with this regulation.
l l
w
l l
I REGULATION COMPLIANCE 10CFR50 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.
10CFR50 Appendix E This Appendix specifies requirements for emergency plans. FSAR Chapter 13.3 describes the provisions of the emergency plans which ensure that Waterford meets the requirements of Appendix E.
100FR50 Appendix G This Appendix specifies fracture toughness requirements for reactor coolant pressure boundary components.
Fracture toughness requirements of Appendix G have been satisfied by alternate methods of evaluation as
! discussed in FSAR Sections 5.2.1 and 5.2.3.
10CFR50 Appendix H This Appendix specifies reactor vessel material surveillance program require-ments. Technical Specification.-4.4.9.1.2 and operating procedures will be im-2- plemented to ensure compliance with this Appendix. Further discussion can be found in FSAR Section 5.3.1.6.
10CFR50 Appendix J This Appendix specifies containment leak rate testing requirements. Tech-nical Specifications 3.6.1.2 and 3.6.1.3 ensure Waterford compliance with this Appendix.
10CFR50 Appendix K This Appendix specifies features of acceptable ECCS evaluation models.
As discussed in FSAR Section 6.3 and 15.6 Waterford is in compliance '
with this Appendix.
l l
{
l l
l l
l :
l ..
REGULATION COMPLIANCE GDC 1 Structures, systems, and components im-portant to safety are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The structures, systems, and components im-portant to safety are listed in Table 3.2-1. Recognized codes and standards are applied to the equipment in these classi-fications as necessary to assure a quality product in keeping with the required safety function. The total quality assurance program is described in Chapter 17 and is applied to the safety class 1, 2 and 3, and seismic category I items contained in this table. The intent of the quality assurance program is to assure sound engineering in all phases of design and construction through conformity to regulatory require-ments and design bases described in the license application. In addi* ion, the program assures adherence to specified standards of workmanship and implementation of applicable codes and standards in fabrication and construction. It also includes the observance of proper preope-rational testing and maintenance procedures (Chapter 14) as well as the documentation of the foregoing by keeping appropriate records. The total quality assurance program of the applicant and its principal contractors meets the quality-related requirements of Appendix B to 10CFR50.
Records are maintained which demonstrate that the requirements of the quality asaurance program are satisfied. This documentation shows that appropriate codes, standards and regulatory require-l mer.ts are observed, specified materials i
are used, correct procedures are utilized, rualified personnel are provided and that the finished parts and components meet the applicable specifications for safe and reliable operation. These records are available so that any desired item of information is retrievable for reference.
Ihese records of the design, fabrication, erection and testing of structures, systems, and components important to safety are maintained as required by the LP&L quality assurance program.
l
-REGULATION COMPLIANCE GDC 2 The integrity of systems, structures and components important to safety is included )
in the reactor facilities design evalua- ;
j tions. The structures, systems and compo- l nents important to safety are designed to withstand the effects of natural phenomena without loss of capability to perform their safety functions. Those structures, systems l 3
and components vital to the shutdown ca-pability of the reactor are designed to withstand the maximum probable natural phenomenon expected at the site determined from. recorded data for the site vicinity with appropriate margin to account for uncertainties in historical data. Those structures, systems and components vital to the mitigation and control of incident conditions are designed to withstand the
) effects of a loss-of-coolant accident coincident with the effects of the safe shutdown earthquake. The structures,
- systems and components important to safety are listed in Table 3.2-1.
For further discussion, see the following sections: 2.3 Meteorology, 2.4 Hydrologic j Engineering, 2.5 Geology, Seismology and j Geotechnical Engineering, 3.2 Classification i of Structures, Components and Systems, 1 3.3 Wind and Tornade Loadings, 3.4 Water j~ Level (Flood) Design, 3.5 Missile Protection, 3.7 Seismic Design, 3.8 Design of Category I Structures, 3.9 Mechanical Systems and Components, 3.10 Seismic Qualification of Seismic Category I Instrumentation and Electrical Equipment, and 3.11 Environmental i Design of Mechanical and Electrical Equipment.
GDC 3 Noncombustible and fire resistant materials are used wherever practical throughout the i facility, particularly in areas containing l critical portions of the plant such as l containment structure, control room and j components of the Engineered Safety Features l
Systems.
Safety related systems are designed and located to minimize the effect of fires or explosions on redundant components.
Facilities for the storage of combustible material are designed to ninimize both the probability and the effects of a fire.
, - - - rmy6 --- ---r-. - -q-- p'em m- -y-- ----m 3 ye-p -
yr ,-yw w - y yo---py- *- -,-- ge-y--
REGULATION COMPLIANCE Equipment and facilities for fire pro-tection, including detection, alarm and extinguishment are provided to protect both plant and personnel from fire or explosion and the resultant release of toxic vapors. Both wet and dry type fire-fighting equipment are provided.
Normal fire protection is provided by deluge systems, halon systems, sprinklers, hose lines and portable extinguishers.
The Fire Protection System is designed such that failure of any component of the system will not impair the ability of redundant equipment to safely shutdown and isolate the reactor or limit the release of radioactivity to the environ-ment in the event of a postulated accident.
The Fire Protection Systems are provided with test hose valves for periodic testing.
All equipment is accessible for periodic inspection.
Fire protection for cable systems is dis-cussed in Subsection 8.3.3 and the Plant Fire Protection System in Subsection 9.5.1.
GDC 4 Structures, systems and components important to safety are designed to accommodate the effects and to be coppatible with the pressure, temperature, humidity, chemical and radiation conditions associated with ;
normal operation, maintenance, testing, 1 and postulated accidents, including a loss- l of-coolant accident in the area in which they are located.
Protective walls and slabs. local missile shielding, or restraining devices are .
provided to protect the containment and j Engineered Safety Features Systems within !
and without the containment against damage l from missiles generated by equipment failures. The concrete enclosing the Reactor Coolant System serves as radiation shielding and as an effective barrier against internal missiles. Local missile barriers are provided for control element drive mechanisms. Penetratic s and piping extending to and including isolation valves are protected from damage due to pipe
l REGULATION COMPLIANCE whipping, and are protected from damage by external missiles, where such protection is necessary to meet the design bases.
Non-seismic category piping is arranged or restrained so that failure of any non-seismic category piping will not cause radioactivity to be released to the en-vironment nor prevent essential seismic Category I structures or equipment from mitigating the consequences of such an accident.
Seismic Category I piping has been arranged or restrained such that, in the event of rupture of a seismic Category I pipe which causes a loss-of-coolant accident, resulting pipe movement, will not result in loss of containment integrity and adequate En-gineercl Safety Features Systems operation will be maintained.
The containment interior structure is designed to sustain dynamic load which could result from failure in major equip-ment and piping, such as jet thrust, jet impingement, and local pressure transients, where containment integrity is needed to cope with the conditions.
The external concrete shield protects the steel containment vessel from damage due to external missiles such as tornado pro-pelled missiles. The functional capability of any safety related structures, systems or components located outdoors (e.g.,
co611ng towers) are designed for protection against externally generated missiles.
For those components which are required to operate under extreme conditions such ,
as design seismic loads or containment l post accident environmental conditions, 4 the manufacturers submit type test, ope- l rational or calculational data which !
substantiate this capability of the equipment.
For further discussion, refer to the following sections: 3.3 Wind and Tornado Loadings, 3.4 Water Level (Flood) Design, 3.5 Missile Protection, 3.6 Protection Against Dynamic Effects Associated with
l REGULATION COMPLIANCE l l
the Postulated Rupture of Piping, 3.7 )
Seismic Design, 3.8 Design of Category I Structures, 3.11 Environmental Design i of Mechanical and Electrical Equipment, '
and 6.0 Engineered Safety Features.
I GDC 5 As per the Louisiana Power & Light letter (LPL-362) of October 19, 1971 to Dr P A Morris (then with the AEC) , Unit No. 4 is no longer being considered for construction; therefore, this criterion is not applicable.
