ML19317F430

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Forwards OL Supplemental Review Plan,Review Schedule & Review Documentation
ML19317F430
Person / Time
Site: Oconee  
Issue date: 02/16/1972
From: Morris P
US ATOMIC ENERGY COMMISSION (AEC)
To: Case E, Knuth D, Maccary R
US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8001140673
Download: ML19317F430 (27)
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11. R. Denton, CRL REVII"J PLtJi - nCONEE NUCLEAR P0iER STATION, tr,IITS 2 AND 3 DOCKET NOS AND 50-287 The Operating License Review Plan and Schedule for Oconec Units 2 a'nd 3 are attached. DRS SRS and RD assistance is requested as designated in the attached Review Plan and according to the Schedule. URS, SRS and T4 are requested to consider this Sche.fule in light of current work i

load and advise the project leader 'ay February 22, 1972 if the requestod dates cannot he txt.

f It is requested that the Technical Erecifications ns nuhnitted in the FSAR be included early_ in the reviev, no that ppropri:*te questions and comr.cnts can be provided to the opplicant with the initial additional infornation request.

l Tne rosponaibility for review of this facility is assiened to l'W. Eranch No. 4, with I. reltior assigned as Troject Leader.

R. Fer :oro vill assist in the_ review of this project. Croups assigned responalbility in the review Flan are requested to conduct reviews in accordance vith a achedule that is nubstantially shorter than the one encioned. Just hoe this vill be done vill be the subject of future directives. In the r antit:c, e dryone should try to collapse this achedule by not revicuinn items previously approved, by being sure to get t:anagenent positicas early en unresolved '

itcas. and by increasing ef fieleney as nuch as possible.

Oise g IkT A Mct Peter A. liorris, I)irector l

Division of reactor Licensing

Enclosures:

1.

Operating License Suppicm ntal Review ria for Oconee ducicar D

b Power Station, Units 2 and 3 2.

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le h Operating License Supplemental ~

Review Plan for Oconee Nuclear Power Stationi Units 2 & 3 Docket Nos. 50-270/287

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. Introduction' Duke Power Company submitted its Final Safety Analysis Report on the three-unit Oconee Nuclear Station on June 2, 1969.'The entire three-unit station was. evaluated by'the Regulatory staff and presented to

'the ACRS in August and September, 1970. The ACRS elected to write their letter < nsidering only Unit 1.

Our subsequent Safety Evalua-tion report issued December 29, 1970 considered'all three units but reflected the fact that only Unit No. 1 was,being considered for licensing. The scope and~ purpose of this review is to high'ight

<those aspects of Units 2 and 3-that are significantly different from Unit 1 and to consider any new information, AEC positions or policies 1

arising.since the earlier review.

The Oconce Nuclear Power Station employs for the reactor units a pressurized water reactor nuclear steam supply syctem furnished by Babcock and Wilcox Company (B&W).

- The applicant has requested licenses to operate each unit at a core

. power-level of 2568 MWt. All core performance data are based on this

. power level as well as site parameters, principal structures, engi-neered' safety systems and accident evaluations. This core power level is considered to be the ultimate power level for each of the

'three units. The reactor core and steam. supply systems for Units 2 Hand 3 are nearly identica1'to Unit 1 which-will be the first B&W sys-

~ tem to go into operation.

II.

Review and Evaluation of Technica1'Information j

Each group assigned to review a section* of the FSAR is expected to evaluate the section for the following-

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Content appropriate.for operating license stage.

2. ; Agreement.with the intent of AEC positions stated in Regulations, Proposed Regulations, Safety Guides, or Information Guides.

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Unreviewed and unresolved safety issues.

'*The scope of the review is limited by.the previous review al mentioned in the' introduction although all: review' sections are listed for.

completeness.

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Treatment of specific cvaluations as assigned in Part III.

5.

Adequacy of-referenced reports.

'6 Adequacy of the portions of theLTechnical Specifications that are

. frelated to the FSAR Section being reviewed.

Where more than one group is designated, each group is responsible for the review of the area of the subject consistent with the special competence of the group.

