ML17212A861

Forwards Itemized Review of Compliance W/Significant Rules & Regulations,Per 10CFR,in Response to NRC 810929 Request

ML17212A861
Person / Time
Site:	Saint Lucie
Issue date:	10/02/1981
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-81-429, NUDOCS 8110070252
Download: ML17212A861 (56)

Text

RKGUL RY INFORMATION DISTRIBUT SYSTEM (RIDS)

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VAMEl RECiIPIENT" AFFILKA'TION 'ECIP KKSEiVHUT'iD'iG, Division of'icensing

SUBJECT:

Forwards itemized r eview, of covlpliance w/significant< rules S.

regul ations< per 10CFR ini r esponse'o NRG 810929

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FLORIDA POWER & LIGHT COMPANY October 2, 1981 L-81-429 Office of Nuclear Reactor Regulation Attention: Nr. Darrell G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Hr. Eisenhut:

Re: St. Lucie Unit 2 Docket No. 50-389 Review of St. Lucie Unit 2 Against Significant Rules and Re ulations Per your request of'eptember 29, 1981, FPL is enclosing an itemized review of the compliance df St. I ucie Unit 2 with particularly significant rules and regulations of Title . 10 of the Code of Federal Regulations. Enclosure 8 of the January 8, 1981 t1emorandum 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 Florida Power 5 Light Company that St. Lucie Unit 2 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 Florida Power 8 Light, the NSSS vendor, and our consultants. Additional verification is

.provided the independent review of the NRC staff. All of these

.'factors, by

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taken together, provides reasonable assurance that the public health and safety will be protected.

If we can be of further service, please advise.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/TCG/ah Attachments cc: J P. O'Rei lly, Director, Region II (w/o attachments)

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Harold F. Reis, Esquire (w/o attachments)

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'07 2W 1 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. FP&L will adhere to those definitions in applicable documents.

19.4 This regulation governs the interpretation of regulations by the NRC and does not impose any specific requirements.

19.5 This regulation gives the address'f the NRC and does not impose any specific requirements'9.11 This regulation governs the posting of notices to workers. St. Lucie Unit 2 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. St. Lucie Unit 2 compliance with this regulation is ensured through Health Physics and administrative procedures 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. St. Lucie Unit 2 compliance with this regulation is ensured through appropriate Health Physics and administrative procedures.

19. 14 When required by and in accordance with this regulation, FP6L will allow workers to consult privately with an NRC inspector, and allow a worker representative to accompany the inspection.

19.15 This regulation governs NRC consultation with workers during inspection. St. Lucie Unit 2 will comply 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. St. Lucie Unit 2 complies with this regulation.

1072W 2 REGULATION COMPLIANCE 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 requirement s.

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. St. Lucie Unit 2 is in compliance with this regulation.

1072W 3 REGULATIONS COMPLIANCE 10CFR

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 Conformance 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 When required by and in accordance with this regulation, individuals will submit an appropriate written, signed statement. Appropriate health physics procedures and administrative procedures control this proces s.

(b) When required by this regulation, the accumulated dose for any individual permitted to exceed the exposure limits specified in 20.101(a) is determined by the use of Form NRC-4. Appropriate health physics procedures and administrative procedures control this process.

(FSAR Chapter 13.3)

20. 103(a) Compliance with this regulation is ensured through the implementation of appropriate health physics procedures relating to air sampling for radioactive materials, and bioassay of individuals for internal contamination 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, ll, the FSAR, are installed and operated to maintain levels and 12 of of airborne radioactivity as low as reasonably achievable'hen necessary, as determined by St. Lucie Unit 2 administrative and Health Physics guidelines, additional precautionary procedures are utilized to limit the potential for intake of radioactive materials.

20. 103(c ) The requirements of this regulation are ensured by the proper use of approved respiratory protection equipments St. Lucie Unit 2 administrative procedures

,a 1072' 4 REGULATION COMPLIANCE 10CFR 20.103(c) incorporate fully the stipulations of Regulatory (Cont'd) 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) St. Lucie Unit 2 will make the appropriate notification in accordance with this regulation.

20.103(f) The St ~ Lucie Unit 2 Respiratory Protection Program will in full accordance with the requirements of be 20.103(c) prior to start-up.

20. 104 Conformance with this regulation is assured by appropriate FP&L policies regarding employment of St.

Lucie Unit 2 individuals under the age of 18 and the Health Physics Manual restricting these individuals access to restricted areas.

20. 105 (a) Section 12.3 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 St. Lucie Unit 2 procedures, Technical and administrative policies which

'pecifications, 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 St.

Lucie Unit 2 procedures and applicable Technical Specifications which provide adequate sampling and analyses, and monitoring of radioactive materials in effluents before and during their release. Monitoring effluent releases is carried out in accordance with Regulatory Guide 1.21 and is discussed in Chapter ll of the FSAR. Appropriate surveys and monitoring devices document this compliance.

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P'072W 5 REGULATION COMPLIANCE 10CFR 20.106(b) FP&L does not propose limits higher than those spec-20.106(c) ified 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 ll 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 provisions 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.

'0. 108 This regulation provides an alert to the licensee that additional bioassay may be necessary or desirable in order to aid in determining the extent of an individual's exposure to concentrations of radioactive material. FP6L complies with this requirement.

20. 201 The surveys required by this regulation are performed at adequate frequencies and contain such detail as to be consistent with the radiation hazard being evaluated. St Lucie Unit 2 Health Physics Manual and

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applicable health physics procedures require these surveys and provide for their documentation in such a manner as to ensure compliance with the regulations of 10CFR Part 20.

, 20.202(a) The St. Lucie Unit 2 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 monitoring equipment.

REGULATION COMPLIANCE 10CFR 20.202(b) The terminology set forth in this regulation is accepted and conformed to in applicable St, Lucie Unit 2 procedures, Technical Specifications, and those portions of the St Lucie Unit 2 Health Physics Manual

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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 appropriate plant health physics procedures, as well as the St. Lucie Unit 2 Health Physics Manual.. The control and other protective measures set forth in the regulation are maintained under the surveillance of the Health Physics group ~

20. 203(d) Each Airborne Radioactivity Area, as defined 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 surveillance requirements necessary to determine airborne radioactivity levels.

20.203(e) The area and room posting requirements set forth in this regulation pertaining to radioactive materials are complied with through the implementation of appropriate health physics procedures, and portions of the St.

Lucie Unit 2 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 St. Lucie Unit 2 Health Physics Manual.

107 2W 7 REGULATION COMPLIANCE 10CFR 20.204 The posting requirement exceptions described in this regulation are used where appropriate and necessary at St. Lucie Unit 2. Adequate controls are provided within the St. Lucie Unit 2 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 St. Lucie Unit 2 Health Physics Manual and appropriate health physics procedures. These procedures also provide for the necessary documentation to ensure an auditable record of compliance.

20.206 The requirements of 10CFR19.12 referred to by this regulation are satisfied. Appropriate health physics procedures set forth requirements for all radiation workers to receive this instruction on a periodic basis.

20.207 The storage and control requirements for licensed materials in unrestricted areas are conformed to and documented through the implementation of St. Lucie Unit 2 health physics procedures and applicable portions of the St- Lucie Unit 2 Health Physics Manual.

20. 301 The general requirements for waste disposal set forth in this regulation are compiled with through appropriate administrative procedures. Chapter 11 of the FSAR describes the Radioactive Waste Management System installed at St. Lucie Unit 2.

20.302 No such application for proposed disposal procedures, as described in this regulation, has been made or is by FPGL.

20. 303 No plans for waste disposal by release into sanitary sewerage systems, as provided for in this regulation, are contemplate by St. Lucie Unit 2.

20.304 Disposal of wastes by burial in soil (i.e., onsite burial), as provided for in this regulation, is not being contemplated by St Lucie Unit 2.

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107 2W 8 REGULATION COMPLIANCE 10CFR

20. 305 Specific authorization, as described in this regulation, is not currently being sought by FP6L for treatment or disposal of wastes by incineration.