GDC 10 In ANSI N18.2, Nuclear Safety Criterion for the Design of Pressurized Water Reactor Plants (January 1973), plant conditions are categorized in accordance with their anticipated frequency of occurrence and risk to the public, and design requirements are given for each of the four categories. The categories covered by this criterion are Condition I - Normal Operation and Condition II-Faults of Moderate Frequency.
The design requirement for Condition I is that margin shall be provided between any plant parameter and the value of that paramet2r which would require either automatic or manual protective action.
This condition is met by providing an i adequate control system (refer to Section !
7.7). The design requirement for Con-dition II is that such faults shall be
> accomodated with, at most, a shutdown of the reactor, with the plant capable of returning to operacion after corrective action. On Waterford-3, this condition is met by providing an adequate protective system (refer to Section 7.2 and Chapter 15).
Specified acceptable fuel design limits are stated in Subsection 4.4.1. Operating limits, to ensure specified acceptable fuel design limits are met, are prescribed in the Technical Specifications (limiting I conditions for operations) which support Chapters 4 and 15. Operator action, aided by the control systems and monitored by
- plant instrumentation, maintains the plant within technical specification li-mitations. For further discussion see l
i l'
l - . .- --
REGULATION COMPLIANCE the following sections: 4.2.1 Reactor Fuel, 5.0 Reactor Coolant, 5.4.7 Decay Heat Removal, 7.2 Reactor Protective System.
GDC 11 In the power operating range, the com-bined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coeffi-cient, and the moderator pressure coeffi-cient to an increase in reactor power in the power operating range is a decrease in reactivity, i.e., the inherent nuclear feedback characteristics, are not positive.
The reactivity coefficients for this reactor are listed in Table 4.3-4 and are discussed in detail in Section 4.3.
GDC 12 Power level oscillations will not occur.
The effect of the negative power coeffi-cient of reactivity (refer to criterion 11), together with the coolant temperature program maintained by control element assemblies (CEAs) and soluble boron, provide fundamental mode stability. Power level is monitored continuously by neutron flux detectors (refer to Chapter 7) and by reactor coolant temperature difference measuring devices.
Power distribution oscillations are detected by neutron flux detectors.
Axial mode oscillations are suppressed by means of CEAs. Radial oscillations are expected to be convergent. It is a design objective that azimuthal xenon oscillations be convergent. Monitoring and protective requirements imposed by criterion 10 and 20 are discussed in those responses and in Chapter 4.
GDC 13 Instrumentation is provided to monitor significant process variable that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary (RCPB) , and the con-tainment and its associated systems.
Controls are provided for the purpose of maintaining these variables within the limits prescribed for safe operation.
The principal process variables to be monitored and controlled are neutron
REGULATIud COMPLIANCE level (reactor power), axial neutron flux shape, CEA position, reactor coolant tem-perature, reactor coolant pump speed, pressurizer liquid level and pressure, and stem generator level and pressure.
In addition, instrumentation and control are provided for monitoring of activity level and monitoring and controlling baron concentrations in the reactor coolant.
Control room indication is provided for all parameters required for normal operation and accident conditions.
The Plant Protection System (PPS) consists of the Reactor Protective System (RPS) and the Engineered Safety Features Ac-tuation System (ESFAS) . The RPS monitors the reactor operating conditions and effects reliable and rapid reacto- . rip if any monitored variable or combination of monitored varaibles deviates from the permissible operating range to a degree that a safety limit may be reached (refer to Section 7.2). The ESFAS monitors plant operating conditions and initiates ESF operation in the event of a certain postulated accident .efer to Section 7.3).
The non-nuclear safety grade Core Operating Limit Supervisory System (COLSS) aids the operator with an independent indication of the proximity to specified core ope-rating limits and an alarm when one of these limits is reached.
In-core instrumentation is provided to supplement information on core power distribution and to provide a means fdr calibration of out-of-core flux detectors.
Instrumentation is provided to monitor plant variables and systems under post-accident conditions and to follow the course of the accident, as described in Section 7.5.
- The instrumentation and control systems are described in detail in Chapter 7, and the boronometer and the process radiation monitor are discussed in subsection 9.3.4.
- - - - ~ . . .-%
1 REGULATION COMPLIANCE GDC 14 Reactor Coolant System (RCS) components i
are designed in accordance with ASME l
' Code,Section III, Division I. The l
establishment and implementation of ope-rating quality control, inspection, and I testing as required by this standard and allowable reactor pressure-temperature operations within allowable limits, ensure the integrity of the RCS.
The RCPB accommodates system pressures and temperatures attained under all ex-pected modes of unit operation including all anticipated transients, and maintains the stresses within applicable stress limits.
Piping and equipment pressure parts of the RCPB are usually assembled and erected by welding. Planged, screwed or com-pression joints, when used, are in compliance with applicable codes. Welding procedures, are employed which produce welds ci complete fusion and free of unacceptable defects. All welding pro-cedures, welders and welding machine operators are qualified in accordance with the raquirements of Section IX of the ASME Boiler and Pressure Vessel Code for the materials to be welded.
Qualification records, including the results of procedure and performance qualification tests and identification symbols assigned to each welder are main-tained.
The pressure boundary has provisions for in-service inspection in accordance with j Section XI of the ASME Boiler and Pressure
! Vessel Code, to ensure the continued structural and leaktight integrity of the boundary (see also response to Cri-terion 32 and Subsection 5.2.4). For the reactor vessel, a material surveillance program conforming with the requirements of Appendix H to 10CFR50 is given in Subsection 5.3.1.6.
Means are provided to detect significant leakage from the RCPB with monitoring i
' readouts and alarms in the control room, ,
as discussed in Subsection 5.2.5.
- ,. : _-- - ~ :- - __
y
l REGULATION COMPLIANCE J l
GDC 15 The design criteria and bases for the i RCPB are described in the response to Criterion 14.
The operating conditions established for noimal steady and transient plant ope-rations are discussed in Chapter 5. The normal operating limits are selected so that an adequate margin exists between them and the design limits. The plant control systems maintain the plant variables well within the established opere. ting limits. Plant transient response charac-teristics and pressure and temperature distributions during normal operations are considered in the design as well as the accuracy and response of the instruments and controls.
These design techniques ensure that a satisfactory margin is maintained between the plant's normal operating conditions, including design transients, and the design limits for the RCPB.
The RPS minimizes the deviation from normal operating limits in the event of anticipated operational oc:urrences, (ANSI N18.2 Condition II occurrences) .
Analyses for this plant show that the design limits for the RCPB are not exceeded in the event of any ANSI N18.2 Condition II occurrence: Faults of Moderate Frequency. For further discussion refer to the following sections: 5.2 Integrity of Reactor Coolant Pressure Boundary, 5.4.1 Reactor Coolant Pumps, and 7.2 Reactor Trip System.
GDC 16 The Containment System is designed to provide for protection of the public from the consequences of a loss-of-coolant accident, based on a postulated break of the reactor coolant piping up to and in-cluding a double-ended break of the largest reactor coolant pipe.
The containment vessel, Shield Building, and'the Engineered Safety Features Systems are designed to safely withstand all i
l
- ~ - - - - - - - - - - -
_. _ . _ _ - _ _ _ _ - - . _ ~_.
REGULATION COMPLIANCE l internal and external environmental con-ditions that may reasonably be expected to occur during the life of the plant, including both short and long term effects following a loss-of-coolant accident. Due consideration has been given to all site factors and local environment as they relate to public health and safety. For further discussion, see the following sections: 3.8.4.1.1 Shield Building, 3.8.2 Design of Steel Containment, 6.2 Containment Systems, 15.0 Accident Analysis.
GDC 17 A summary description of the electric power system is provided in Section 8.1.
Full descriptions of the offsite and on-site power systems are included in Sections 8.2 and 8.3, respectively. All onsite emergency and vital equipment, as required to meet the safety function defined above, is redundant, with each division fed from separate and independent engineered safety feature (ESF) buses.