If the information is complete, in agreement with AEC positions and l

. there are no open issues, the reviewer should prepare an appropriate i

documentation of his. review and a section of the report to the ACRS.

If information is not complete or there is some unresolved issue, the reviewer should prepare a statement of the unresolved issue and a request for additional information. Since all sections of the FSAR have already received an in-depth regulatory review, maximum reliance will be placed on this prior review.

III. Operating License Review Plan (Specific Evaluations)

The. designation of the group having responsibility for review and preparation of comments, questions and report sections is indicated in_ parentheses beside each paragraph.

In those instances where more

- than one. group is designated, each group will be responsible for the specified review subject in those areas of their special competence.

Subject titles coincide with corresponding parts of the FSAR. The general design criteria referred to are those published in the Federci Register,; July 7, 1971.

Items.having an -* are judged to have been adequately reviewed and no additional review is required.

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_ Responsibility For Specific Evaluations Oconee Nuclear Power Station 2 & 3 Docket Nos. 50-270/287

- 1.0 Introduction and Summary (PWR'4) 1.1 Review and evaluate the overall station design with regard to differences between Units-2 and 3 and Unit 1 and changes that have taken place since 1970.

(PWR 4) 1.2 Review and evaluate the quality assurance program for plant operation for changes since 1970 to assure compliance with current AEC criteria, Appendix B of

_10 CFR 50, Quality Assurance Criteria for Nuclear Power Plants.

(PWR 4) 1.3 Review principal plant design for con-formance with the current issue of the General Design Criteria for Nuclear Power Plants, Appendix A of 10 CFR 50, General Design Criteria for Nuclear Power Plants. Note:

Plant already conforms to prior issue of the GDC.

  • l.4 Review and evaluate the adequacy of the information presented in topical reports referenced in the FSAR and the status of research and development programs which affect the planned operation of the facility.

(PWR 4) 1.5 Review and evaluate the adequacy o'f the application with regard to earlier ACRS expressed concerns about the design or operation of Oconee Unit 1 or of comparable facilities.

' (PWR 4) 1.6 Review and evaluate the adequacy of the application with regard to the Safety Guides.

(These have been issued since the 1970 review.)

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' l.7 Review and evaluate the adequacy of the application with regard to code classification in areas where Units 2 and 3 differ from Unit 1.

(Was done through December, 1970.)

  • 1.8 Review and evaluate the adequacy of the application with regard to seismic design and earthquake instrumentation (Safety Guide 12).
  • 1.9 For any component within the reactor coolant boundary that is designed and/or fabricated in a foreign country, evaluate the qualification of the designer and/or manufacturer.
  • l.10 For any component associated with radio-logical safety, other than components within the reactor coolant pressure boundary, that is designed and/or fabri-cated in a foreign country, evaluate the qualification of the designer and/or manufacturer.
  • l.11 Review and evaluate the capability of structures, components, and piping to withstand the effects of the design basis tornado.

(PWR 4) 1.12 Review and evaluate the adequacy of the' application with regard to safety classifications.

(PWR 4) 1.13 Review and evaluate possible inter-actions among the three nuclear units at the plant site.

(Confirmatory of-prior review and evaluation.)

2.0 Site and Environment (PWR 4) 2.1 Review and evaluate changes that could effect interactions among Units 1, 2 and 3 since initial FSAR review.

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  • 2.2 Review and evaluate the site for con-formance with the guidelines of 10 CFR 100, Reactor Site Criteria, with respect to population distribution with-in the low population zone.
  • 2.3 Review and evaluate the meteorological data presented in the FSAR and the ade-quacy of the planned meteorological pro-gram considering the effects of the proposed Units 2 and 3.

(S&RS) 2.4 Review and evaluate the adequacy of the planned radiological monitoring program to the extent of changes in AEC requirements since 1970.

  • 2.5 Review and evaluate site ground water conditions with regard to inadvertent releases of radioactive liquids and the resultant effects on sources of potable water in the area.

(PWR 4).