20.401 The requirements of this regulation are complied with through the health physics procedures pertaining to

, records of surveys, radiation monitoring and waste disposal. The retention periods specified for such records are also provided for in these specifications and procedure s.

20.402 St. Lucie Unit 2 will establish an appropriate inventory and control program to ensure strict accountability for all licensed radioactive materials.

Reports of theft or loss of licensed material are required by reference to the regulations of 10CFR. 20.403 Notifications of incidents, as described in this regulation, are assured by the St. Lucie Unit 2 Health Physics Manual and appropriate plant procedures, which also provide for the necessary assessments to determine the occurrence of such incidents.

20.405 Reports of overexposures to radiation and the occurrence of excessive levels and concentrations, as required by this regulation, are provided, for in appropriate health physics procedures.

20.407 The personnel monitoring report required by this regulation is provided for in the appropriate health physics procedures.

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 FP&L procedures.

20.409 The notification and reporting requirements of this regulation, and those referred to by it, are satisfied by the provisions of FPL6L procedures.

Ws 1072W 9 REGULATION COMPLIANCE 10CFR

21. 1 This regulation states the general purposes of this part and does not impose 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 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. St. Lucie Unit 2 is in compliance with this requirements

21. 7 This regulation provides for the granting of exemptions from 10CFR21 regulations, provided such exemptions are authorized by law and will not endanger life or property or the common defense and security, and are otherwise in the public interests It imposes no specific requirements.

21.21(a) This paragraph governs the evaluation of a deviation and the reporting of same to a responsible officer.

St. Lucie Unit 2 compliance is assured through appropriate Quality Assurance procedures.

21.21(b) This paragraph governs the reporting of a deviation to the NRC. St. Lucie Unit 2 compliance is ensured through appropriate Quality Assurance 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.

"r A! ~ 10 72W 10 I

RE GULATION COMPLIANCE 10CFR

21. 41 As required by this regulation, the licensee shall permit authorized representatives of the Commission to make inspections as allowed by this part.

21. 51 As required by this regulation, FP&L will maintain appropriate records as identified in this procedure.

Compliance is ensured through appropriate administrative and Quality Assurance procedure.

21. 61 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.

50. 36a Technical Specifications will be prepared which include items in each of the categories specified, including:

(1) safety limits and limiting safety settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

50. 36 The radiation Technical Specifications will include specifications which require compliance with 10CFR50.34a (releases as low as is reasonably achievable), and that ensure that concentrations of radioactive effluents released to unrestricted areas are within the limits specified in 10CF20.106 The reporting requirements of 10CFR 50.36a (a) (2) will also be included in these specifications.

50. 46 The St. Lucie Unit 2 Emergency Core Cooling System meets the requirements of this regulation (FSAR Sections 6.3) .

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.

1072W 11 RE GULATION COMPLIANCE 10CFR 50.55a(a) (1) In accordance with this paragraph, structures, systems, and components are designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with their safety importance. Applicable discussions are provided in the FSAR 50.55a (a) (2) This paragraph is a general statement leading into paragraphs (c) t hrough (i) of the regulation.

50.55a (b) This paragraph merely provides guidance concerning the approved Edition and Addenda of Section III and XI of the ASME Boiler and Pressure Vessel Code.

50.55a (c) These paragraphs delineate the codes and standards to 50.55a (d) which various components must adhere. St . Luc ie 50.55a (e) Unit 2 components subject to 50.55a meet the design and 50.55a (f) construction requirements as discussed in FSAR Section 5.2.4.

50.55a (g) This paragraph delineates the ASME Code preservice inspection requirements to which applicable components must adhere. As discussed in FSAR Subsection 5.2.4, St. Lucie Unit 2 meets these requirements.

50.55a (h) In accordance with this regulation, St. Lucie Unit 2 protection systems meet IEEE 279-1971

'n 50.55a (i) accordance with this regulation, Fracture Toughness requirements of Appendices G and H have been satisfied at St. Lucie Unit 2.

50.55a (j) This paragraph specifies exemptions for certain facilities. It is not applicable to St Lucie Unit 2.

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50. 70 This regulation discusses requirements for allowing plant access and providing adequate office facilities for an NRC inspector. FP&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.

50. 80 This regulation provides that the license may not be transferred without NRC consent. No application for a transfer of a license is involved in the St. Lucie Unit 2 proceeding ~

107 2W 12 REGULATION COMPLIANCE 10 CFR

50. 81 This regulation permits the creation of mortgages, pledges, and liens on licensed facilities, subject to certain provisions. FP&L licensed facilities are not mortgaged, pledged, or otherwise encumbered as those terms are used in this regulation.

50.82 This requirement provides for the termination of licenses. Should FP&L a request a license termination, FP&L will comply this regulation.

50. 109 This regulation specifies the conditions under which the NRC may require the backfitting of a facility.

FP&L will comply with this regulation.

10CFR50 Appendix B Chapter 17 of the FSAR describes 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 Section 13.3 describes the provisions of the emergency plans which ensure that St. Lucie Unit 2 meets the requirements of Appendix E.

10CFR50 Appendix G This Appendix specifies fracture toughness requirements for reactor coolant pressure boundary components.

Fracture toughness requirements of Appendix G have been satisfied. See FSAR Subsection 5.2.1 and 5.2.3 10CFR50 Appendix H This Appendix specifies reactor vessel material surveillance program requirements. A discussion of our compliance can be found in FSAR Subsection 5.3.1.5.

10CFR50 Appendix J This Appendix specifies containment leak rate testing requirements. Technical Specifications will be written to ensure compliance with this Appendix.

10CFR50 Appendix K This .Appendix specifies features of acceptable ECCS evaluation models. As discussed in FSAR Section 6.3 St. Lucie Unit 2 is in compliance with this Appendix.

REGULATION COMP LIANCE All structures, systems and components of the facility are classif1ed accord1ng to their relative important to safety.

Those items vital to safety such that their failure might cause or result in an uncontrolled release of an excess1ve amount of radioactive material are designated seismic Category I. They and items of lesser importance to safety are des1gned, fabricated, erected and tested according to the provisions of recognized codes and quality standards. Discussions of the applicable codes, standards, records and quality assurance program used to implement and audit the operation processes are presented in Section 17.2. A complete set of facility structural, arrangement and system draw1ngs is ma1ntained under the control of FP&L throughout the life of the plant ~ Quality assurance written data and comprehensive test and operating procedures are likewise assembled and maintained by FP&L The classification of seismic Category I structures and safety related systems and components is discussed in Section 3.2.

GDC 2 The structures, systems and components important to safety are designed to withstand the effects of natural phenomena without loss of capability to perform their safety functions. Natural phenomena factored into the design of plant structures, systems and components 1mportant to safety are determined from recorded data for the s1te vicinity with appropriate marg1n to account for uncertainties in historical data.

The most severe natural phenomena postulated to occur at the site in terms of induced stresses is the safe shutdown earthquake (SSE) ~ Those structures, systems, and components vital for the mit1gat1on and control of accident conditions are designed to withstand the effects of a loss of coolant accident (LOCA) coincident with the effects of the SSE. Structures, systems and components vital to the safe shutdown of the plant are designed to withstand the effects of any one of the most severe natural phenomena, including flooding, hurricanes, tornadoes and the SSE (refer to Chapter 2) ~

Design criteria for wind and tornado, flood and earthquake are discussed in Sections 3.3, 3.4 and 3.7, respectively.

GDC 3 Noncombustible and fire resistant materials are used wherever practical throughout the facility, particularly in areas containing critical portions of the plant such as containment structure, control room and components of systems important to safety. These systems are designed and located to minimize the effects of fires or explosions on their redundant components.

Fac1lities for the storage of combustible material are designed to minimize both the probability and the affects of a fire.

Equipment and facilities for fire 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 prov1ded.

1071W-2 REGULATION COMP LIANCE GDC 3 Normal fi.re protection is provided by deluge systems, hose lines (Cont') and portable extinguishers.

