ML011930012

Letter Informing of Issuance of Amendment No. 16, Technical Specification Change No.26 for License No DPR-38

ML011930012
Person / Time
Site:	Oconee
Issue date:	12/22/1975
From:	Purple R Office of Nuclear Reactor Regulation
To:	Parker W Duke Power Co
References
Download: ML011930012 (43)

Text

DEC 2 2 1975 Docket Nos.

50-270 and 50-287 Duke Power Company ATTN: Mr. William 0. Parker, Jr.

Vice President Steam Production Post Office Box 2178 422 South Church Street Charlotte, North Carolina 28242 Gentlemen:

The Commission has issued the enclosed Amendment iko. 1 6, Technical 1 6 Specification Change -No. 2 6 for License No. DPR-38; Amendment No.

Technical Specification Change No. 2 '_for License No. DPR-47; and Amendment No.1 3, Technical Specification Change 1 3 for License 1o.

No. DPR-5S, for the Oconee Nuclear Station, U"its 1, 2, and 3. These amendments are in response to your request dated January 15, 1975.

The amendment incorporates into the Oconee Nuclear Station Technical Specifications changes to the reporting requirements. Changes to your proposal were necessary to meet our requirements. These have been discussed with your staff. The technical specifications are based on Regulatory Guide 1.16. "Reporting of Operating Information - Appendix A Technical Specifications", Revision 4.

We request that you use the formats presented in the Appendices to Regulatory Guide 1.16, Revision 4, for reporting operating information and that you report events of the type described under the section "Events of Potential Public Interest". Instructions for using these is reporting formats are contained in Regulatory Guide 1.16 fa copy enclosed for your se), and ABC report OCE-SS-00I titled "Instructions for Preparation of Data Entry Sheets for Licensee Event Report (LER)

File" (a copy of which was provided you previously). This report is modified by udated instructions dated December 8, 1975, which are enclosed. Copy requirements are summarized in Regulatory Guide 10.1, CG "Compilation of Reporting Requirements for Persons Subject to NRC Regulations", a copy of which is also enclosed. This Guide will assist you in identifying reports that are required by the Commission's regulations set forth in Title 10 Code of Federal Regulations but are not contained in your technical specifications. Reports that are requirea by the regulations have not been repeated in your technical specifications.

SURNAME - -.----------------

DATE 0 1...................

FeOm AEC-318 (Rev. 9-53) AECM 0240 GPO 043-16--81465-1 445-678

Duke Power Company -2 -

DEC 2 2 1975 Copies of the related Safety Evaluation and the Pederal Register Notice also are eonlosed.

Sinceely,,

Original 7igned bM X A. Pu-rP1Q _..._-i Robert A. PUrple, Chief Operating Reactors Branch 01 Division of Reactor Licensing

Enclosures:

1. Amendment No. 1 6

2. Amendment No, 1 6

3. AmendmentNo.13

4. Regulatory Guide 1,16

5. Updated Instrutions

6. Regulatory Guide 10.1

7. Safety Evaluation

8. Fedoral Register Notice cc w/enclosures 3 cc W/enolosurs 4 inaming:

Mr. William L. Porter Mr, Elnqr Whitten Duke Power Company State Clearinghouse P. 0. Box 2178 Office 4f the Govornor 422 South Church Street Divisio, of Admisistration Charlotte, North Carolina 28242 129S Peodleton Street Fourth loer Mr. Troy 8. Conner Col~mb 0, South Carolina Conner 4 Knott 1747 Pennsylvania Avenue, NK DISTRIBION 20006 Docket Fle "(3) ORB#1 Reading Washingtonj, D.C. Local PDR NRC PDRs (3)

TBAberna' 9 y, TIC KRGol ler Oconee Public Library RAPurple 201 South Spring Stret TJCarter GZech SMSheppard Walhallas South Carolina 29691 JMcGough SIari SVarga DEisenhut Hononable, Reese -A, Hubbard NDube BJones (4)

County Supervisor of Oconee County BScharf (15) MHebron Walhalla, South Carolina 29621 JSattznaat PCollins AESteen ACRS 116)

CE LD EPLA (2) 01 &E (3)

TP -~r hs*Zechue* .. (see.note_ --

4 .