2.6 Review and evaluate the Oconee Station cooling water system for conformance with the proposed Safety Guide, Ultimate Heat Sink.

  • 2.7 Review and evaluate the maximum flood level estimated for the site, the ade-quacy of the design flood level and the ability of the on-site reservoir to accommodate it.
  • 2.8 Review and evaluate the adequacy of the intake structure to provide a con-tinuous source of water under all environmental conditious.
  • 2.9 Review and evaluate any new information regarding geology and seismology of the site having a bearing on the adequacy of the foundation design.

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  • 2.10 Review-and evaluate the adequacy of the acceleration values and response spectra proposed for use in the design of the structures, systems and equipment.
  • 2.11 Review and evaluate the adequacy of the proposed equipment and program for mon-itoring earthquakes (Safety Guide No.12, Instrumentation for Earthquakes).

3.0 Reactor

  • 3.1 Review and evaluate the nuclear design.
  • 3.2 Review and evaluate the thermal and I

hydraulic design.

  • 3.3 Review and evaluate the adequacy of the current analytical methods used to calculate thermal and hydraulic characteristics of the plant. Ascer-tain the conservatism of the computer programs involved in multi-dimensional analysis of power distribution.
  • 3.4 Review and evaluate the fuel systems.

(DRS) 3.5 Review and evaluate the design of the core internals with respect to plans for vibration monitoring.

  • 3.6 Review and evaluate techniques for detection of failed fuel elements.

Evaluate the adequacy of the action to be taken upon detection of failed fuel.

  • 3.7 Review and evaluate the proposed pro-gram for surveillance of the pressurized fuel elements at high burnup with re-spect to assuring the fuel elements maintain their integrity while under-going anticipated transients near end of life.

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  • 3.8 Review and evaluate the analysis of the reactor, the reactor vessel, and the internals to withstand stresses imposed by carthquake loads.
  • 3.9 Review and evaluate the proposed use and functional performance of in-core and out-of-core instrumentation.
  • 3.10 Review and evaluate the potential consequences of improper fuel loading.

within the fuel assembly and within the core.

3.11 Review and evaluate the potential for (DRS).

and consequences of a single rod uncon-trolled withdrawal if different from previous FSAR review.

4.0 Reactor Coolant System

  • 4.1 Review and evaluate significant differences from other previously approved systems pertaining to the design, fabrication and installation of the equipment and piping in the reac-tor coolant system.
  • 4.3 Review and evaluate adequacy of primary system leak detection methods, and the response to be taken upon detection of leaks.
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(DRS) 4.5 Review and evaluate the adequacy of the reactor vessel materials surveillance j

program for neutron irradiation damage for conformance with proposed Appendix H, 10 CFR 50, Reactor Vessel Material Sur-veillance Program Requirements.

  • 4.6 Review and evaluate primary system components relative to sensitization during fabrication.
  • 4.7 Review and evaluate the seismic design of the Class I and Class II (Seismic) equipment and piping. Evaluate the

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seismic design methods and the re-straints against seismic forces.

Determine the potential for generation o

of missiles and pipe Whip. Determine the adequacy of restraint or protection against missiles and pipe ship.

(DRS) 4.8 Review and evaluate adequacy of design, in-service inspection procedures, and quality control measures for primary coolant pump flywheels for conformance with Safety Guide No.14 Reactor Coolant Puro Flywheel Integrity.

(DRS) 4.9 Evaluate the, adequacy of the preopera-tional vibration program as executed on Unit 1 - if possible.

(DRS) 4.11 Review and evaluate the fracture tough-ness specification for Reactor Coolant System materials for conformance with Appendix G, 10 CFR 50, Fracture Tough-ness Requirements.

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'(DRS) 4.12 Review and evaluate the safety and operating limits (i.e., heatup and cooldown rates, maximum pressure, and minimum operating conditions) to be included in the Technical Specifications.

(DRS) 4.13 Review and evaluate the design of the reactor vessel and cavity for confor-mance with Safety Guide No. 2, Thermal Shock to Reactor Pressure Vessels.