The Fire Protection System is designed such that a failure of any component of the system:

a) will not cause a significant release of radioactiv1ty to the environment b) will not impair the ability of redundant equipment to safely shutdown and isolate the reactor or limit the release of radioact1vity to the environment in the event of a LOCA.

All equipment is accessible for per1od1c inspection- The Fire Protection System is described in Subsection 9.5.1.

GDC 4 Structures, systems and components important to safety are designed to accommodate the effects of and to be compatible with the pressure, temperature, humidity, and radiation conditions associated with normal operation, maintenance, test1ng, and postulated accidents including a LOCA, in the area in which they are located.

Protect1ve walls and slabs, local missile shielding, or restraining devices are provided to protect the containment and engineered safety features systems within the containment against damage from missiles generated by equipment failures-The concrete enclosing the Reactor Coolant System serves as radiation shielding and an effective barrier against internally generated missiles. A missile shield is provided for control element drive mechanisms. Penetrations and piping extending outward from the containment, up to and 1ncluding isolation valves are protected from damage due to pipe whipping, and are protected from damage by external missiles, where such protection is necessary to meet the des1gn bases.

Non-seismic Category I piping is arranged or restrained so that failure of any non-seismic Category I piping will neither cause a nuclear accident nor prevent essential seismic Category I structures or equipment from mitigating the consequences of such an accident.

Seism1c Category I piping is arranged or restrained such that ina the event of rupture of a seismic Category I pipe which causes LOCA, resulting pipe movement will not result in loss of containment integrity or adequate engineered safety features systems operation.

The structures inside the containment vessel are designed to sustain dynamic loads wh1ch could result from failure of major equipment and pip1ng, such as jet thrust, jet impingement and local pressure transients, where containment integrity is needed to cope with the conditions.

1071W-3 A

REGULATION COMP LIANCE GDC 4 The external concrete Shield Building protects the steel (Cont') containment vessel from damage due to external missiles such as tornado propelled missiles.

For those components which are required to operate under extreme conditions such as design seismic loads or containment post-LOCA environmental conditions, the manufacturers submit type test, operational or calculational data which substantiate this capability of the equipment.

Refer to Sections 3.5, 3.6, 3.7 and 3.11 for details on missile protection, pipe rupture and jet impingement protection, seismic design criteria and environmental qualification, respectively.

GDC 5 Because St Lucie Unit 2 and Unit 1 share the plant site, they use the same cooling water and makeup water sources which are adequate for the requirements of both units in full operation.

Startup transformers may be paralleled under administrative control. No other structures, systems or components important to safety, are shared between St Lucie Unit 2 and Unit 1 except for seismic instrumentation and ultimate heat sink.

St Lucie Units 1 and 2 are designed using the "slide along" concept. The following facilities are shared by both nuclear units:

a) ultimate heat sink b) domestic water and fire protection system c) sw'itchyard, telemetering and load dispatch equipment d) seismic instrumentation e) site and offsite environmental monitors f) service building and guard houae g) steam generator blowdown treatment facility h) hypochlorite generator system i) makeup demineralizer regeneration (water treatment) system g) turbine oil storage tank k) auxiliary steam supply system

1) PAX communication system All facilities listed are constructed so that no single failure can in any way preclude safe shutdown of the plant.

1071WH REGULATION COMP LIANCE GDC 5 An accident in one unit does not affect safe shutdown of the (Cont') other unit. An accident in any of the shared features may result in reduced load operation of either or both units, but the capability for safe shutdown is unaffected by such an accident.

In the unlikely event of a loss of the preferred shutdown power, l

both St Lucie Units and 2 have their 'own 100 percent capacity redundant diesel generator sets which are ava1lable for safe shutdown.

The ultimate heat sink supplies emergency cooling water to both St Luc1e Units l and 2 The canal has sufficient cross-sectional flow area after liquefaction to mitigate the consequences of a LOCA on one unit while safely shutting down the other unit.

The systems and components which are interconnected between St.

l Lucie Units and Subsection l. 2.4.

2 but not normally shared, are discussed in GDC 10 In ANSl N18.2, plant conditions has been categorized in accordance with their anticipated frequency of occurrence and risk to the publ1c, and design requ1rements 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 requ1rement for Condition I is that marg1n shall be provided between any plant parameter and the value of that parameter which would require either automatic or manual protective action; it is met by providing an adequate control The des1gn requirement for system (refer to Section 7.7) ~

Condition II is that such faults shall be accommodated with, at most, a shutdown of the reactor, with the plant capable of return1ng to operat1on after correct1ve act1on; it is met by providing a Reactor Protective System (refer to Section 7.2) ~

Specified acceptable fuel design limits are stated in Section 4.4. Minimum margins to specified acceptable fuel design limits are prescribed in the Technical Specifications (Limiting Conditions for Operations) which support Chapters 4 and 15. The plant 1s designed such that operation within Limiting Conditions for Operation, with safety system settings not less conservative than Limitiag Safety System Settings prescribed in the Technical Specifications, assures that specified acceptable fuel design limits will not be violated as a result of anticipated operational occurrences. During non-accident conditions, operation of the plant within Limit1ng Conditions for Operation ensures that spec1fied acceptable fuel design 11mits are not approached within the minimum margins. Operator action, aided by the control systems and monitored by plant instrumentation, mainta1ns the plant within Limiting Conditions for Operat1on during non-accident conditions.

1071W-5 REGULATION COMP LIANCE GDC ll In the power operating range, the combined response of the fuel temperature coefficient, the moderator temperature coefficient, the moderator void coeff1cient, and the moderator pressure coefficient to an increase in reactor power in the power operation range is a decrease in reactivity, i.e , the inherent nuclear feedback characteristics are not positive. The reactivity coefficients are discussed in detail in Section 4.3 ~

GDC 12 Power level oscillations will not occur. The effect of the negative power coefficient of reactivity (refer to Criterion ll) ~ together with the coolant temperature soluble program maintained by boron, provide control element assemblies (CEAs) and 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 distribut1on oscillations are detected by neutron flux detectors. Axial mode oscillations are suppressed by means of CEAs. Rad1al oscillations are expected to be convergent It 1s a design objective that az1muthal xenon oscillations be convergent. Monitoring and protective requirements 1mposed by Criterion 10 and 20 are discussed in those responses and in Chapter 4.

GDC 13 Instrumentation is provided, as required, to monitor and maintain significant process variables which can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Controls are prov1ded for the purpose of maintaining these variables within the limits prescribed for safe operation.

N The principal var1ables and systems monitored include neutron level (reactor power); reactor coolant temperature, flow, and pressure; pressurizer liquid level; steam generator level and pressure; and containment pressure and temperature. In addition, instrumentation is provided for continuous automatic monitoring of process radiation level and boron concentration 1n the Reactor Coolant System.

The following is provided to monitor and maintain control over the fission process during both transient and steady state periods over the lifetime of the core:

a) Twelve independent channels of nuclear instrumentation, which constitute the primary monitor of the fission process. Of these channels, the four wide range logarithmic safety channels are used to monitor the reactor from startup through full power; four will monitor the reactor in the power range and are used to 1nitiate a reactor shutdown in the event of overpower; two linear power range channels are ut111zed for control purposes and two channels for startup and extended shutdown.

1071W-6 REGULATION COMP LIANCE GDC 13 b) Two independant CEA Position Indicating Systems (Cont')

c) A boronometer, which determines the boron concentration in the reactor coolant by neutron absorption is provided as a backup to the primary method of determining soluble po1son concentration by sampl1ng and analys1s of reactor coolant water (see Subsection 9.3.4).

d) Manual and automatic control of reactor power by means of CEAs e) Manual regulation of coolant boron concentrations.

Incore instrumentation is provided to supplement information on core power distribution and to provide for cal1bration of out-ofmore flux detectors.

Instrumentation measures temperatures, pressures, flows, and levels in the Main Steam System and auxiliary systems and is used to maintain these variables within prescribed limits.

The Reactor Protective System is designed to monitor the reactor operating conditions and to effect reliable and rapid reactor trip if any one or a combination of conditions deviate from a preselected operating range.