12/16/7S 1..- AECM 0240 DATE 0Re53) 12I/3/75)

-I---------------- -

12/247s Form AEC-3 18 (Rev. 9-53) AECM 0240 GPO .43--16--81465-1 44d5-,378

UNITED STATES NUCLEAR REGULATORY COMMIS;ION WASHINGTON. D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 1 8 License No. DPR-38

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Duke Power Company (the licensee) dated January 15, 1975, complies with the standards and requirements of the Atomic Energy Act of of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; and D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-38 is hereby amended to read as follows:

'Ž?_%J10A

"1B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No.2 8 -.'

3. This license amendment is effective January 1, 1976.

FOR THE NUCLEAR REGULATORY COMMISSION Origina' signed by, H., A. Puripi]e Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing

Attachment:

Change No. , G to Ithe Technical Specifications Date of Issuance: DEC 2 2 1975

.................. . .............................................. ..I ............................................ ............................................. .......................................

A C F Rm E-3 .........................

0240 Form .A.C-318 (Rev. 9-53) A.ECM: *g uý S, GiOVKgMN'grPRNT pqII*NGI OFIC*I* 1[74-526-¶166

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment Ro. 16 License No. DPR-47

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Duke Power Company (the licensee) dated January 15, 1975, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; and D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-47 is hereby amended to read as follows:

40 %oUTlOA, M

cc 6~9~

"11B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised,by issued chan~es thereto through Change No.2, 1."

3. This license amendment is effective January 1, 1976.

FOR THE NUCLEAR PRGULATORY COMMISSION 01191nal signed by P4 A.PurplqL ,

Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing

Attachment:

Change No. p I to ýthe Technical Specifications Date of Issuance: 2 4 '275 AURNAME P Forml AE._3 (*Ro. 9-53) AEC 0240

• W* 9 GOVERN*)MENT PRINTING OPPICEI 1974-826-186

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 DUKE POWER COMPANYý DOCKET NO. 50-287 OCONEE NUCLEAR STATION, NIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No. DPR-55

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Duke Power Company (the licensee) dated January 15, 1975, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; and D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B of Facility License No. DPR-55 is hereby amended to read as follows:

0oUTIOA, 1/2?6_191'

"1B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change 'No. 1 3 ""

3. This license amendment is effective January 1, 1976.

FOR THE NUCLEAR REGULATORY COMMISSION sined by nOrigin X A.Purple Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing

Attachment:

Change No. 1 , to the Technical Specifications Date of Issuance: DEL 22 1975 Form AEC-318 (Rev. 9-53) AECM 0240

• U' S; GOVE'RNMENT PRINmTING OF!cgi i974-a-;ie"

ATTAChMENT TO LICENSE AMENDMENTS AMENDMENT NO.1 6 TO FACILITY LICENSE NO. DPR-38 CHANGE NO.2 3 TO.TECHNICAL SPECIFICATIONS; AMENDMENT NO.1 6 TO' FACILITY LICENSE NO. DPR-47 CHANGE NO. I 1 TO TECILNICAL SPECIFICATIONS; AMENDMENT NO. i V TO 'FACILITY LICENSE NO. DPR-55 CHANGE NO. . a TO TECHNICAL SPECIFICATIONS DOCKET NOS. 50-269, 50-270, AND 50-287 Revise Appendix A as follows:

Remove Pages Insert New Pages ic i

ii ii iii .

iii iv iv v v vi vi 1-5 1-5 (blank) 3.1-19 3.1-19 3.1-19a 3.1-20 3.1-20 4.2-1 4.2-1 4.2-2 4.2-2 4.2-3 4.2-3 4.4-1 4.4-1 4.4-2 4.4-2 4.4-3 4.4-3 4.4-4 4.4-4 4.4-7 4.4-7 4.4-8 4.4-8 4.4-9 4.4-9 4.4-10 4.4-10 4.13-1 4.13-1 6.1-2 6.1-2 6.1-4 6.1-4 6.2-1 6.2-1 6.6-1 6.6-1 thru thru 6.6-12 6.6-9