5.0 Containment System and other special Structures

  • 5.1 Review and evaluate significant design differences from other previously approved facilities.
  • S.2 Review and evaluate the applicant's calculations of the pressure transient, in the containment following the pos-tulnted LOCA.
  • 5.3 Review and evaluate the acceptability of the proposed containment leak rate.
  • S.4 Review and evaluate the seismic design methods used for plant structures.
  • 5.5 Review and evaluate the adequacy and appropriateness of structural design criteria set forth in the FSAR.
  • 5.6 Review and evaluate the capability of the containment and other appropriate structures to withstand, without loss of integrity, the effects of: missiles and jets generated inside the structure, turbine missiles, tornado-borne missiles and hurricane or seismic forces.

(DRS) 5.7 Evaluate the acceptance tests for con-formance with Safety Guide 18, Structural Acceptance Test for Concrete Primary Reactor Containments.

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  • 5.8 Review and evaluate the internal missile protection features.
  • S.9 Review and evaluate penetcation design and methods for assuring containment leak tightness at various times in the plant life. Review and evaluate the design and isolation criteria for air locks, pipe pentrations, instrument lines, and electrical penetrations.

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Evaluate ability to test seam welds for leak tightness.

  • 5.10 Verify that material testing for installed equipment and building mate-rials use appropriate code techniques.

Consider steel containment vessel, concrete, Cadweld splices, scalants, automatic vacuum relief devices, coatings, valves, pumps and tendons.

  • 5.11 Review and evaluate the containment isolation systems.
  • 5.12 Review and evaluate the containment emergency cooling system.

(PWR 4/S&RS) 5.13 Review and evaluate the adequacy of the containment air ventilation and filtration' systems to confine and minimize releases of radioactive materials to unrestricted areas following a design basis accident.

(DRS) 5.14 Review and evaluate the containment leakage surveillance techniques for conformance with 10 CFR 50, Appendix J, Reactor Containment Leakage Testing for Water Cooled Reactors.

(Note: Review Item 5.16)

i Review and evaluate post-LOCA conditions and their long term effect on the con-

  • S.15 Include hydrogen build-up, tainment.

potential metal water reaction, contam-ination, and corrosive properties of

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containment materials (Safety Guide No. 7).

In conjunction with site analysis, review and evaluate the acceptability

  • S.16 of the postulated containment leak rate at the containment design pres-sure and at lower pressure, and review and evaluate the proposed leakage rate testing program.

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Engineered Safety Features _

Review and evaluate the performance 6.0 of the emergency core cooling system

  • 6.1 (ECCS) for conformance with the policy stated in the Interim Policy Guide dated June 19, 1971.

Review and evaluate the interlocks on the isolation valves associated with

  • 6.2 the core flooding tanks.

Review and evaluate the long-term core cooling capability following a

  • 6.3 LOCA, including the adequacy of water supply for maintaining long-term cooling capability.

Review and evaluate the ability of engir.o. red safety feature components

  • 6.4 to function in the accident environment.

Review and evaluate the ability of the containment fan cooling and recircula-

  • 6.5 tion spray systems to depressurize the containment and limit subsequent repressurization.

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~(PWR 4) 6.6 Review and evaluate the design margins on adequacy of positive suction head for ECCS pumps and containment spray Conformance with Safety Guide 1, pumps.

Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps _ will not be required.

  • 6.7 Review and evaluate the design of engineered safety features to withstand single. failures without loss of the function of heat transfer to the ulti-mate heat sink.
  • 6.8 Review and evaluate the capability for and adequacy of proposed in-service testing and inspection of engineered safety feature components.

(PWR 4) 6.9 Review and evaluate the safeguards design for ability to detect failures during operation.

(PWR 4) 6.10 Review and evaluate the extent that

" field run" piping is to be used and the manner in which the final runs are to be checked against design predic-tions. Unit 1 experience will be apolicable here.

(PWR 4) 6.11 Review and evaluate the systems provided to limit hydrogen buildup subsequent to a LOCA for conformance with Safety Guide 7, Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Accident.