Alternate power systems are provided as follows:
a) Several 230 kv transmission lines, any of which is capable of supplying power for the engineered safety fea-tures in the event of loss of auxil-iary transformer powcr.
b) Two half-capacity auxiliary trans-formers directly connected to the main generator 25 kV isolated phase bus to supply power for the unit under normal operating conditions. The transformers also provide auxiliary j power to the unit when the main gene- '
rator is disconnected from the 230 kV system and the unit is carrying its own auxiliaries.
c) Two half-capacity start-up transformers to provide start-up pcver and full capacity standby auxiliaries service 1 (engineered safety features loads) from the 230 kV switchyard.
REGULATION COMPLIANCE d) Two independent on-site diesel generator sources are each capable of supplying 100 percent power for one of the two redundant Engineered Safety Features System trains in the event of a loss of auxiliary transformer power and start-up transformer power.
The Transmission System will provide re-liable sources of offsite power for sup-plying the station auxiliary power system for plant start-up, shutdown, or at any time that power is unavailable from the station main generator. All transmission lines approach the plant along a common right-of-way on independent structure.
Although in the same right-of-way the two lines are spaced sufficiently far apart that a falling transmission tower cannot involve other line.
In the event of a loss of all offsite power sources, standby onsite diesel generators and station batteries provide the necessary power for safe shutdown or, in the event of an incident, to restrict the consequerces to within acceptable limits. Both the onsite ESF dc and standby ac power systems consist of redundant and independent power sources and distribution systems such that a single failure will not prevent either system from performincy its safety functions.
A review of systems stability is performed to confirm that a very small probability exists of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission net-work, or the loss of power from the onsite electric power supplies.
GDC 18 Electrical power systems important to safety are designed to permit appropriate periodic inspection and testing of im-portant areas and features such as wiring,
~
REGULATION COMPLIANCE insulation, connections, and switchboards, to assess the continuity of the systems and to detect deterioration, if any, of their components. Capability is provided to periodically test the operability and functional perfdrmance of the components of the systems. The diesel generators are started and loaded periodically on a routine basis and relays, switches, and buses are inspected and tested for operation and availability on an individual basis.
Transfers from normal to emergency sources of power are made to check the operability of the systems and the full operational sequence that brings the systems into operation.
For those components which are required to operate under extreme conditions, such as design earthquake seismic loads or containment post-accident environment
- parameters, the manufacturers submit type test; operational or calculational data which substantiates this capebility of the equipment (refer to Sections 3.10 and 3.11) .
For further discussion, refer to the fol- '
lowing Subsections: 8.3.1.2 Analysis AC Power Systems, 8.3.2.1 Analysis of
) DC Power Systems, Technical Specification
' (Emergency Power System Periodic Tests).
GDC 19 Following proven power plant design phi-losophy, all control stations, switches, controllers and indicators necessary to operate and shut down the nuclear unit and maintain safe control of the facility are located in one common control room.
The design of the main control room l
(Section 6.4) permits safe occupancy during abnormal conditions without per-sonnel receiving radiation exposure in excess of five rem whole body or its equivalent. Shielding is designed to maintain tolerable radiation exposure levels (See Section 12.1) in the main ;
control room for postulated accident conditions, including a loss-of-coolant accident. The main control room is
_ _ . _ __ _ _ _ . _ _ _ _ _~.;-- . _c. _ , - --,__--._- _- -
=
. _ _ ~ _ _
l REGULATION COMPLIANCE l pressurized relative to the outside at-mosphere following the occurrence of a radiological accident. Food, water and other habitability systems are provided for main control room personnel for the duration of any postulated accident.
Positive air pressure is maintained in the main control room after receipt of a safety injection actuation signal, a containment purge isolation signal or a high radiation signal. The Main Control Room Air Conditioning System is provided with radiation and toxic chemical detec-tors and alarms. The main control room is isolated during a postulated toxic chemical accident. Provisionn are made for main control room air to be recir-l culated through high-efficiency particu-late and charcoal fileters following l
i any accident. Emergency lighting is pro-vided (see Subsection 9.5.3) .'
Alternate controls and instruments at a location outside the main control room are available for those items of equipment required to bring the plant to, and maintain it in, a hot standby condition.
It is also possible to reach a cold shut-down condition from locations outside of the main control room in a reasonable period of time through the use of suitable procedures (see Subsection 7.4.11.
GDC 20 A Plant Protection System (PPS) is provided to monitor reactor and plant operating conditions and automatically initiate a reactor trip when the monitored variable or combination of variables approach l
[
specified limiting safety system settings.
These limiting safety system settings are
' selected to ensure that the design basis anticipated operational occurrences do not cause acceptable fuel design limits (linear heat rate and departure from l nucleate boiling ratio (DNBR) to be l
! exceeded. Section 7.2 describes specific ;
reactor trips and provides the list of anticipated operational occurrences accommodated.
l l
J
REGULATION COMPLIANCE Reactor trip is accomplished by de-ener-gizing the control element drive mechanism (CEDM) holding latch coils through the interruption of the CEDM power supply.
The control element assemblies (CEAs) are thus released to drop into the core, ra-pidly inserting negative reactivity to shut down the reactor. The CEDMs are described in Subsection 3.9.4.
The PPS also functions to monitor certain accident conditions and autanatically initiate various required Engineered Safety Features Systems and their support systems when the monitored variables reach their set points. The parameters that automatically actuate ESF are described in Section 7.3. Controls are provided for manual actuation of ESF.
l The specified acceptable fuel design I limits on linear heat rate and DNBR are intended to enforce the principal thermal hydraulic design basis given in Subsection 4.4.1, i.e., the avoidance of thermally-induces fuel damage during normal steady-state operation and during anticipated operational occurrences. The specified acceptable fuel design limit on linear heat rate is specifically intendad to prevent fuel melting.
Clad strain lLmits are not explicitly
[
addressed by the specified acceptable fuel des 39n limits on linear heat rate cnd min; um DNBR. However, the specified acceptacle fuel design limits, in con-junction with the limiting conditions for operation, define possible reactor ope-l rating conditions that are considered in the calculation of clad strain.
GDC 21 The PPS is designed to provide high func-tional reliabilitr and in-service testa-bility. The protectxon system is designed to comply with the requirements of IEEE Standard 279-1971. No single failure will result in the loss of the protection function. The protection channels are independent, e.g., with respect to piping, wire routing, mounting, and supply of
L REGULATION COMPLIANCE power. This independence permits testing and the removal from service of any com-ponent or channel without loss of the protection function.
Each channel of the PPS, from the sensors up to the final actuation device, is capable of being checked by comparison of outputs of sbnilar channels that are presented on indicators and/or rccorders on the control board. Trip units and logic are tested by inserting a signal into the measurement channel ahead of the readout and, upon application of trip level input, observing that a signal passes through the trip unit and the logic to the logic output relays. The logic output relays are tested individually for initiation of trip action. The parallel trip circuit breakers that supply power to the CEDM holding coils may be tested during reactor operation without effecting a reactor trip.
The benefit of a system that includes four independent and redundant channels is that the system can be operated, if need be, with up to two channels out of service (one bypassed and another tripped) and still meet the single failure criteria.
The only operating restriction while in this condition (effectively one-out-of-two logic) is that no provision is made to bypass another channel for periodic testing or maintenance. The system logic must be restored to at least a two-out-of-three condition prior to removing another channel for maintenance or testing.
Plant Protection System reliability and testability are discussed in Subsections 7.2.2 and 7.3.2.
I GDC 22 The PPS conforms to the provisions of IEEE Standard 279-1371. Four independent mea-surement channels complete with sensors, i
sensor power supplies, signal conditioning units, and bistable trip units are provided
_ _ . _ _ _.-. . . . . .m -
REGULATION COMPLIANCE for each protective parameter monitored by the protection systems. The measure-ment channels are provided with a high degree of independence by separate con-nections of the channel sensors to the process systems. Power to the channels is provided by independent nuclear ins-trumentation buses (see Chapter 7)'.