The containment pressure, temperature, and radiat1on instrumentat1on is designed to function during normal operation and the postulated accidents.

The instrumentation and control systems are described in detail in Chapter 7.

GDC 14 The RCPB is defined in accordance with 10CFR50, Section 50.2(v) and ANSI N18.2, Section 5.4.3.2 (see response to criterion 55) ~

Accordingly, Reactor Coolant System (RCS) components are des1gned to meet the requirements of the ASlE Code, Sect1on III. To establish operating pressure and temperature limitations dur1ng startup and shutdown of the Reactor Coolant System, the fracture toughness rules defined in Appendix G of the ASME Code,Section III, is followed. Quality control, inspection, and testing as required by this standard and allowable reactor pressure temperature operations ensure the integrity of the RCS.

The RCPB is designed to accommodate the system pressures and temperatures attained under all expected modes of unit operation, ixm.luding all anticipated transients, and to ma1ntain the stresses within applicable limits-

REGULATION COMP LIANCE GDC 14 Piping and equipment pressure parts of the RCPB are assembled (Cont') and erected by weld1ng unless applicable codes permit flanged, screwed, or compression joints. Welding procedures are employed which produce welds of complete fusion and free of unacceptable defects. All welding procedures, welders, and weld1ng mach1ne operators are qualified in accordance with the requi,rements of ASME Code,Section IX for the materials to be welded.

Qualification records, including the results of procedure and performance qual1fication tests and identificat1on symbols assigned to each welder are maintained.

The pressure boundary has provisions for inservice inspect1on in accordance with the requirements of ASME Code,Section XI, to ensure continuance of the structural and leaktight integrity of the'oundary (see also response to Criterion 32). For the reactor vessel, a material surveillance program conforming with tbe requirements of Appendix H of 10CFR50 is given in Section 5.3.

GDC 15 The design criteria and bases for the reactor coolant pressure boundary are described in the response to Criterion 14.

The operating conditions for normal steady state and transient plant operations are established conservat1vely (see Subsection 4.4.3). Limits are selected so that an adequate margin exists between normal operating and the design 11mits. The plant control systems are designed to ensure that plant variables are maintained well within the established operating limits. The plant transient response character1stics and pressure and temperature distr1butions during normal operations are considered in the design as well as the accuracy and response of the 1nstruments and controls. These design techniques ensure that a satisfactory margin is maintained between the plant' normal operating conditions, including design transients, and the design limits for the reactor coolant pressure boundary'he RPS functions to minimize dev1ation from normal operating limits in the event of anticipated operational occurrences (ANSI N18.2 Condition II Occurrences) ~

Analyses show that the design limits for the reactor coolant pressure boundary are not exceeded in the event of any ANSI N18.2 Condition II, of Reactor Coolant Pressure Boundary, and in Chapter 7, Instrumentation and Controls.

GDC 16 The containment system is designed to protect the publ1c from the rad1ological consequences of a LOCA, based on a postulated break of reactor coolant piping up to and including a double ended break of the largest reactor coolant pipe-

4 a.L C

1071W-8 REGULATION COMP LIANCE GDC 16 The containment vessel, Shield Building, and the associated (Cont') engineered safety features systems are designed to safely sustain all internal and external environmental conditions that may reasonably be expected to occur during the life of the plant, including both short and long term effects of a design basis accident.

Leak tightness of the containment system and short and long term performance are analyzed 1n Section 6.2.

GDC 17 Offsite power is transmitted to the plant switchyard by three physically 1ndependent 240 kV tranmiss1on lines. During normal plant operation, the station auxiliary power ia normally supplied from the main generator through the plant auxiliary transformers. Upon loss of power from the auxil1ary transformers or of a unit generator, there is a "fast, dead" automatic transfer to the startup transformers thus providing continuity of power.

In the event of a loss of the offsite power sources, two emergency onsite diesel generator sets and redundant sets of station batteries provide the necessary ac and dc power for safe shutdown or, in the event of an acc1dent, prov1de the necessary power to restrict the consequences to within acceptable limits.

The onsite emergency ac and dc power systems consist of redundant and independent power sources and distribution systems such that a single failure does not prevent the systems from performing their safety function.

Refer to Sections 8.2 and 8.3 for further discuss1on of offsite power sources and onsite power sources respectively.

GDC 18 Electrical power systems important to safety are designed to permit appropriate periodic inspection and testing of important areas and features such as wiring, insulations connections, and switchboards to assess the continu1ty of the systems and to detect deterioration, if any of their components. Capability is provided to periodically test the operability and functional performance of the components of the systems. The diesel generator sets are started and loaded periodically on a routine basis and relays, sw1tches, 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 operat1on.

Refer to Section 8.3 and the Technical Specif ications GDC 19 The control stations, switches, controllers and indicators necessary to operate or shut down the unit and maintain safe control of the facility are located in the control room.

1071W-9 REGULATION COMP LIANCE GDC 19 The design of the control room permits safe occupancy during (Cont') design basis accident conditions. The control room is isolated from the outside atmosphere during the initial period following the occurrence of an accident. The control room ventilation system recirculates control room air through HEPA and charcoal filters as discussed in Subsection 9.4.1 and Section 6.4 ~

Radiation detectors and alarms are provided. Emergency lighting 18 provided as discussed in Subsection 9.5.3 ~

Alternate local controls and local instruments are available for equipment required to bring the plant to and maintain a hot standby condition. It is also possible to attain a cold shutdown condition from locations outside of the control room through the use of suitable procedures. (Refer to Subsection 7.4. l) ~

GDC 20 The Reactor Protective System mon1tors reactor operating conditions and automatically initiates a reactor trip when the monitored variable or combination of variables exceeds a prescribed operating range. The reactor tr1p setpoints are selected to ensure that anticipated operational occurrences do not cause specified acceptable fuel design limits to be v1olated. Specific reactor trips are described in Sect1on 7.2.

Reactor trip is accomplished by deenergizing the control element drive mechanism holding latch coils through the interruption of the CEDM power supply. The CEAs are thus released to drop 1nto the core reducing reactor power.

The Eng1neered Safety Features Actuation System monitors potential accident conditions and automatically initiates eng1neered safety features and their supporting systems when the monitored variables reach prescribed setpoints. The parameters which automatically actuate engineered safety features are described 1n Section 7.3 ~ Manual actuation is prov1ded to the operator.

GDC 21 The protection systems are des1gned to prov1de high functional reliability and in"service testability by designing to the requirements of IEEE 279 1971 and IEEE 338 1971. The systems are designed such that the s1ngle failure criteria and performance requirements are met with three channels in service A coincidence of exceeding any two like sensor trip parameters generates a trip signal. However, four measurement channels with electrical and phys1cal separation are provided for each parameter. To enhance plant ava1lability, a fourth channel is provided aa a spare and allows bypassing of one channel while mainta1ning the requisite twomut-of-three system

1071W-10

~t)

REGULATION COMP LIANCE GDC 21 Each channel of the protection system, including the sensors up (Cont') to the final actuation device is capable of being checked during reactor operat1on. Those channels that can affect plant operation are tested during scheduled reactor shutdown.

Measurement sensors of each channel used in protection systems are checked by observing outputs of similar channels which are presented on indicators and recorders in the control room. Trip un1ts and logic are tested by inserting a signal into the measurement channel ahead of the readout and, upon application of a trip level input, observing that a signal is passed through the trip units and the logic to the logic output relays. The logic output relays are tested individually for init1ation of trip action.

Protection system reliability and testability are discussed in Sections 7.2 and 7.3.

GDC 22 The protection systems conform to the prov1sions of IEEE 279 1971, as explained in Subsection 7.2.2.3 ~ Four independent measurement channels complete with sensors, sensor power supplies, signal conditioning units and bistable trip units are provided for each protective parameter monitored by the protection systems. The measurement channels are provided with a high degree of independence by 'separate connections of the channel serisors to the process systems. Power to the channels is provided by two independent emergency power supply sources.

The protective system is functionally tested to ensure satisfactory operation pr1or to installation in the plant.