(PWR 4/SRS)

  • 6.12 Review and evaluate the function of the engineered safety equipment in the post accident recirculation mode including leak detection, passive failure, and maintenance.

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(PWR 4) 6~13 Review and evaluate the equipment in the fuel pool area that mitigates the consequences of a fuel handling acci-dent for conformance with Gafety Guide 13, Fuel Storage Facility Design Basis.

(DRS) 6.14 Review and evaluate the system for switching to the recirculation mode of safety injection.

7.0 Instrumentation and Control

  • 7.1 Review and evaluate the systems for actuation and control of engineered safety features and process safety systems. Evaluate the diversity of 3x systems actuation and the ability of the systems to withstand design basis natural phenomena and the post-LOCA environment.
  • 7.2 Review and evaluate the instrumentation and control systens associated with

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process radiation monitors, refueling interlocks, area radiation monitors, site radiation monitors, and containment remote monitoring instrument systems.

Evaluate the capability of these systems to provide necessary surveillance and control required to assure conformance with the radiation exposure and radio-activity release limits of 10 CFR Part 20 and 10 CFR Part 100.

  • 7.3 Review and evaluate the instrumentation control systems related to protective function, such as the emergency power system, reactor protection system, engi-neered safeguards systems, and contain-ment isolation system and containment atmosphere control systems for confor-mance with IEEE-279, IEEE Criteria for Nuclear Power Plant Protection System.

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(PWR 4) 7.4 Review and evaluate the control room and auxiliary control station for the capability to perform necessary opera-tions under normal and abnormal I

conditions.

(DRS) 7.5 Review and evaluate the proposed test program for the control and instrumen-tation system circuitry for conformance with IEEE-366, Installation, Inspection, and Testing Requirements for Instrumen-tation and Electrical Equipment During Construction of Nuclear Power Generating Stations and IEEE-338, Criteria for Periodic Testing of Nuclear Power Generating Stations Protection Systems.

(DRS) 7.6 Review and evaluate the provisions to prevent common mode failures in the sys-tems which perform a safety function.

  • 7.7 Review and evaluate the instrumentation for sampling and monitoring the course of an accident.

(PWR 4) 7.8 Review and evaluate communication sys-tems including those systems to be.used under the industrial security plan.

8.0 Electrical Systems (DRS) 8.1 Review and evaluate the electrical power systems for confermance with General Design Criterion 17, Electrical Power Systems, IEEE-308, IEEE Criteria for Class IE Electrical Systems for Nuclear, Power Generating Stations, Safety Guide 6, and Independence Between Redun-dant Standby (Onsite) Power Sources and -

Between Their Distribution Systems.

  • 8.2 Review and evaluate the plant's inter-action with the external grid system, the relative independence of transmis-sion lines and the interaction among units at the site.
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  • 8.3 Review and evaluate the onsite power system with respect to the switchyard bus and breaker arrangement, its con-nection and interlocks, and system independence.
  • 8.4 Review and evaluate the switching cir-cuits for load shedding from the emer-gency buses.
  • 8.5 Review and evaluate the adequacy of the physical and electrical separation employed for emergency power sources, cables and containment penetrations.

9.0 Auxiliary and Emergency Systems (PWR 4) 9.1 Audit previous review and evaluation of fuel storage and handling facilities for conformance with Safety Guide 13, Fuel Storage Facility Design Basis.

-0.2 Review and evaluate the fuel pit cool-ing and cleanup system including water purity control, normal and maximum cooling capacity level and water drain control and pool leakage effects.

  • 9.3 Review and evaluate the chemical and volume control system, and boron re-covery system.

(PWR 4) 9.4 Review and evaluate station fire pro-tection systems including areas of automatic coverage, emergency power, and unsafe conditions the fire protec-tion system might create.

(PWR 4) 9.5 Review and evaluate station instrument and service air systems and their emergency power sources.

  • 9.6 Review and evaluate the component cool-ing water system and the service water system, the emergency power sources, and the effects of system failure.