c-Section Page 1.5.4 Instrument Channel Calibration 1-3 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 QUADRANT POWER TILT 1-4 1.7 CONTAINMENT INTEGRITY 1-4 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1-1 2.1 SAFETY LIMITS, REACTOR CORE 2.1-1 2.2 SAFETY LIMIT, REACTOR COOLANT SYSTEM PRESSURE 2.2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE 2.3-1 INSTRUMENTATION 3 LIMITING CONDITIONS FOR OPERATION 3.1-1 3.1 REACTOR COOLANT SYSTEM 3.1-1 3.1.1 Operational Components 3.1-1 3.1.2 Pressurization, Heatup, and Cooldown Limitations 3.1-3 3.1.3 Minimum Conditions for Criticality 3.1-8 3.1.4 Reactor Coolant System Activity 3.1-10 3.1.5 Chemistry 3.1-12 3.1.6 Leakage 3.1-14 3.1.7 Moderator Temperature Coefficient of Reactivity 3.1-17 3.1.8 Single Loop Restrictions 3.1-19 3.1.9 Low Power Physics Testing Restrictions 3.1-20 3.1.10 Control Rod Operation 3.1-21 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS 3.2-1 3.3 EMERGENCY CORE COOLING, REACTOR BUILDING COOLING, REACTOR 3.3-1 BUILDING SPRAY, AND PENETRATION ROOM VENTILATION SYSTEMS ii 22, !975

Section Page 4.5-.6 4.5.2 Reactor Building Cooling Systems 4.5- 10 4.5.3 Penetration Room Ventilation System 4.5-12 4.5.4 Low Pressure Injection System Leakage 4.6-1 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTING 4.7-1 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7-1 4.7.1 Control Rod Drive System Functional Tests 4.7-2 4.7.2 Control Rod Program Verification 4.8-1 4.8 MAIN STEAM STOP VALVES 4.9-1 4.9 EMERGENCY FEEDWATER PUMP PERIODIC TESTING 4.10-1 4.10 REACTIVITY ANOMALIES 4* l1-1 4.11 ENVIRONMENTAL SURVEILLANCE 4.12-1 4.12 CONTROL ROOM FILTERING SYSTEM 4.13-1 4.13 FUEL SURVEILLANCE 4.14-1 4.14 REACTOR BUILDING PURGE FILTERING SYSTEM 4.15-1 4.15 IODINE RADIATION MONITORING FILTERS 4.16-1 4.16 RADIOACTIVE MATERIALS SOURCES 5.1-1 5 DESIGN FEATURES 5.1-1 5.1 SITE 5.2-1 5.2 CONTAINMENT 5.3-1 5.3 REACTOR 5.4-1 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 6.1i-1 6 ADMINISTRATIVE CONTROLS 6.1-1 6.1 ORGANIZATION, REVIEW, AND AUDIT 6.1-1 6.1.1 Organization 6.1-2 6.1.2 Review and Audit 6.2-1 6.2 ACTION TO BE TAKEN IN THE EVENT OF AN INCIDENT 12 9/2 ', / 1 :

REPORTABLE TO THE COMMISSION iv

.E: 2) 1975

3.1.8 Single Loop Restrictions-Specification The following special limitations are placed on sfngle loop operation in addition to the limitations set forth in Specification 2.3.

3.1.8.1 Single loop operation is authorized for test purposes only.

3.1.8.2 At least 23 incore detectors meeting the requirements of Technical Specification 3.5.4.1 and 3.5.4.2 shall be available throughout this test to check gross core power distribution.

3.1.8.3 The pump monitor trip setpoint shall be set at no greater than 50 percent of rated power.

3.1.8.4 The outlet reactor coolant temperature trip setpoint shall be set at no greater than 610F.

3.1.8.5 At 15 percent of rated power and every 10 percent of rated power above 15 percent, measurements shall be taken of each operable incore neutron detector and each operable incore thermocouple, reactor coolant loop flow rates and vessel inlet and outlet temperature, and evaluation of this data determined to be at ceptable before proceeding to higher power levels.

3.1.8.6 A report covering single loop operation, permitted by Specification 3.1.8, shall be submitted within 90 days after completion of testing.

This report shall :include the data obtained together with analyses and interpretations of these data which demonstrate:

(1) Coolant flows in the idle loop and operating loop are as 26 predicted. 2 (2) Relative incore flux and temperature profiles remain es- I sentially the same as for four pump operation at each power level taking into account the reduced flow in single loop operation.

(3) Operating loop temperatures and flows are obtained which justify the revised safety system setting prescribed for the temperature and flow instruments located in the operating loop (which must sense the combined core flow plus the cooler bypass flow of the idle loop).

Subsequent single loop operation shall be contingent upon Commission approval.