Environmental and seismic qual1fications are also performed utilizing type tests and specific equipment tests (refer to Sections 3 '0 and 3.11)- t GDC 23 Protective system 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 gravity insertion of the CEAs into the core. Redundancy, channel 1ndependence, and separation 1ncorporated in the protective system des1gn minimize the possibility of the loss of a protection function under adverse environmental conditions (refer to Sections 7.2 and 7.3) ~

GDC 24 The protect1on systems are separated from the control instrumentation systems so that failure or removal from service of any control 1nstrumentation system component or channel does not inhibit the function of>the protect1on system. Separation of protection and control systems is discussed in Subsection 7.2.2.3.

107 1W-ll REGULATION COMP LIANCE GDC 25 Reactor shutdown with CEAs is accompl1shed completely independent of the control functions since the trip breakers interrupt power to the CEA drive mechanisms regardless of ex1sting control signals. The design is such that the system can withstand accidental withdrawal of controll1ng groups without exceeding acceptable fuel design limits. Analysis of possible reactivity control malfunctions is given in Section 15.4. The Reactor Protection System will prevent specified acceptable fuel design 11mits from being exceeded for any event of moderate frequency.

GDC 26 Two independent reactivity control systems of different design principles are provided- The first system, using CEAs, includes a pos1tive means (gravity) for inserting CEAs and is capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including specified anticipated operational occurrences, specified acceptable fuel design limits are not exceeded. The CEAs can also be mechanically driven into the core. The appropriate margin for stuck rods 1s provided by assuming in the analyses of anticipated operational occurrences that the highest worth CEA does not fall intothe core.

The second system, the Chem1cal and Volume Control System (CVCS), us1ng neutron absorbing soluble boron, 1s capable of reliably compensat1ng for the rate of reactivity changes resulting from planned normal power changes (1ncluding xenon burnout) such that acceptable fuel design limits are not exceeded. This system is capable of holding the reactor subcritical under cold conditions.. For a further description, see Subsection 9.3.4.

Either system is capable of making the core subcrit1cal from a hot operating condition.

For further discussion, see Sections 7 4 and 7.7.

GDC 27 The reactivity control systems that prov1de the means for making and holding the core subcr1tical under postulated accident conditions are discussed in Sections 9.3 and 4.3. Combined use of CEAs and chemical shim control by the CVCS provides the shutdown margin required for plant cooldown and long-term xenon decay, assuming the highest worth CEA 1s stuck out of the core.

During an acc1dent, the Safety Injection System (SIS) funct1ons to inject concentrated boric acid into the Reactor Coolant System for short term and long term cooling and for reactivity control. Details of the system are given in Section 6.3.

The Safety Injection System, in conjunction with the combined capabilities of the reactivity control systems is available to maintain short and long term cooling of the core even in the event a CEA of highest worth is stuck out of the core. Upon receipt of a safety injection actuation signal, the SIS functions to infect borated water from the refueling water tank into the Reactor Coolant System.

107 1W-12 REGULATION . COMP LIANCE 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 d1vided 1nto three sets: a shutdown set, a regulating set, and a part length set. These sets are further subdivided into groups as necessary. Administrative procedures and control interlocks ensure that the amount and rate of reactivity increase is limited to predeterm1ned 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 (see Sections 4.3 and 7.7) ~

The maximum rate of reactivity addition that may be produced by the CVCS is too low to induce any s1gnif1cant pressure forces that might rupture the RCPB or d1sturb the reactor vessel internals.

The RCPB (Chapter 5) and the reactor internals (Chapter 4) are designed to appropriate codes (refer for instance to the response to Criterion 14) and 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 limited deformation which will not impair the capability of cooling the core.

GDC 29 Plant condit1ons designated as Condition I and Condition II in ANSI N 18.2 are carefully considered in the design of the Reactor Protective System and the reactivity control systems.

Consideration of redundancy, independence and testability in the design, coupled with careful component selection, overall system testing, and adherence to detailed quality assurance, assure an extremely high probability that safety functions are accomplished in the event of anticipated operational occurrences. For additional discussion, see the responses for Criteria 10, 13, 15 and 20.

GDC 30 The RCPB components are designed, fabricated, erected, and tested in accordance w1th ASME Code, Section III- RCPB components are classified as Quality Groups A and B as defined in Subsection 3.2.2. Accordingly, they receive all of the quality measures appropriate to that classification.

Detection and ident1fication of reactor coolant leakage is discussed in Subsection 5.2.5. The system is designed to detect and 1dentify the source of reactor coolant leakage.

1071W-13 REGULATION COMP LIANCE GDC 31 All th'e RCPB components are des1gned and constructed in accordance with ASME Code Section III, and comply with the test and 1nspection requirementa of this code. These test and inspection requirements assure that flaw sizes are limited so that the probability of failure by rap1d propagation is extremely remote. Particular emphasis is placed on the quality control applied to the reactor vessel, on which tests and 1nspections exceed1ng ASME Code requirements are performed. The tests and inspections performed on the reactor vessel are summarized in Sections 5.2 and 5.3 ~

Carbon and low-alloy steel materials that form part of the pressure boundary are tested for fracture toughness to the acceptance criteria and requirements of 10CFR50, Appendix G (see Section 5.2) ~

GDC 32 Provisions are made in the des1gn for inspection, testing and surveillance of the RCS boundary as required by ASME Code Section III and Section XI, as applicable.

The reactor vessel surveillance program conforms with ASTM E-185, "Standard Recommended Practice for Surve1llance Tests for Nuclear Reactor Vessels", as revised in 1973 ~ The details of the reactor surveillance program are given in Subsection 5.2 4.

GDC 33 Reactor coolant makeup during normal operation is provided by the CVCS. The design incorporates a high degree of funct1onal reliability by provision of redundant components and an alternate path for charging ~ The charging pumps can be powered from either onsite or offsite power sources, including the onsite emergency diesel generators. The system is described in Subsection 9.3.4 GDC 34 The transfer of fission product decay heat and other residual heat from the reactor core is accomplished by the steam generators and the shutdown cooling system at such a rate that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Residual heat removal capacity is provided with sufficient redundancies in design that in the event of a s1ngle active failure or a s1ngle limited leakage passive failure the system can still perform its function. The steam generator auxiliaries and the Shutdown Cooling System are des1gned to operate from either offsite or onsite electric power sources. See Subsection 5.4.7 for addit1onal information on Residual Heat Removal.

GDC 35 The Safety Injection System provides cooling water at a rate sufficient (l) to assure that the zirconium~ater reaction ia limited to a negligible rate of less than one percent and (2) to assure that the fuel remains in a eoolable geometry. Therefore, compliance with the 1ntent of the acceptance criteria for emergency core cooling systems for light water power reactors of 10CFR50, paragraph 50.46(b) is satisfied.

1071W-14 REGULATION COMP LIANCE GDC 35 The SIS design includes adequate prov1sions to ensure that the (Cont') required safety functions are provided with a single, active failure relying on either onsite or offsite electrical power supply (see Section 6.3)-

GDC 36 The Safety Injection System is designed to facilitate inspection of all critical components. Those components located external to the containment structure are readily accessible for periodic inspect1on to ensure system leak-tight integrity. Components located inside containment are designed to permit inspection for leak-tightness during ma1ntenance and refueling shutdowns.

Reactor vessel internal structures, reactor coolant piping and water injection nozzles are designed to permit visual inspection and/or nondestructive inspection techniques (where these are applicable).

The actual location, arrangement and installation of the system components provide the necessary access for the capability of complying with the periodic inspection requ1rements of Section XI of the ASME Code (refer to Section 6.3 and the Technical Specifications).