Functional diversity is incorporated in the system design to the extent that is practical to prevent loss of the protec-tion function. ;
The PPS is functionally tested to ensure satisfactory operation prior to installation in the plant. Environmental and seismic .
qualifications are also performed utilizing type tests and specific equipment tests as discussed in Sections 3.10 and 3.11.
GDC 23 PPS trip channels are designed to fail into a safe state or into a state established as acceptable in the event of loss of power supply or disconnection of the system. A loss of power to the CEDM holding coils results in insertion of all CEAs by gravity into the core. Redundancy, channel inde-pendence, and separation are incorporated in the PPS design to minimize the possibi-lity of the loss of protection function under adverse environmental conditions f.cee Chapter 7 and the response to Cri-terion 22).
- GDC 24 The PPS is separated from the control instrumentation systems so that failure or removal from service of any control instrumentation system component or ,
channel does not inhibit the function - <
l of the PPS and will leave intact a pro-tection system satisfying all reliability,
! redundancy and independence requirements
. of the protective system (refer to Section t l 7. 2) . !
l i i
i
- ; - n~ - - ..: ::_- . _ _ - : - .. ...._.:~:~:_____-_. . . _ , --
. . .. - ~_ . - -
REGULATION COMPLIANCE 4
GDC 25 Shutdown of the reactor is accomplished
-by opening of the reactor trip breakers that interrupt power to the CEDM holding coils. Actuation of the trip breakers is independent of any existing control signals.
The protection system is designed such that specified acceptable fuel design limits are not exceeded for specified single malfunctions of the reactivity control systems, including the withdrawal of a single full or part length CEA.
A definition of the specified single malfunctions of the reactivity control, systems accommodated by the protection system design is included in Section 7.2.
Analyses of specified control malfunctions are provided in Chapter 15.
GDC 26 Two independent reactivity control systems of different design principles are provided.
The first system, using CEAs, includes a positive means (gravity) for inserting CEAs, and is capable of reliably control-ling reactivity changes to ensure that under conditions of normal operation, including anticipated operational occur-rences, specified acceptable fuel design limits are not exceeded. The CEAs can be mechanically driven into the core.
The appropriate margin for stuck rods is provided by assuming in the analyses of anticipated operational occurrences that the highest worth CEA is stuck out of the core.
The second system, the Chemical and Volume Control System (CVCS) , uses neutron absorbing soluble boron and is capable l of reliably compensating for the rate of i
reactivity changes resulting from planned normal power changes (including xenon burnout) such that acceptable fuel design !
limits are not exceeded. This system is capable of holding the reactor suberitical under cold conditions. For a further description, see Subsection 9.3.4.
l i
l
.. _. -. . .-- .. c -_ _. .-
l REGULATION COMPLIANCE Either system is capable of making the core suberitical from a hot standby con-dition. ]
For further discussion, see Sections 7.4 )
and 7.7.
GDC 27 The reactivity control systems, which provide the means for making and holding the core subcritical under postulated accident conditions, are discussed in Section 4.3 e.nd Subsection 9.3.4. Com-bined use et CEAs and chemical shim control by the Chemical and Volume Control Systems (CVCS) provides the shutdown margin required for plant cooldown and long-term xenon decay, assuming the highest worth CEA is stuck out of the core.
During an accident, the Safety Injection System injects concentrated boric acid into the Reactor CQolant System for long-term and short-tera cooling and for reactivity control. Dctails of the system are given in Section 6.3.
GDC 28 The bases for CEA design include ensuring that the reactivity worth of any one CEA is not greater than a preselected maximum value. The CEAs are divided into three sets: a shutdown set, a regulating set, and a part length set. These sets are further subdivided into groups as neces-4 sary. Administrative procedures and control interlocks ensure that the amount and rate of reactivity increase is limited to predetermined values. The regulating
' groups are withdrawn only after the shutdown groups are fully withdrawn.
The regulating groups are programmed to move in sequence and within limits which prevent the rate of reactivity addition and the worth of individual CEAs from exceeding limiting values. For the spe-cified list of design bases anticipated operational occurrences, the CEA positions aremonitored by the RPS, and a trip is initiated in the event that specified acceptable fuel design limits are approached (see Sections 4.3 and 7.7) .
1 l
~ . . . . - -.
REGULATION COMPLIANCE The maximum rate of reactivity addition that is produced by the CVCS is too low to induce any significant pressure forcec that might rupture the RCPB or disturb the reactor vessel internals.
The RCPB (refer to Chapter 5) and the reactor internals (refer to Chapter 4) are designed to appropriate codes (refer to the response .to Criterion 14) . They can accommodate the static and dynamic loads associated with an inadvertent, sudden release of energy, such as that resulting from a CEA ejection or a steam line break, without rupture and with 1Luited deformation that will not impair the capability of cooling the core.
GDC 29 Tha1 design bases anticipated operational occurrences considered in the design of the RPS and the reactivity control systems are defined in Section 7.2.
1 Consideration of redundancy, independence, and testability in the design, coupled with careful component selection, overall system testing, and adherence to detailed quality assurance, ensure an extremely high, probability that safety functions are accomplished in the event of antici-pated operational occurrences (refer to Chapters 4, 7 and 9).
GDC 30 The RCPB components are designed, fabri-cated, erected, and tested in accordance :
with the ASME Code,Section III. All major components are classified safety class 1 i as specified in Subsection 3.2.2. Accor- ,
dingly, they receive all of the quality assurance measures appropriate to that classification. l Detection and identification of reactor :
coolant leakage is discussed in Subsection j 5.2.5. The system is designed to detect, :
and, to the extent practical, identify ,
the source of reactor coolant leakage.
Further discussion relating to qu&lity f of the RCPB is contained in Section 5.2.
.__..___..7,_-._.
l r
j REGULATION COMPLIANCE l
' constructed in accordance with ASME Code,Section III and comply with the test and inspection requirements of these codes.
These requirements ensure that flaw sizes are limited so that the probability of failure by rapid propagation is extremely remote. Particular emphasis is placed upon the quality control applied to the reactor vessel, on which tests and ins-pections exceeding ASME code requirements areperformed. These tests and inspections are summarized in Sections 5.2, and 5.4.
Carbon and low-alloy steel materials that form part of the pressure boundary are assessed for fracture toughness in accordance with Branch Technical Position MTEB 5-2, l
Fracture Toughness Requirements. Through
- this approach, the available test data is used to estimate fractare toughness in the same terms as the new requirements set forth by Appendix G of 10CFR50.
Excessive neutron-induced changes of the reactor vessel material due to neutron radiation is prevented by providing an annulus of coolant water between the re-actor core and the vessel. In addition, to minimize the effects of irradiation on material toughness properties on core beltline materials, restrictions are placed on upper limits for those residual chemical elements that directly influence the nil ductility transition temperature (NDTT) shifts. This is accomplished through material specifications for the plates and deposited welds. Specifically, upper limits are placed en copper, phos-phorus, sulfur and vanadium. ;
The maximum integrated fast neutron flux .
exposure of the reactor vessel wall oppo- l site the midplane of the core is less l than 3.68 x 1019 nyt. This value assumes
~
a 40 year vessel design life, with the -
plant at the design power level 80 percent ;
j of the time. The maximum expected increase in transition temperature is about 160F.
The actual change in material toughness f properties due to irradiation are verified !
periodically during plant lifetime by a ,
material surveillance program conforming l
REGULATION COMPLIANCE to the requirements of ASTM-E-185 as revised in 1973. Based on the reference nil ductility temperature (RT rating restrictions are appliNST)'asUE*"
ne-cessary to limit vessel stresses.
The thermal stresses induced by the in-jection of cold water into the vessel, following a LOCA, were examined. The test results and analysis show that there is no gross yielding across the vessel wall using the minimum specified yield strength in the ASME Boiler and Pressure Vessel Code,Section III, Division 1.
GDC 32 Provisions are made in the design for ins-pection,. testing and surveillance of the RCS boundary as required by ASME Boiler and Pressure Vessel Code Section XI and Section III, Division 1, as applicable.