Bases The purpose of single loop testing is to (1) supplement the 1/6 scale model test information, (2) verify predicted flow through the idle loop, (3) verify that changes in power level do not affect flow distribution or core power 3.1-19 S1975

clistribution, and (4) demonstrate that limiting safety system settings (pump monitor trip setpoint and reactor coolant outlet temperature trip setpoint) can be conservatively adjusted taking into account instrument errors.

Limiting the pump monitor trip setpoint to 50 percent uf rated power and the reactor coolant outlet temperature trip setpoint to 610°F to perform this con firmatory testing assures operation well within the core protective safety limits shown in Figure 2.1-3, Curve 2.

Incore thermocouples will be installed and data will be taken to check outlet core temperature profiles. These data will be used in evaluating test results.

3.1-19a ou 2' 1975

'3.1. 9 Low Power Physics Testing Restrictions Specification The following special limitations are placed on low power physics testing.

3.1.9.1 Reactor Protective System Requirements

z. Below 1720 psig shutdown bypass trip setting limits shall apply in accordance with Table 2.3-lA - Unit 1.

2.3-1B - Unit 2.

2.3-IC - Unit 3.

b. Above 1800 psig nuclear overpower trip shall be set at less than 5.0 percent. Other settings shall be in accordance with Table 2.3-lA - Unit 1.

2.3-lB - Unit 2.

2.3-IC - Unit 3.

3.1.9.2 Startup rate rod withdrawal hold shall be in effect at all times. This applies to both the source and intermediate ranges.

Bases Technical Specification 3.1.9.2 will apply to both the source and intermediate ranges.

The above specification provides additional safety margins during low power physics testing. ( N 3.1-20 4 - Ift 41975

4.2 REACTOR COOLANT SYSTEM SURVEILLANCE Applicability pressure boundary.

Applies to the surveillance of the Reactor Coolant Sy~tem Objective To assure the continued integrity of the Reactor Coolant System pressure boundary.

Specification 4.2.1 Prior to initial unit operation, an ultrasonic test survey shall be made of Reactor Coolant System pressure boundary welds as required to establish preoperational integrity and bas'eline data for future inspections.

4.2.2 Post-operational inspections of components shall be made in ac and cordance with the methods and intervals indicated in IS-242 Code, IS-261 of Section XI of the ASME Boiler and Pressure Vessel 1970, including 1970 winter addenda,.except as follows:

IS-261 Item Component Exception-Primary Nozzle to Vessel 1 RC outlet nozzle to be 1.4 inspected after approxi Welds mately 3 1/3 years operation. 2nd RC outlet nozzle to be inspected after approx. 6 2/3 yrs.

operation. 4 RC inlet nozzles and 2 core flooding nozzles to be in spected at or near end of interval 3.3 Primary Nozzle to Safe End Not Applicable Welds Not Applicable 4.3 Valve Pressure Retaining Bolting Larger than 2" Not Applicable 6.1 Valve Body Welds Not Applicable 6.3 Valve to Safe End Welds Not Applicable 6.6 Integrally Welded Valve Supports Not Applicable 6.7 Valve Supports & Hangers DLc 22 1975 4.2-1

4.2.3 The structural integrity of the Reactor Coolant System boundary shall be maintained at the level required by the original ac ceptance standards throughout the life of the station. Any evidence, as a result of the tests outline& in Table IS-261 of Section XI of the code, that defects have dcveloped or grown, shall be investigated, including evaluation of comparable areas of the Reactor Coolant System.

4.2.4 The results of the Inservice Inspections performed pursuant to

• 2 Specifications 4.2.1, 4.2.2, and 4.2.3 shall be reported to the 21 Commission within 90 days of completion. I*

4.2.5 To assure the structual integrity of the reactor internals through out the life of the unit, the two sets of main internals bolts (connecting the core barrel to the core support shield and to the lower grid cylinder) shall remain in place and under tension. This will be verified by visual inspection to determine that the welded bolt locking caps renain in place. All locking caps will be inspected after hot functional testing and whenever the internals are removed from the vessel during a refueling or maintenance shutdown. The core barrel to core support shield caps will be inspected each refueling shutdown.