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9.7 Review and evaluate ventilation systems for the Containment Building, Turbine Building, and Auxiliary Building.

Evaluate the capability to monitor and isolate the Control. Room and Fuel Build-ing ventilation s conditions (e.g.,ystems during accident fire, LOCA, and Post-LOCA H2"""'I"8)*

(PWR 4) 9.8 Review and evaluate equipment and floor drainage systems including reliability of level instrumentation and alarms, lower limits of leak detection and sur-veillance of radioactivity of drains.

  • 9.9 Review and evaluate the residual heat j

removal system, its emergency power sources and the effects of system failure.

(PWR 4) 9.10 Review and evaluate the capability of the interlocks on the isolation valves between the residual heat removal sys-tem or other low pressure systems and the reactor coolant system to prevent a LOCA due to deliberate or inadvertent operation of the valves.

Evaluate interlocks with PWR Position Paper -

Motor Operated Valves in the Residual Heat Removal System in PWR Plants (February 6, 1971).

10.0 Steam and Power Conversion System (PWR 4) 10.1 Review and evaluate the effects of transients initiated by turbine and generator trips with and without normal control actions.

_(PWR 4) 10.2 Review and evaluate main steam transients initiated.by the rejection and acceptance ioad with and without normal control action.

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Review and evaluate the capability of

  • 10.4 the system to remove residual heat during emergency conditions.

Review and evaluate the provisions for

  • 10.5 preventing and mitigat'ng the effects of turbine-generator missiles such as QA, inservice inspections and tests, and equipment orientation.

Review and evaluate the design features

  • 10.6 previded to minimize the effect of a steam line rupture.

f Review and evaluate the provisions made 10.7

-(DRS) for inservice inspection of the secondary system.

Radioactive Waste and Radiation Protection 11.0 Review and evaluate the design and 11.1 (S&RS) capability of the radioactive waste system to collect, confine, process, dispose, and monitor radioactive waste within the limitations of 10 CFR 20, 10 CFR 50, and 10 CFR 100 under normal and abnormal operating conditions. Evaluate conformance with the intent of proposed changes of 10 CFR 20 and 10 CFR 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low As Practicable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, concerning radiation exposures and releases of radioactive materials to' unrestricted areas.

Review and evaluate the estimated normal 11.2 (SERS) releases and possible additional means for reducing planned or accidental release of radioactive waste to unre-stricted areas.

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(PWR.4) 11.3 Review and evaluate the planned pro-visions for protection of personnel and selected equipment / materials against radiation during'either nor-mal or abnormal operating conditions.

  • 11.4 Review and evaluate the criteria for control room shielding.

12.0 Conduct of Operation

  • 12.1 Review and evaluate the t<!chnical qualifications of the applicant and principal contractors.

(RO) 12.2 Review and evaluate the fr)llowing programs, plans and concepts:

Plant staffing and responsi2Llity a.

assignment.

b.

Training of plant personnel for conformance with Safety Guide No. 8, Personnel Selectit3 and Training.

  • c.

Preparation and maintenance of written procedures.

d.

Normal modes of operation.

  • e.

Preparation and maintenance of records.

  • f.

Operational review and audit.

g.

Emergency plans for conformance with 10 CFR 50, Appendix E, Emergency Plans for Protection and Utilization Facilities.

(RO) 12.3 Review and evaluate industrial security program for conformance with the 33 cri-teria in the draft proposed Appendix K to 10 CFR 50 per P. A. Morris guidance memo dated January 18, 1972.

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~ Initial Tests and Operations (PWR 4/RO) 13.1 Review and evaluate the following:

a.

Preoperational test program, b.

Initial criticalit.y test program.

c.

Post-criticality test program, d.

Proposed operating restrictions.

14.0 Safety Analysis _

  • 14.1 Review and evcluete the analysis of the following faults to assure that the analytical techniques are conservative and consistent with the plant design and the expecimental data:

a.

Reacter coolant pipe rupture.

b.

Steam pipe rupture, c.

Steam generator tube rupture, d.

Reactor coolant pump locked rotor.