Tne reactor vessel surveillance program conforms with ASTM-E-185, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," as revised in 1973. The details of the reactor surveillance program are given in Section 5.2. Samplo pieces taken from the same shell p. ate material used in fabrication of the beltline region of the reactor vessel are installed between the core and the vessel inside wall. These samples are removed and tested at intervals during vessel life to provide an indication of the extent of the neutron-induced changes in mechanical properties at the vessel wall. Charpy tests are performed on the samples to develop a Cahrpy tran-sition curve. By comparison of this curve with the Charpy curve and drop 7 weight tests for specimens taken at the beginning of the vessel life, the change of RT is determined and operating proceduresakhdstedasrequired.
The survefilance program described in Section 5.3 includes provisions that com-ply with the NRC regulation, Reactor Vessel Material Surveillance Program Re-l
l REGULATION COMPLIANCE quirements, 10CFR50, Appendix H, published in the Federal Register on July 17, 1973.
The only exception between the Waterford-3 Surveillance Program and the Appendix H requirements is in Appendix H,Section II.C.2, Attachments to the Reactor Vessel:
"In adhering to the requirements of placing the surveillance specimens as close as possible to the reactor vessel wall, the capsule holders are attached to the cladding of the reactor vessel and are not major load-bearing components. The method of attachment was described to the NRC in the CE Topical Report CENPD-155P, CE Pro-cedures for Design, Fabrication, Installa-tion and Inspection of Surveillance Specimen Holder Assemblies. In the Topical Report Evaluation, the NRC concluded that the proposed method of attachment did comply with the intent of this section which was to ensure that the method of attachment did not cause any degradation of the base material, prevent in-service inspections, nor produce any unacceptable loads in the reactor vessel. By such placement, tem-perature, flux spectrums, and fluence differences between the surveillance specimens and the reactor vessel are mi-nLuized, thereby permitting more accurate assessment of the changes in the reactor vessel properties."
GDC 33 Reactor coolant makeup during normal ope-ration is provided by the Chemical and Volume Control System (CVCS). The design incorporates a high degree of functional reliability by provision of redundant components and an alternate path for charging. The charging pumps can be powered from either onsite or offsite i power sources, including the onsite emergency diesel generators.
There are three charging pumps associated t with the CVCS. One of these pumps is normally in operation balancing the letdown purification flow and the reactor coolant pump controlled bleed-off flow rate. A complete system functional description is provided in Subsection 9.3.4. ,
1 l
l
\
- " ~ '
REGULATION COMPLIANCE It is not the function of the CVCS to provide protection against small breaks; this safety function is provided by the Safety Injection System (SIS) . The CVCS does have the capability, with only one charging pump available, of replacing the flow loss to the Reactor Building for leaks in the reactor coolant piping up to 0.50 inch equivalent diameter.
However, loss of this CVCS capability in no way compromises the safety of the reactor plant.
GDC 34 The' Shutdown Cooling System provides residual heat removal for reactor coolant temperature of less than 350 F. For temperatures greater than 350F, the steam generators provide this function. The design incorporates sufficient redundancy, interconnections, leak detection, and isolation capability to ensure that re-sidual heat removal is accomplished, assuming failure of a single active com-ponent. Either system removes fission product decay heat at a rate that prevents violation of acceptable fuel design limits and the design conditions of the RCPB.
The Shutdown Cooling System and the steam generator auxiliaries are designed to operate either from offsite or onsite power sources.
l Further discussion is included in Sub-section 9.3.6 for the Shutdown Cooling System and in Chapter 10 for the Steam l
and Power Conversion System.
GDC 35 Emergency core cooling is provided by the Safety Injection System (SIS) des-cribed in Section 6.3. This system pro-l vides adequate borated cooling water to remove heat at a rate sufficient to maintain the fuel in a coolable geometry and to ensure that zirconium-water reaction is limited to a negligible amount (less than one percent) , Detailed analysis is performed to verify that the system per-formance is adequate to satisfy the new NRC Acceptance Criteria for ECCS for Light Water Power Reactors (10CFR50, Appendix K, January 4,1974) . Details
REGULATION COMPLIANCE of this analysis are provided in Sub-section 6.2.1.5, Section 6.3, and Chapter 15.
GDC 36 The Safety Injection System layout ar-rangement and design facilitates access to all critical components. All pumps, valves, and piping external to the Reactor Building are readily accessible for periodic inspection to ensure system leaktight integrity. Valves, piping, and tanks inside the Reactor Building are inspected for leaktightness during plant shutdowns for refueling and main-tenance.
Reactor vessel internal structures, reactor coolant piping, and safety in-jection nozzles are accessible for visual inspection for wear due to erosion, cor-rosion, or vibration and nondestructive inspection techniques in accordance with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code.
Details of the inspection program are described in Chapters 5, 6 and 16 as appropriate. !
GDC 37 The Safety Injection System provides the ]
testing capability required to demonstrate .
system and component operability. Testing l is conducted during normal plant operation l with the test facilities arranged so that they will not interfere with performance of the systems or with the initiation of control circuits, as described in Sub-section 6.3.4.
The SIS permits periodic-testing of the delivery capability up to a locatiLn as l close to the core as practicable. Periodic injection into the RCS from the SIS during normal operation is not practical.
During normal operation, RCS pressure exceeds high pressure safety injection (HPSI) pump shutoff head. Periodic pressure testing of the HPSI System to assure system integrity is possible using the cross connection from the charging pumps in the CVCS.
REGULATION COMPLIANCE With the plant at operating pressure, operation of high and low pressure safety injection pumps is verified by recircu-lation back to the refueling water storage pool (RWSP). This permits verification of flow path continuity in the high pres-sure injection lines and suction ?.ines from the RWSP.
In addition, the low pressure safety in-jection pumps are used as shutdown cooling pumps during normal plant cooldown. The pumps discharge into the safety insection header via the shdtdown cooling heat exchangers and the low pressure injection lines.
Borated water from the safety injection tanks is bled through the recirculation test line to verify flow path continuity from each tank to its associated main safety injection header.
During refueling, blowdown tests provide additional evidence of safety injection tank operability. Inadvertent HPSI pump actuation at the beginning of plant cool-down does not cause RCS heatup/cooldown ibnitations to be exceeded. Relief valves on the shutdown cooling (SDC) lines provide protection from accidental HPSI pump operation during SDC. Thus, no cests are rqquired to cover this particular aspect.
The operational sequence tnat brings the Safety Injection System into action, including the transfer to alternate power sources, can be tested in parts as described in Section 0.3, Subsection 7.3.2 and Section 8.3.
GDC 38 The containment Spray System, consisting l
of two pumps and two shutdown heat ex-changers, and the Containment Cooling l
System, consisting of four fan coolers, function as emergency containment heat removal systems. Each of these systems has the full heat removal capability required for the most severe postulated i
- ~ ' ' w - . ~_ ,_ _ _ _
-,g w_ w- ,ewh-e .
REGULATION COMPLIANCE loss-of-coolant accident.
The systems are provided with the emer-gency onsite power necessary for their operation assuming a loss of offsite power. The systems taken together pro-vide the necessary capability for contain-ment heat removal assuming a single failure in either system or in the emer-gency onsite power supply.
The Containment Spray System and the Containment cooling System are described in Subsection 6.2.2.
GDC 39 The Containment Spray System essential equipment except for risers, distribution header piping, spray nozzles and the containment sump are located outside of the containment.
The containment uump, spray piping, and nozzles may be inspected for leaktight-ness during plant shutdowns for refueling and maintenance. Piping, ptmps, heat uxchangers, and valves external to the containment structure are readily acces-l sible for periodic inspection to check i system leaktight integrity.
1 Portions of the Containment Cooling System entirely within the ontainment can be inspected at appropriate intervals during refueling shutdowns. Cooling water systems external to the containment which service the Contairment Cooling System are accessible for inspection at any time during plant operation.