GDC 37 The Safety Injection System is designed to permit appropriate periodic pressure and functional testing. The structural and leaktight integrity, operability, and performance of the SIS components and system are assured through testing conducted during normal plant operation under conditions as close to design as practicable. The operational sequence that bring the SIS into action, 1ncluding transfer to alternate power sources, is designed to be tested in parts in such a manner as to verify the operability of the actuation system as a whole-Periodic pressure testing of the high pressure safety injection portion of tha SIS to assure system integrity is possible using the cross connection from the charging pumps in the CVCS. Flow path continuity in the high pressure injection lines and suction lines from the refuel1ng water tank (RWT) is assured with the plant at operating pressure by the recirculation of HPSI and LPSI pump discharge back to the RWT. Since LPSI pumps are used as shutdown cooling pumps during normal operation their operability is further demonstrated. Borated water from the safety injection tanks (SITs) may be bled through the recirculation test lines to verify flow path continuity from each tank to its associated main safety injection header.

During refueling, blowdown tests will provide additional evidence of SIT operability. Preoperational testing of the SIS provides additional assurance of SIS performance (see Section 6.3 and the Technical Specifications) ~

107 1W-15

~ ~

0'EGULATION COMP LIANCE GDC 38 The conta1nment heat removal systems described 1n Subsection 6.2.2 consists of the Containment Spray System and the Containment Cooling System. The Containment Spray System consists of two trains, each containing a containment spray pump, shutdown heat exchanger and spray header. The Containment Cooling System consists of four fan coolers. One spray pump and two containment fan coolers have the capacity to reduce containment pressure and temperature following a design basis accident and maintain them at acceptably low levels.

Both the Containment Spray and the Containment Cooling Systems are provided with emergency onsite power necessary for their operation, assuming a loss of offsite power. They are provided with offsite power from the startup transformers if normal onsite power is not

-ava1lable. The Containment Spray and Containment Cooling Systems are provided with redundant equipment so that when assuming a single failure or a failure of an emergency ons1te power supply, 100 percent conta1nment cooling capability is available.

GDC 39 The Containment Spray System essential equ1pment except for risers, distribution header piping, spray nozzles and the containment sump are located outside of the containment. The containment sump, the spray piping, and the spray nozzles within the containment can be inspected during refueling shutdowns. Associated equipment outside the containment can be visually inspected at any time.

The Containment Cooling System is entirely within the containment.

It can be 1nspected at appropriate intervals during refueling shutdowns. Cooling water systems external to the conta1nment which service the Containment Cooling System are accessible for inspect1on at any time during plant operation.

Inserv1ce inspections of the Containment Spray System and the Containment Cooling System are performed as indicated 1n Section

6. 6.

GDC 40 System piping, valves, pumps, fans, heat exchangers, and other components of the containment heat removal system are designed to permit appropriate periodic testing to assure their structural and leaktight integrity. The components are arranged so that each component can be tested periodically for operability and required funct1onal performance.

The containment cooling units are normally in operation.

The operational sequence that would bring the containment heat removal system 1nto action, including the transfer to alternate power sources, can be tested. With the plant at operating pressure, the containment spray pumps and valves may be operated by recirculation back to the refueling water tank. This permits verification of flow path continuity in the suction lines from the refueling water tank to the first containment spray isolation valve outside the containment. The spray isolation valves can be tested independently of the spray pumps (refer to Section 6.2 and the Technical Specifications).

C 1071M-16 REGULATION COMP LIANCE GDC 41 As discussed in Subsection 6.5.2, the Containment Spray System, in conjunction with the Iodine Removal System, provides a function of remov1ng fission products from a post-accident containment atmosphere. The Iodine Removal System removes radio-iodines from the containment atmosphere following a LOCA by adding controlled amounts of hydrazine to the containment spray water.

The Shield Building Ventilation System consists of two full capacity redundant fan and filter systems and is designed, consistent with the functioning of other engineered safety features systems, to reduce the concentration and quantity of fission products released to the environment follow1ng a LOCA by establishing and maintaining a subatmospheric pressure within tha Shield Building annulus to ensure that post-accident activity leakage from the containment vessel is routed through the charcoal filter system (refer to Subsection 6.2.3) ~

Hydrogen control and sampl1ng systems are provided to prevent the buildup of dangerous concentratio of hydrogen in the

~

containment following a LOCA. The hydrogen control system consist of two full capacity hydrogen recombiners and redundant hydrogen sampling systems. A Continuous Containment/Hydrogen Purge System is also available. The hydrogen recombiners, which are the primary means of control, provide control of hydrogen concentration in the conta1nment without any release to the environment. The hydrogen sampling system can analyze the containment atmosphere either by passing a sample through the automatic hydrogen analyzer or by util1zing a grab sample (refer to Subsections 6.2.3, 6.2.5 and 9.4 8) ~

The Shield Building Ventilation System, Containment Spray System/Iodine Removal System, and the containment hydrogen control system have suitable redundancy to assure that for normal onsite or for offsite electrical power system failure, their safety functions can be accomplished, assuming a single failure.

GDC 42 All components of the Shield Building Ventilation System and the containment hydrogen control system are accessible for physical inspection. Ducts, plenums, and casings are provided with access doors for internal 1nspection.

The only components of the containmant atmosphere cleanup systems inside the Shield Bu1lding are the duct work of the SBVS, hydrogen recombiners and the containment spray nozzles and piping. These can be inspected dur1ng shutdown.

Specific inspection programs are d1scussed in Subsection 6.2.5.4 for the combustible gas control systems and components, Subsection 6.5.1.4 for the filter systems that are required to perform a safety related function following a design basis accident and Subsection 6.5-2.4 for the Containment Spray System.

~ 107 1W-17 1

REGULATION COMP LIANCE GDC 43 The Shield Build1ng Ventilation system, Containment Spray r system, and hydrogen control and sampling systems are designed and constructed to permit periodic pressure and funct1onal testing. For the purpose of periodically test1ng the retentive capability of the filter systems, test panels are placed 1n, the filter housings 1n locations which allow the panels to be subjected to the same air flow as the filters. These are periodically removed and tested.

High efficiency particulate (HEPA) and charcoal filters are located outside the containment for convenience in testing and inspection. Period1c tests are described in Subsection 6.5.1.4.

Active components of the Shield Building Ventilation System, hydrogen analyzer, hydrogen recombiners and Containment Spray System can be tested periodically for operability and required functional performance.

The full operational sequence that would bring the systems into action, includ1ng the transfer to alternate po~er sources, and the design air flow capability can be tested (refer to Subsections 6.2.5.4, 6.5.1.4 and 6.5.2.4) ~

GDC 44 The cooling water systems which function to remove the combined heat load from structures, systems and components important to safety under normal operating and accident conditions, are the Component Cooling Water System and the Intake Cooling Water System.

The Component Cooling Water System is a closed loop system which removes heat from the shutdown heat exchangers, Containment Cooling System and other essential and nonessential components as described in Subsection 9.2.2.

The Intake Cooling Water System is an open loop system. which removes heat from the Component Cool1ng System and transfers it to the ultimate heat sink as described in Subsection 9.2.1.

The primary and secondary sources of water for the utimate heat sink are as follows. The intake cooling water pumps normally take water from the Atlantic Ocean through the circulating water intake condu1ts and canal. In the event of interruption of water from this source, water is taken through the emergency cooling water canal from Big Mud Creek. The ultimate heat sink is discussed in Subsection 9.2.5.

Each system is normally pressurized permitting leakage detection by routine surveillance or monitoring instrumentation.

Electrical power for the operation of each system may be supplied from offsite or onsite emergency power sources, with distribution arranged such that a single fa1lure does not prevent the system from performing its safety function.

REGULATION COMP LIANCE GDC 45 The Component Cooling Water Systam and Intake Cooling Water System are des1gned to permit periodic inspection, to the extent practical of important components,. such as heat exchangers, pumps, valves and accessible pip1ng. Each system is normally pressurized permitt1ng leakage detection by routine surve1llance or monitoring instrumentation (refer to Subsections 9.2.1-4 and 9.2.2.4 and Sect1on 6.6) ~

GDC 46 Both the Component Cooling Water and Intake Cooling Water Systems are in operation during normal plant operat1on or shutdown. The structural and leaktight integrity of the Component Cooling Water and Intake Cooling Water Systems components and the operability and performance of their active components are demonstrated in this way. The operation of pumps and heat exchangers are rotated on a scheduled basis to monitor operat1onal capability of redundant components. Data can be taken periodically during normal plant operation to confirm heat transfer capabilities (refer to Subsections 9.2.1.4 and 9.2.2.4) ~

The systems are designed to permit testing of system operability encompassing s1mulation of emergency reactor shutdown or LOCA cond1tions, including the transfer between normal and emergency power sources.