Fuel rupture during handling operation.

e.

f.

Control rod ejection.

Improper positioning of fuel assembly.

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h.

Uncontrolled RCCA withdrawal.

  • 14.2 Review and evaluate the analytical techniques, methods of analysis and the results for the following accidents as performed for recent PWR's:

RCC assembly misalignment.

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b.

Chemical and Volume Control System malfunct on.

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coolant loop, d.

Excessive heat removal due to feedwater system malfunction.

Excessive load increase, e.

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Loss of forced reactor coolant flow.

Loss of external electrical load g.

and/or turbine.

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Loss of normal feedwater.

i. Loss of all AC power to the station auxiliaries,
j. Waste gas decay tank rupture.

(SRS) 14.3 Evaluate analysis for conformance with Safety cuide 4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Acci-dent for Pressurized Water Reactors -

(Examine existing DRL analysis completed prior to 1971.)

15.0 Technical Specifications (PWR 4/DRS/S&RS) 15.1 Review and evaluate the technical specifications for those areas not (RO) addressed in Oconee 1 Tech Specs.

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ITS 2 AND 3_

REVIEW SCHEDULE - TECHNICAL - OCONEE NUCLEAR IV.

................ 6/2/69 Application Submitted for Units 1, 2 and 3)...........................

2/10/72 (for Units 2 and 3).....................................

1.

Initial Meeting with Applicant P Branch.......................

2/28/72

. 2.

Initial Draft Request for Additional Information - To R

................... 3/13/72

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Technical Meeting with Appli cant....................................

3/20/72 4.

t Formal Request for Additional Information - To DRL Managemen..........................

3/27/72

~5.

Formal Request for Additfonal Information - To Applicant.............

6.

Interim Report Drafts - To RP Branch.............

................. Delete 7.

Interim Report - To DRL Management.....................................................

5/1/72 8.

Responses from Applicant...................................................................

5/22/72 9.

Second Draft Request for Additional Information - To RP Branch.

................... 5/30/72 10.

t Formal R'equest for Additional Information - To DRL Managemen.....

........... 6/5/72 11.

Formal Request for Additional Information - To Applicant...................

12.

Responses from Applicant..................... *. *.********************* *

      • .,,,. 7/31/72 13.

Draft Sections of ACRS Report - To RP Eranch.....................

14.

ACRS Report - To DRL Management.......................

15.

Tr ansmi t ACRS Re po r t....................

16.

ACRS Meeting...................

17.

Safety Evaluation.....................

18.

Issued 1-28-72

.I.

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Review Documentation - Oconee Nucicar Power Station, Units 2 & 3 V.

For each part of the review set forth in the review plan, the assigned reviewer should provide a report to the project 1cader which contains a concise description of the safety issue; a discussion three parts:

that summarizes the significant information; and the conclusions and recommendations that are based on the information presented in the discussion.

The description should state the safety issue clearly, with only the The discussion design details needed to understand the problem.

should state the essential information provided by the applicant, the analysis that was performed, the conclusions reached and a summary The of the actions taken by the applicant as a result of our review.

conclusion should summarize the significant points of our position and indicate the applicant's compliance with them.

The content of the report should be appropriate for the subject being documented. For example:

The reports that discuss an issue that is applicable to Oconee 2/3 for which a specific regulatory position must be developed will require a complete explanation of the issue and the alternatives.

The reports that discuss a generic issue that is being reviewed on all plants of a given type and for which a regulatory position is being developed should completely explain the new information presented on the particular plant; however, portions of the issue that have been resolved and documented in previous applications should be sentioned only briefly.or by reference.

The reports that discuss compliance with a regulatory position, as forth in published or proposed Regulations, Safety Guides, Infor-set mation Guides or PWR Position Papers, can be very brief. The descrip-tion need only identify the safety issue, the discussion should summarize the essential information provided by the applicant which indicates compliance, and the conclusions should indicate the com-pliance with regulatory position.

Each report, appropriate for the phase of review being documented, should be forwarded to the project leader at the completion of the initial review, and within one month of the receipt of the applicant's final submittal on the subject.