In-service inspections of the Containment Spray System and Containment Cooling System are performed as indicated in Section 6.6.
GDC 40 System piping, valves, pumps, fans, heat exchangers, and other components of the containment heat removal system are de-signed to permit appropriate periodic testing to assure their structural and leaktight integrity. The components are
1 1
REGULATION COMPLIANCE arranged so that each component can be tested periodically for operability and required functional performance.
Three of the four containment cooling units are normally in operation. The fourth unit will be rotated in service with the other three for normal contain-ment cooling. Transfer to alternate power sources can also be tested.
The operational sequence that would bring the Containment Spray System into action, including the transfer to alternate power sources, can be tested. With the plant at operating pressure, the contain-ment spray pumps acid valves may be operated by recirculation back to the refueling water storage pool. This will permit verification of flow path continuity in the spray lines and suction lines from the refueling water storage pool to the first isolation valve outside the containment.
Testing of the Containment Spray System and Containment Cooling System is per-formed ac indicated in Subsection 6.2.2.
GDC 41 The Shield Building Ventilation System (SBVS) , which consists of two full ca-3 pacity redundant fan and filter systems, is designed consistent with the functioning of other associated systems, to reduce the concentration and quantity of fission products released to the environment following postulated accidents, including a loss-of-coolant accident. This is established by maintaining a subatmospheric pressure within the Shield Building annulus to ensure that post accident activity leakage from the steel containment is routed through the filterrsystem., This system is described in Subsection 6.2.3.
The Containment Atmosphere Release System l (CARS) and the Hydrogen Recombiner System, l discussed in Subsection 6.2.5, prevents l the buildup of dangerous concentrations of hydrogen in the containment following
6 REGULATION COMPLIANCE a loss-of-coolant accident. Operation will normally be initiated when the hydrogen concentration within the containment reaches a predetermined set point as determined by containment air sampling.
The Containment Spray System, discussed in Subsection 6.5.2, provides for the removal of iodine from the containment atmosphere following a LOCA.
The ope-ration is initiated by the containment spray actuation signal.
The Shield Building Ventilation System, the Containment Atmosphere Release System, Hydrogen Recombiner System, Hydrogen Analyzer System and Containment Spray System have suitable redundancy to assure that for onsite electrical power system operation only, or for offsite electrical
- power system operation only, their safety functions can be accomplished, assuming a single failure.
GDC 42 The only components of the containment atmosphere cleanup systems inside the Shield Building are the ductwork of the SBVS, hydrogen recombiners and the containment spray nozzles and piping.
These can be inspected during shutdown.
The balance of equipment is located in the Reactor Auxiliary Building, where it is accessible for physical inspection.
Ducts, plenums, and casings will be pro-vided with access doors for internal inspection at approrpiate times, i
specific inspection programs are dis-cussed in Subsecticn 6.2.5.4 for the combustible gas control systems and ccm-ponents, Subsection 6.5.1.4 for the filter systems that are required to per-
- j. form a safety related function following l a design basis accident and Subsection ,
6.5.2.4 for the Containment Spray System.
I
~D REGULATION COMPLIANCE CDC 43 The Shield Building Ventilation System is designed and constructed to permit periodic pressure and functional testing. i For purpose of periodically testing the l retentive capability of the filterrsystem, test cartridges are placed in the filter housing in locations which allow the cartridges to be subjected to the same air currents as the beds. These are periodically removed and tested.
High efficiency particulate (HEPA) and carbon filters, associated with the Shield Building Ventilation System, are located outside the containment for convenience for testing and inspection. Periodic tests are described in Subsection 6.5.1.4.
Active components of the Shield Building Ventilation System, Containment Atmosphere Release System, Hydrogen Recombiner System, Hydrogen Analyzer System and Containment Spray System can be tested periodically for operability and required functional performance.
The full operational sequence that would bring the SBVS, CARC, hydrogen recombiners and Containment Spray System into action, including the transfer to alternate power sources and the design capability, can be tested. Testing provisions are discussed in Subsections 6.2.5.4, 6.5.1.4 and 6.5.2.4.
GDC 44 The Component Cooling Water System (CCWS) and the Auxiliary Component Cooling Water System (ACCWS) are designed to transfer heat from structures, systems and com-ponents important to safety, to the cooling towers. Two redundant, completely independent trains are provided, each of which is capable of removing the heat associated with normal operation or accident conditions.
The Component Cooling Water System is a closed loop cooling water system that includes three full capacity pumps, two heat exchangers and two dry cooling towers.
1
... y REGULATION- COMPLIANCE The cooling tower is pumped by the component cooling water pumps , through the dry cooling towers and the tube side of the CCW heat exchangers, through the components being cooled and back to the pumps.
The Auxiliarf Component Cooling Water System ramovea heat, if required, from the CCW3 via the CCW heat exchangers and dissipa:es it to the atmosphere. The ACCWS consists of two independent loops which include two CCWS heat exchangers (shell side) , two full capacity pumps, two wet type, mechanical draft cooling towers and two cooling tower basins, each of which store sufficient water to complete a safe shutdown based upon the occurrence of a LOCA and minimum safeguards operation.
The piping, valves, pumps and heat ex-changers in each system are arranged so that the system safety functions can be performed assuming a single system failure. The essential' headers of each system will each be automatically isolated from the nonessential headers during emergency mode of operation.
Each system is normally pressurized per-mitting leakage detection by routine sur-veillance or monitoring instrumentation.
l Electric power for the operation of each
- system may be supplied from off-site or i onsite emergency power sources, with distribution arranged such that a single failure will not prevent the system from performing its safety function.
The CCWS and ACCWS are discussed in Sub-section 9.2.2.
GDC 45 The CCWS and ACCWS are designed to permit the required periodic inspections of heat exchangers and piping. Three CCW pumps are provided, two of which serve the two system loops used in normal ope-l The third pump can operate on ration.
either loop, allowing inspection and maintenance of a pump while maintaining redundant system capability.
~ _ - - - . _ .-
1 REGULATION COMPLIANCE In-service inspection of the CCWS and ACCWS is performed as discussed in Section 6.6.
GDC 46 Two CCW pumps are normally operating, one per loop. Normally, both dry towers are continuously operated. Therefore, the structural and leaktight integrity of the components and the operability of their active components are demonstrated in this way. Data is taken periodically during normal plant operation to confirm heat transfer characteristics.
The ACCW pumps, wet towers and CCW heat exchangers are operated periodically to ensure their operability and to confirm performance requirements.
The systems are designed to permit testing of system operability, encompassing si-mulation of emergency reactor shutdown or LOCA conditions including the transfer between normal and emergency power sources.
The testing procedures are discussed in Subsection 9.2.2.
GDC 50 The containment structure, including access openings and penetrations, is designed !
with sufficient conservatism to accommodate, without exceeding the design leak rate, the transient peak pressure and temperature associated with a postulated reactor coolant piping break, up to and including a double-ended rupture of the largest reactor coolant 4 pipe. The containment design basis is discussed in Subsection 6.2.1.1.
The containment structure and Engineered Safety Features Systems are evaluated for various combinations of energy release.
The analysis accounts for system thermal and chemical energy, and for nuclear decay heat. The cooling capacity of containment heat removal systems is adequate to prevent overpressurization of the structure, and to return the containment to near at-mospheric pressure as discussed in Sub-section 6.2.2.
l
?>
REGULATION COMPLIANCE GDC 51 The containment vessel material (ASTM-SA516 Grade 70), is normalized to refine the grain structure, which results in improved ductility. In addition, the actual mecha-nical and chemical properties of the ma-terial are documented and are within the limits for minimum ductility defined in ASTM-A516.
The containment vessel is built to Sub-section NE of Section III of the ASME Boiler and Pressure Vessel Code which re-quires that materials shall be impact-tes-ted at a temperature at least 30F below the lowest metal service temperature.
These tests do not determine the nil-duc-tility transition temperature of the ma-terial but ensure that this temperature is at or below the test temperature.