GDC 50 The containment structure, including access openings and penetrations, is designed to accommodate, without exceeding the design leak rate, the transient peak pressure and temperature associated with a design basis accident.

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 Safety Injection System ia designed such that no single active failure could result in significant metal~ater reaction. The combined cooling capacity of two containment cooling un1ts and one containment spray train is adequate to prevent over pressurization of the structure, and to return the containment to near atmospheric pressure (refer to Subsection 6.2. l).

GDC 51 As specified in Subsection 3.8.2, the material selected for the containment vessel is carbon steel normalized to refine the grain which results 1n improved ductility. In addition, the actual mechanical and chemical properties of the material are documentad and are within the limits for minimum ductility.

The containment vessel is built to Subsection NE of Section III of the ASME Code, and 1n accordance with this code the materials including weld specimens are 1mpact tested.

The design of the vessel reflects consideration of all ranges of temperature are loading conditions which apply to the vessel during operation, maintenance, testing and postulatad acc1dent conditions.

1071W-19 r ~ ~

REGULATION COMP LIANCE GDC 51 All seam welds in the vessel are 100 percent radiographed and (Cont') the acceptance standards of the radiographs ensure that flaws in welds do not exceed the maximim allowed by the ASME Code.

Since 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 (refer to Subsection 3.8.2).

GDC 52 The containment vessel is designed so that initial integrated leak rate testing can be performed at design pressure after completion and installation 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 of the containment.

Periodic integrated leakage rate testing are carried out in accordance with the requirements of Appendix J to 10CFR50. A description of the periodic integrated leakage rate testing is provided in Subsections 6.2.1.6 and 6.2.6.

GDC 53 The absence of insulation of the containment vessel permits periodic inspection of the exposed interior surfaces of the vessel. The lower portions of the containment vessel are totally encased in concrete and are not accessible for inspection after the acceptance testing. There is no need for any special in"service surveillance program due to the rigorous design, fabrication, inspection and pressure testing the containment vessel receives prior to operation.

Provisions are made to permit periodic testing at containment design pressure of penetrations which have resilient seals or expansion bellows to allow leak tightness to be demonstrated (refer to 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 isolation system design requires a double barrier on all of the above systems not serving accident consequence limiting systems so that no single active failure can result in loss of isolation or intolerable leakage-These lines are provided with isolation valves as indicated in Subsection 6.2.4.

Valves isolating penetrations serving engineered safety features systems will not automatically close with a containment isolation actuation signal (CIAS), but may be closed'y remote manual operation from the control room to isolate any Engineered Safety Feature when required.

v 1071W-20 REGULATION COMP LIANCE GDC 54 Proper valve closing time is achieved by appropriate selection (Cont') of valve, operator type and operator size. Refer to Subsection 6.2.4 for additional isolation valve information.

To ensure continued integrity of the containment isolation system, periodic closure and leakage tests are performed as stated in Subsection 6. 2.4.4.

GDC 55 Except for Safety Infection System lines, shutdown cooling lines, certain sample lines off the Safety Injection System or Reactor Coolant System, and Chemical and Volume Control System charging and letdown lines, the reactor coolant pressure boundary as defined in 10CFR50 is located within the containment. Isolation provisions for these lines are as indicated in Subsection 6.2.4 The safety injection, shutdown cooling and charging lines are closed seismic Category I piping systems outside the containment with isolation valves as indicated in Subsection 6.2.4.

GDC 56 Lines which connect directly to the containment atmosphere and are not used to mitigate the effects of a LOCA are provided with two valves in series, one inside and one outside the containment. These containment isolation valves are either capable of automatic actuation or normally locked closed.

Lines which connect directly to the containment atmosphere and are used for mitigating the affects of a LOCA are provided with a double containment barrier which consists of the closed piping system pressure boundary outside the containment and one isolation valve capable of remote manual actuation-Automatic isolation valves, upon loss of power, are selected to failmlose, fail-as-is, or failmpen, whichever position provides the greater safety. Isolation valves are located as close to the containment as practical. Refer to Subsection 6.2.4 for detailed information regarding containment isolation.

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 which are either automatic, or locked closed, or capable of remote manual operation, and located outside the containment as close to tha containment as practical (refer to Subsection 6.2.4).

GDC 60 The Waste Management System is described in Sections 11.2, 11.3'nd 11.4, and is designed to provide controlled handling and disposal of liquid, gaseous, and solid wastes. The Waste Management System is designed to ensure that the general public and plant personnel are protected against exposure to radioactive material to meet the intent of 10 CFR 20 and 10 CFR 50, Appendix I.

REGULATION COMP LIANCE GDC 60 'All liquid and gaseous radioactive releases from the Waste (Cont') Management System are accomplished on a bath basis. All radioactive effluents are sampled prior to release to ensure c'ompliance with 10 CFR 20 and 10,CFR 50, Appendix I and to determine release rates. Radioactive effluents which do not meet release limits are not discharged to the environment. The Waste Management System is designed with sufficient holdup capacity and flexibility for reprocessing of wastes to ensure that releases are as low as reasonably achievable.

The Waste Management System is designed to preclude the inadvertent release of radioactive material.

All storage tanks in the liquid waste and gaseous waste systems are administratively controlled to prevent the addition of waste to a tank which is being discharged to the environment. Each discharge path ia provided with a radiation monitor which alerts plant personnel and initiates automatic closure of redundant isolation valves to prevent further releases in the event of noncompliance with 10 CFR 20 (see Section 11. 5 for details).

GDC 61 Most of the components and systems in this category are in frequent use and no special testing is required. Those systems and components important to safety which are not normally operating are tested periodically, e.g., temperature alarms in the Fuel Pool System (Subsection 9.1.3) and radiation alarms in the fuel pool area, and the fuel handling equipment (prior to each refueling) ~

The spent fuel storage racks are located to provide sufficient shielding water 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 10CFR 20.

The Waste Management System (Chapter 11) is designed to permit controlled handling and disposal of liquid, gaseous, and solid wastes which will be generated during plant operation. The principal design criterion is to ensure that plant personnel and the general public are protected against exposure to radiation from wastes in accordance with limits defined in 10CFR20.

The fuel pool is located within the Fuel Handling Building. The liquid waste processing equipment and the gaseous waste storage and disposal equipment are located within a separate area of the Reactor Auxiliary Buildings Both of these areas provide confinement capability in the event of an accidental release of radioactive materials, and both are ventilated with filtered discharges to the vent pipe which is moni.tored.

107 1W-22 REGULATION COMP LIANCE GDC 61 Analysis (Section 15.7) indicate that the accident release of (Cont') the maximum activity content of a gas decay tank does not result in doses in excess of the limits set forth in 10CFR100.

The Fuel Pool Cooling System is designed to prevent damage to the spent fuel which could result in radioactivity release to the plant operating areas or the public environs (Subsection 9 1.3) ~

The fuel pool is designed to withstand the postulated tornado driven missiles, seismic event or cask drop without loss of pool water.

GDC 62 The new and spent fuel storage and handling facilities are described in Subsections 9.1.1 and 9.1.2. New fuel is stored in air. Spent fuel is stored in borated water. The spacing is sufficient to maintain a subcritical keff for the new and spent fuel assemblies when in unborated water.

GDC 63 There are no residual or decay heat removal systems in the Waste Management System The Fuel Pool and Waste Management Systems are provided with appropriate radiation indication and alarms. In addition, alarms are provided in the event of a reduction in fuel pool level.

GDC 64 Means are provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss of coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents.

Radioactive waste management and monitoring is discussed in Chapter ll Area and airborne monitoring is discussed in Subsection 12.3.4.