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If the applicant _'s information is not sufficient to complete the major 'part of the evaluation, the discussion should summarize the

- type of information that is missing and a request for the additional information that is necessary to resolve the issue should be prepared.

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IPeltier E. J. 31och DR IVKaras (2)

RRossi S. II. Eanauer, D F. Schroeder, DEL T. R. Vilson, DRL

~

Assistant Directo DEL E. C. Cese, DRS' Assistant Directors, RS R. V. Kicckcr. DRL'

~

DRL Eranch Chiefs DRS Eranch Chiefs EEVIEW PLNI - OCONER UUCLEA POWER STATION, itIITS 2 A','D 3 DOCKET KOS. 50-270 AND 50-28 1/

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The Operating License Review P in and Schedule for Oconeo Units 2 and 3 DRS assistance'ikrequested in those areas desinnnted in are attached.

the attached Review Plan and according to the Schedule. DRS is requested to consider this Schedule in lighkof,its current work load and advise us by February 22, 1972 if the req [uested ' dates '.cannot be vet.

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It is requested that the Tehhnical Specifications as nubmitted in the r5AR be included early in/' our reviewkso that appropriate questions ami coteents can he provided to the appliennt with the initisl additional infornation request.

/

The renpoesibility for/ review of this facility is assitned to P No. 4, with I. Peltier assirned as Project 'I.eader.

R. 2crnero vill assist in the review of this project. Croups assigned responsibility in the' Review Tian are requested to conduct reviews in secordance tif th a schedule that is cubstantially shorter than the one enclosed. Just hev this will be donc silll be the subject of future directives., In the reantice, everyene should try to collapse this schedule by'not reviewing items previously approved, by being; sure to get r.anager.ent positions early on tmresolved itens, and by increasin:; ef ficiency cs r:nch as possible.

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's g6 Peter A. Morris. Director Division of Reactor Licensing

Enclosures:

1.

Operating License Suppleeental

. _.,,. _ _, - -,, _ _ __ v _,___

2.

Reviev Schedule.

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h Forum AEC-310 (Rev.9-M) AECM 0240 ' M Qd; '

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DISTRI3UTION Docket (2)

DRL Reading DR Panding PWR-4 Reading PAMorris CO (2)

IPeltier PWKaras (2)

~M-RRossi E. J. Elech, DR

5. II.11annuer, DR -

F. Schroeder, I q.

T. R. Vilson, ERLg

/asiotant Directoro DRL g

E. C. Case, DRS

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Assis t ant Di rt*c tu: s, P'IS R. U. Klecker, DRL DRL Branch t%1efs ins T> ranch Chiefs RTVIEW PLAN - OCO'iEE NUCLEAR 'OER STATII, r';ITS 2 AND 3 4

DOCKET NOS. 50-270 AND 50-207 h Operatinct License Reviev Pinn\\for Oc:mec ';uclear To"cr Statien, Units 2 and 3, is encioned. The ScKec'u'.a and fornat for the docu:nentation of the review nre encloqe! alco. Creeps assi> ned responsibility in the Feview Plan cre

, ested to conduct rev!eus in accordance with this schedule.

It is requested that the Technical Specifications ss subritted in the PSAR be ' included early in your reviev, sa th'at appropriate questions and cotments can he provided to the applicant \\ith the initial additional infornation request.

N responalbility To-the reviev of this factIity !c t-s.irced to pit. Ernnch No. 4, with I. Peltier a sirmd as ProjectyInder.

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Peter A. Torris, Director Divinion of Peactor Lf censing

Enclosures:

1..,0perating' License Supple:r. ental Review Plan for oconee Nuc1 car l

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' Foweg Station, Units 2 and 3 g

01 4pg 2.

Bevisw Schedule ~

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Review Documentation.4 DRL:PWR-4, DRL PWR-4 DRL:AD/PWh DRL:AD DRL:DIR E

IAPeltie emp we RCDeYoung:

TRWilson PAMorris SURNAtfE k

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_2/__/72

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