The design of the vessel reflects consi-deration of all ranges of temperature and loading conditions which apply to the vessel during operation, maintenance, testing and postulated accident conditions.
Because this vessel is post weld heat treated, residual stresses from welding are minimal. Steady state and transient stresses are calculated in accordance with accepted methods.
All pressure boundary double butt welds that comprise the containment vessel are 100 percent radiographed and the acceptance standard of the radiographs ensures that flaws in welds do not exceed the maximum allowed by ASME Code.
Containment boundary design is discussed in Subsection 3.8.2.
GDC 52 The containment vessel is designed so l that initial integrated leakage rate i testing can be performed at design l pressure after completion and installation '
l of penetrations and equipment.
Provisions are made in the containment design to permit periodic leakage rate !
tests, at reduced or peak pressure, to !
verify the continued leaktight integrity f of the containment.
f
. - - .- = . . = = . -z. :=2z.m=- . ..- =n .-.::r:m ::: -: :::-:^
N REGULATION COMPLIANCE Periodic integrated leakage rate testing will be carried out in accordance with the requirements of Appendix J to 10CFR50.
A description of the periodic integrated leakage rate testing is provided in Sub- I I
section 6.2.6.
GDC 53 The absence of insulation on the contain-ment vessel permits appropriate periodic inspection of the accessible interior and exterior surfaces of the vessel.
The lower portions of the containment i vessel are totally encased in concrete I and will not be accessible for inspection after?theracceptanceetesting. There ;
will be no need for any special in-service j surveillance program due to the rigorous design, fabrication, inspection and pres-sure testing the containment vessel receives prior to operation. Visual inspection of the accessible interior and exterior surface of the containment vessel will be made.
Provisions are made to permit periodic testing of penetrations which have resilient seals or expansion bellows to allow leak-tightness to be demonstrated at contain-ment design pressure. Inspection and testing of the containment is carried out in accordance with Appendix J of 10CFR50.
Provisions for testing and inspection are discussed in Subsection 6.2.6.
GDC 54 Piping penetrating the containment vessel shell is designed to withstand at least
. a pressure equal to the containment vessel maximum internal pressure. The design bases require a dobble barrier on all of the above systems not required to Ibnit the consequence of accidents, so that no single failure or malfunction of an active component can result in loss of isolation or intolerable leakage. Valves are designed to a maximum allowable leakage 7
of 1/10 of a standard cubic foot of air per hour per inch of diameter of nominal valve size at containment design pressure.
Valves isolating penetrations serving
~ .__---- r., ,
)
, REGULATION COMPLIANCE Engineered Safety Feature Systems will not automatically close with the contain- I ment isolation actuation signal (CIAS), j but may be closed by remote manual ope-ration from the main control room to isolate any Engineered Safety Feature System when required. Proper valve closing tbnes are achieved by appropriate selection of valve, operator type, and operator size. To ensure continued in-tegrity of the containment isolation system, periodic closure and leakage rate tests will be performed to insure that leakage will be within specified limits based upon maintaining post accident site boundary doses within acceptable guidelines.
Design and isolation requirements for piping systems penetrating the containment are provided in Subsection 6.2.4.
GDC 55 Except for the safety injection and CVCS charging lines, the reactor coolant pres-sure boundary as defined in 10CFR50 is located within the containment. The safety injection and CVCS charging lines are closed seismic Categoty I piping systems outside containment with isoln-tion valves that meet the isolation cri-teria of GDC 55. Isolation valves are located as close to the containment as practical.
Valves covered by the above criterion are described in Subsection 6.2.4.
GDC 56 The lines which connect directly to the containment atmosphere and penetrate the primary containment are provided with two valves in series where they penetrate the containment, so that failure of one active component will not prevent iso-lation. Each of these lines meet the isolation criteria of GDC 56.
As described in Subsection 6.2.4, the l
containment sump penetrations contain l valves outside the containment which are j never open during normal operation. In
, , y REGULATION COMPLIANCE addition, the lines form a closed seismic Category I system outside of containment and, after a LOCA, the suction portion of these lines inside containment are covered by water.
Valves covered by the above criterion are described in Subsection 6.2.4.
GDC 57 Each line that penetrates the reactor containment, and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere, has at least one containment isolation valve located outside the con-tainment as & lose to the containment as practical. Each of these lines meet the isolation criteria of GDC 57.
valves covered by the above criterion are described in Subsection 6.2.4.
GDC 60 The facility controls the release of radioactive materials in gaseous and liquid effluents and handles radioactive solid wastes produced during normal reactor operation, including anticipated operational occurrences. The raioactive waste management systems minimize the potential for an inadvertent release of radioactivity from the facility and ensure that the discharge of radioactive wastes is maintained in accordance with the limits 10CFR50, Appendix I. Radioactive materials which do not meet release limits will not be discharged to the environment.
The Waste Management System is designed with sufficient holdup capacity and flexibility for processing of wastes to I
ensure that releases are as low as rea-sonable achievable. ;
l The Solid Radwaste System is capable of handling all radioactive solid wastes produced by the plant for shipment offsite.
The Radioactive Waste Processing System, the design criteria, and the amounts of estimated releases of radioactive effluents to the environment are described in Chapter 11.
w - - _ ,
- y - - -
. .y REGULATION COMPLIANCE GDC 61 The Fuel Pool Cooling System, Fuel Hand-ling System (FHS) , and Radioactive Waste Processing System ensure adequate safety normal and postulated accident conditions.
The Fuel Pool Cooling System provides cooling to remove residual heat from the fuel stored in the spent fuel pool. Data is taken periodically during normal plant operation to confirm heat transfer capa-bilities and differential across compo-nents. The Fuel Pool Cooling System is described in Subsection 9.1.3.
The spent fuel pool meets seismic Category I requirements and is protected against postulated missiles so that no postulated accident could cause excessive loss-of-coolant inventory.
Most of the components and systems in this category are in frequent use and no special testing is required. These systems and components important to safety which are not normally operating are tested periodi-cally at appropriate intervals.
The spent fuel storage racks are covered by water which provides sufficient shielding over stored fuel assemblies to limit radiation at the surface of the water to no more than 2.5 mr/hr during the storage period. The exposure time during refueling is limited so that the integrated dose to operating personnel does not exceed the limits of 10CFR20. Adequate shielding is provided as described in Section 12.3.
Radiation monitoring is provided as dis-cussed in Sections 11.5 and 12.3.
Individual components that contain radioactivity are located in confined areas and are ventilated through appro-priate filtering systems as discussed in Subsection 9.4.2.
l l
l
\ ..
l 3
REGULATION COMPLIANCE GDC 62 A safe geometric spacing is provided for both new and spent fuel assemblies which are stored in racks in parallel rows. An edge-to-edge spacing remployed for the new fuel storage racks results in a k of 0.98 or less utilizing dry storage,.gg The edge-to-edge spacing for the spent fuel storage racks results in k of 0.95' i
l orlesswithouttakingcreditf8bthe boron in the fuel pool water. New and spent fuel storage is described in Section 9.1.
GDC 63 Monitoring and alarm instrumentation are provided for fuel and waste storage and handling areas for conditions that might contribute to a loss of continuity in decay heat removal and to radiation exposure. Area radiation monitors des-cribed in Section 12.3 are provided to detect and alarm excessive radiation le-vels in the Fuel Handling Building and Waste Management System areas.
The heat generated in the waste storage facilities is low and therefore does not require a specific heat removal system.
The normal area ventilating systems are sufficient.
Control room alarms are provided to alert j the operator to high and low liquid level I and high temperature in the fuel pool. A
( low pressure alarm on the fuel pool pumps' discharge header is provided to warn of interruption of the cooling flow. Ins-trumentation is discussed in Subsection 9.1.3.
GDC 64 Provisions are made for monitoring the containment atmosphere, the f acility i
effluent discharge paths, the operating areas within the plant and the facility environs for radioactivity that could be released from normal operations, from anticipated operational occurrences and from postulated accidents.