107 3W-1 REGULATION COMP L IANCE 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 St Lucie Unit 2 compliance.

55.4 The definitions contained in this part will be adhered to in all applicable St Lucie Unit 2 documents.

55.5 This regulation gives the address of the NRC and does not impose any specific requirements.

55.6 This regulation governs the interpretation of by the NRC and does not impose any 'egulations 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 St Lucie Unit 2 personnel will be in accordance with this regulation.

55.11 This regulation states the requirements for NRC approval of a license application. It does not impose any specific requirements.

107 3W-2 REGULATION COMP L IANCE 55.12 This regulation provides the terms on which a license applicant whose application has been denied may submit a reapplication. 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 requirements.

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 requirements.

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 does not impose any specific requirements.

55.25 This regulation permits the NRC to administer 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.

55. 32 This regulation states that each operator and senior operator license shall expire two years after the date of issuance. It does not impose any specific requirements.

1073W-3 REGULATION COMPLIANCE 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. FP&L will comply with this regulation.

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 regulation requires the licensee to notify'he NRC of any disability referred to in 55.11 of this part. St Lucie Unit 2 compliance is ensured through appropriate administrative 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 requirements.

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. FP6L will comply with this regulation.

55 Appendix A In accordance with this regulation, an operator and senior operator requalification program will be instituted at St Lucie Unit

2. This program will be in accordance with the requirements specified in this Appendix.

71.31 This regulation defines general standards for packaging of radioactive material. St Lucie Unit 2 will employ only packaging which conforms to these standards. Compliance 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 St Lucie Unit 2 will conform to these standards. Compliance is assured through appropriate administrative procedures'

1073M-4 REGULATION COMPLIANCE 71.33 This regulation governs the criticality standards used in the design of fissile material packages. Fissile material packaging used at St Lucie Unit 2 will conform to these standards. FPSL will comply with this regulation.

71. 34 This regulation defines the general conditions used to determine the effect of transport environment on a radioactive material package. Radioactive material packages used at St Lucie Unit 2 will be only those tested in accordance with this regulation. FPSL will comply with this regulation.

71.35 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. FP6L will comply with this regulation.

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. FPSL will comply with this regulation.

71.37 In accordance with this regulation, arrays of fissile material packaging used at St Lucie Unit 2 will be evaluated according to the standard set forth in 71.38, 71.39, and 71.40 of this part. FP&L will comply with this regulation.

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. FPGL will comply with this regulation.

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. FP6L will comply with this regulation.

T s ~ 1 I 1073M-5 r i REGULATION COMPLIANCE 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. FP&L will comply with this regulation.

71.41 This regulation permits the use of packages constructed and authorised for use prior to January 1, 1967, subject to certain conditions. Should such a package be employed at St Lucie Unit 2, this regulation will be complied with.

\

71.42 This regulation defines special requirements for shipments of plutonium. FPSL 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.

1073W-6 REGULATION COMPLIANCE 100. 1 This regulation states the general purposes of Part 100 and does not impose any specific requirements.

100 ~ 2 This regulation establishes the applicability of Part 100 and does not impose any specific requirements.

100.3 The definitions contained in this regulation will be adhered to in applicable St Lucie Unit 2 documents.

100.10 As required by this regulation, site specifics, 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 site 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 center distance have been established and are described in FSAR Section 2.1.

(b) This paragraph applies to sites for multiple reactor facilities. St Lucie Unit 2 and St Lucie Unit 1 are independent to the extent that an accident in one reactor would not initiate an accident in another.

100 Appendix A This Appendix provides geologic and seismic siting criteria. The geologic and seismic characteristics of the St Lucie Unit 2 site are described in FSAR Chapter 2. St Lucie Unit 2 design is in compliance with this regulation as discussed in FSAR Chapter 3.

107 3W-1 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 St Lucie Unit 2 compliance-55.4 The definitions contained in this part will be adhered to in applicable to St. Lucie Unit 2 documents.

55.5 This regulation gives the address of the NRC and does not impose any specific requirements.

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 St Lucie Unit 2 personnel will be in accordance with this regulation.

55.11 This regulation states the requirements for NRC approval of a license application. FP&L will be in compliance with this requirement.

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107 3W-2 REGULATION COMPLIANCE 55.12 This regulation provides the terms on which a license applicant whose application has been denied may submit a reapplication. FPGL will be in compliance with this regulation, 55.20 This regulation provides the scope of the operator and senior operator examinations and does not impose any specific requirements.

55. 21 This regulation describes the contents of the operator written examination and FPBL will be in compliance with this regulation.

55. 22 This regulation describes the contents of the senior operator written examination and FP6L will be in compl,ianLe with this regulation.

55.23 This regulation describes the scope of the operator and senior operator operating test.

FP&L will be in compliance with this regulation.

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 does not impose any specific requirements.

55.25 This regulation permits the NRC to administer 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. It does not impose an additional specific requirement on the Licensee.

55. 31 This regulation describes the conditions of a license and allows the NRC to impose any further conditions itspecific deems appropriate.

requirements.

It does not impose any

55. 32 This regulation states that each operator and senior operator license shall expire two years after the date of issuance. It does not impose any specific requirements.

107 3W-3 REGULATION COMPLIANCE 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. FP&L will comply with this regulation.

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 regulation requires the licensee to notify the NRC of any disability referred to in 55.11 of this part. St Lucie Unit 2 compliance is ensured through appropriate administrative 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 requirements.

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. FP&L will comply with this regulation, 55 Appendix A In accordance with this regulation, an operator and senior operator requalification program will be instituted at St Lucie Unit

2. This program will be in accordance with the requirements specified in this Appendix.

71.31 This regulation defines general standards for packaging of radioactive material. St Lucie Unit 2 will employ only packaging which conforms to these standards. Compliance is assured through appropriate administrative procedures.

71.32 This regulation governs the structural standards used in the design of large quantity (Type B) packaging'ype B packaging used at St Lucie Unit 2 will conform to these standards. Compliance is assured through appropriate administrative procedures.

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107 3W-4 REGULATION COMPLIANCE 71.33 This regulation governs the criticality standards used in the design of fissile material packages. Fissile material packaging used at St Lucie Unit 2 will conform to these standards. FP&L will comply with this regulation.

71.34 This regulation defines the general conditions used to determine the effect of transport environment on a radioactive material package. Radioactive material packages used at St Lucie Unit 2 will be only those tested in accordance with this regulation.

, 71.35 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. FP&L will comply with this regulation.

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. FP&L will comply with this regulation.

71.37 In accordance with this regulation, arrays of fissile material packaging used at St Lucie Unit 2 will be evaluated according to the standard set forth in 71.38, 71;39, and 71.40 of this part. FP&L will comply with this regulation.

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. FP&L will comply with this regulation.

71.39 This regulation merely defines the design considerations and acceptance criteria to be used in evaluating the effects of accident condi,tions on Fi.ssile Class II packages in accordance with 71.37 of this part. FP&L will comply with this regulation.

107 3W-5 REGULATION COMPLIANCE 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. FP&L will comply with this regulation.

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 St Lucie Unit 2, this regulation will be complied with.

71.42 This regulation defines special requirements for shipments of plutonium. FP&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.

r' 107 3W-6 REGULATION COMPLIANCE 100. 1 This regulation states the general purposes of Part 100 and does not impose any specific requirements.

100. 2 This regulation establishes the applicability of Part 100 and does not impose any specific requirements.

100. 3 The definitions contained in this regulation will be adhered to in applicable St Lucie Unit 2 documents.

100.10 As required by this regulation, site specifics, 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 site 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 center distance have been established and are described in FSAR Section 2.1.

(b) This paragraph applies to sites for multiple reactor facilities. St Lucie Unit 2 and St Lucie Unit 1 are independent to the extent that an accident in one reactor would not initiate an accident in another.

100 Appendix A This Appendix provides geologic and seismic siting criteria. The geologic and seismic characteristics of the St Lucie Unit 2 site are described in FSAR Chapter 2. St Lucie Unit 2 design is in compliance with this regulation as discussed in FSAR Chapter 3.

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