Those liquid and gaseous wastes containing radioactive matter are processed by the
.
- s t) l l
REGULATION COMPLIANCE Waste Management System which functions to remove radioactive material from these wastes by filtration, ion exchange or distillation prior to discharge. In the event of high radiation, the wastes will be stored until the radioactivity has decayed sufficiently to permit discharge.
Liquid wastes are grab sampled, and if the contained activity meets applicable limits they may be released with continuous monitoring to the circulating water dis-charge.
Gaseous wastes are compressed and stored in the gas decay tanks. The gas is sam- ,
pled to determine radioactivity concen-tration to assure release limits are not exceeded, and is monitored during release through the plant vent.
Solid wastes that are produced will be packaged in licensed shipping containers and transported offsite for disposal.
Radioactivity of the contents of contai-ners will be monitored.
Area radiation monitors are discussed in 4 Section 12.3.
4 Instrumentation is provided to monitor t plant variables and systems under post accident conditions and to follow the i course of the accident, as described in Section 7.5.
L 4
l t
t
REGULATION COMPLIANCE 55.1 This regulation states the general purpose of Part 55 and does not impose any specific requirements.
55.2 This regulation states the applicability of Part 55 and does not impose any specific requirements.
55.3 This regulation states that only those persons licensed by the NRC to perform operator or senior operator functions may perform those functions. Appropriate administrative procedures ensures Waterford compliance.
55.4 The definitions contained in this part will be adhered to in all tpplicable Waterford documents.
55.5 This regulation gives the address of the NRC and does not impose any specific re-quirements.
55.6 This regulation governs the interpretation of regulations by the NRC and does not impose any specific requirements.
55.7 This regulation states that the NRC may grant exemptions to regulations of this part as it deems appropriate. It does not impose any specific requirements.
55.8 This regulation states that the NRC may impose additional requirements as it deems appropriate. It does not impose any specific requirements.
55.9 This regulation merely provides exemptions from this part and does not impose any specific requirements.
55.10 This regulation describes the required contents of an application of a license as provided for in this part. License applications for Waterford personnel are'irr accordance withothis> regulation.
55.11 This regulation states the requirements for NRC approval of a license application.
It does not impose any specific requirements.
+
I l
1 I
REGULATION COMPLIANCE 55.12 This regulation provides the terms on which a license applicant whose appli-cation has been denied may submit a re-application. It does not impose any specific requirements.
55.20 This regulation provides the scope of the operator and senior operator examinations and does not impose any specific require-ments.
55.21 This regulation describes the contents of the operator written examination and does not impose any specific requirements.
55.22 This regulation describes the contents of the senior operator written examination and does not impose any specific require-ments.
55.23 This regulation describes the scope of the operator and senior operator operating test. It does not impose any specific requirements.
55.24 This regulation permits the NRC to waiver any or all of the requirements for the written and operating tests if the applicant meets the criteria contained in this regulation. It 6:. a not impose any specific requirements.
55.25 This regulation r etmits the NRC to adminis-ter a simulated operating test to an applicant for a license if the criteria in this regulation are met. It does not impose any specific requirements.
55.30 This regulation states that the NRC may issue a license once the requirements of the Act and regulations of the NRC are met and the NRC deems it appropriate.
55.31 This regulation describes the conditions of a license and allows the NRC to impose any further conditions it deems appropriate.
It does not impose any specific requirements.
l l
1 1
. w. p REGULATION COMPLIANCE 55.32 T ;is regulation states that each operator sad senior operator license shall expire
'wo years after the date of issuance.
It does not impose any specific require-ments.
55.33 This regulation describes the required contents of an application for renewal of a license, and the terms under which the NRC will renew a license. It imposes no specific requirements.
55.40 This regulation provides the terms and conditions under which the NRC may modify or revoke a license. It imposes no specific requirements.
55.41 This regul-clon requires the licensee to report to the NRC of any disability referred to in 55.11 of this part. Waterford compliance is ensured through appropriate administra-tive procedures.
55.50 This regulation describes the remedies which the NRC may obtain in order to enforce its regulations, and sets forth those penalties or punishment which may be imposed for violations of its rules.
It does not impose any specific require-ments.
55.60 This regulation merely delineates the proper form an applicant and examining physician must complete and include in a license application pursuant to 55.10 of this part.
It imposes no independent requirements.
35 Appendix A In accordance with this regulation, an l operator and senior operator requalifi-cation program will be instituted at Waterford. This program will be in ac-cordance with the requirements specified in this Appendix.
l l
.>..9 l
l l
REGULATION COMPLIANCE 71.31 This regulation defines general standards for packaging of radioactive material.
Waterford will employ only packaging which conforms to these standards. Com-pliance is assured through appropriate administrative procedures.
71.32 This regulation governs the structural standards used in the design of large quantity (Type B) packaging. Type B packaging used at Waterford will conform to these standards. Compliance is assured through appropriate administrative pro-cedures.
71.33 This regulation governs the criticality standards used in the design of fissile material packages. Fissile material packaging used at Waterford will conform to these standards. Compliance is assured through appropriate administrative procedures.
71.34 This regulation defines the general con-ditions used to determine the effect of transport environment on a radioactive material package. Radioactive material packages used at Waterford will be only those tested in accordance with this regulation. Compliance is ensured through appropriate administrative procedures.
72.35 This regulation merely defines the design acceptance criteria used to evaluate the effects of normal transport on a package in accordance with 71.34 of this part.
It imposes no independent requirements.
71.36 This regulation merely defines the design acceptance criteria used to evaluate the effects of hypothetical accident conditions on a package in accordance with 71.34 of this part. It imposes no independent requirements.
71.37 In accordance with this regulation, arrays of fissile material packaging used at Waterford will be evaluated according to the standard set forth in 71.38, 71.39, and 71.40 of this part. Compliance is ensured through appropriate administrative procedures. l i
i
. ,o REGULATION COMPLIANCE 71.38 This regulation merely defines the design considerations and acceptance criteria to be used in evaluating the effects of accident conditions on Fissile Class I packages in accordance with 71.37 of this part. It imposes no independent require-ments.
71.39 This regulation merely defines the design considerations and acceptance criteria to be used in evaluating the effects of accident conditions on Fissile Class II packages in accordance with 71.37 of this part. It imposes no independent require-ments.
71.40 This regulation merely defines the design considerations and acceptance criteria to be used to evaluate the effect of accident conditions on Fissile Class III packages in accordance with 71.37 of this part.
It imposes no independent requirements.
71.41 This regulation permits the use of packages constructed and authorized for use prior to January 1, 1967, subject to certain conditions. Should such a package be employed at Waterford, this regulation will be complied with.
71.42 This regulation defines special requirements for shipments of plutonium. LP&L does not expect to ship any plutonium as defined in this subpart. Should plutonium shipments as specified become necessary, procedures will be implemented to ensure compliance with this regulation.
l
._7e sm REGULATION COMPLIANCE 100.1 This regulation states the general pur-poses of Part 100 and does not impose any specific requirements.
100.2 This regulation establishes the applica-bility of Part 100 and does not impose any specific requirements.
100.3 The definitions contained in this regu-lation will be adhered to in all applicable Waterford documents.
100.10 As required by this regulation, site spe-cifics, including seismology, meteorology, geology, and hydrology, as well as the exclusion area, low population zone, and population center distance are presented in FSAR Chapter 2, and have been included in the sita evaluation. Also included in the evaluation were reactor design and engineered safeguard features.
100.11 (a) In accordance with this regulation, an exclusion area, low population zone and the population centar distance have been established and are described in FSAR Section 2.1.
(b) This paragraph applies only to multiple reactor facilities and is not applicable to Waterford.
100 Appendix A This Appendix provides geologic and seis-mic siting criteria. The geologic and l
seismic characteristics of the Waterford .
site are described in FSAR Chapter 2. !
Waterford design is in compliance with I
this regulation as discussed in FSAR Chapter 3. ,
l
._ _ _ - - _