4.2.6 Sufficient records of each inspection shall be kept to allow com parison and evaluation o. future inspections.

4.2.7 The inservice inspection program shall be reviewed at the end of five years to consider incorporation of new inspection techniques and equipment which have been proved practical and the conclusions of this review and evaluation shall be discussed with the NRC/ORI 4.2.8 At approximately three-year intervals, the bore and keyway of each reactor coolant pump flywheel shall be subjected to an in-place, volumetric examination. Whenever maintenance or repair activities necessitate flywheel removal, a surface examination of exposed surfaces and a complete volumetric examination shall be performed, if the interval measured from the previous such inspection is greater than 6 2/3 years.

4.2.9 For Unit 1 and Unit 2, a B Type vessel specimen capsule shall be withdrawn after one year of operation and an A Type capsule shall be withdrawn after 11, 17, and 22 years of operation. The with drawal schedules may be modified to coincide with those refueling outages or unit shutdowns most closely approaching the withdrawal schedule. Specimens thus withdrawn shall be tested in accordance with ASTU-E-185-70. For Unit 3, a B Type vessel specimen capsule shall be withdrawn after one year of operation and an A Type capsule shall be withdrawn after 7, 14, and 17 years of operation.

The withdrawal schedules may be modified to coincide with those refueling outages or unit shutdowns most closely approaching the withdrawal schedule. Specimens thus withdrawn shall be tested in accordance with ASTM-E-185-72. The results of these examinations 213 shall be reported to the Commission within 90 days of completion 21 of testing. *'

4.2-2 DEC0 2 297

• j 4.2.10 During the first two refueling periods, two reactor coolant system piping elbows shall be ultrasonically inspected along their longitudinal welds (4 inches beyond each side) for clad bonding and for cracks in both the clad and base metal. The elbows to be inspected are identified in B&W Report 1364 dated December 1970.

Bases The surveillance program has been developed to comply with Section XI of the ASME Boiler and Pressure Vessel Code, Inservice Inspection of Nuclear Reactor Coolant Systems, 1970, including 1970 winter addenda, edition. The program places major emphasis on the area of highest stress concentrations and on areas where fast neutron irradiation might be sufficient to change material properties.

The reactor vessel specimen sbrveillance program for Unit 1 and Unit 2 is based on equivalent exposure times of 1.8, 19.8, 30.6 and 39.6 years. The contents of the different type of capsules are defined below.

AType B Type Weld Material IIAZ Material HAZ Material Baseline Material Baseline Material For Unit 3, the Reactor Vessel Surveillance Program is based on equivalent exposure times of 1.8, 13.3, 26.7, and 30.0 years. The specimens have been selected and fabricated as specified in ASTM-E-185-72.

Early inspection of Reactor Coolant System piping elbows is considered desirable in order to reconfirm the integrity of the carbon steel base metal when explosively clad with sensitized stainless steel. If no degradation is observed during the two annual inspections, surveillance requirements will revert to Section XI of the ASME Boiler and Pressure Vessel Code.

4.2-3 DLECO 2 (f 9

4.4 REACTOR BUILDING 4.A4.1 Containment Leakage Tests Applicability Applies to containment leakage.

Objective To verify that leakage from the Reactor Building is mai tained within allowable limits.

Specification 4.4.1.1 Integrated Leak Rate Tests 4.4.1.1.1 Design Pressure Leak Rate The maximum allowable integrated leak rate, La, from the Reactor Building at the 59 psig design pressure, Pp, shall not exceed 0.25 weight percent of the building atmosphere at that pressure per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.1.1.2 Testing at Reduced Pressure The periodic integrated leak rate test may be performed at a test pressure, Pt, of not less than 29.5 psig provided the resultant leakage rate, Lt, does not exceed a pre-established fraction of La determined as follows:

a. Prior to reactor operation the initial value of the integrated leak rate of the Reactor Building shall be measured at design pressure and at the reduced pressure to be used in the periodic integrated leak rate tests. The leak rates thus measured shall be identified as Lpm and Ltm respectively.

b. Lt shall not exceed La(Ltm/Lpm) for values of (Ltm/Lpm) not greater than 0.7.

2

c. Lt shall not exceed La(Pt/Pp) for values of (Ltm/Lpm) above 0.7.

d. If Ltm/Lpm is less than 0.3, the initial integrated iest results shall be subject to review by the NRC to establish an acceptable value of Lt.

4.4.1.1.3 Conduct of Tests

a. The test duration shall be at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except that if both the following conditions are met, the test duration shall be at least 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />s:

(1) All test conditions, including the test procedure, shall be similar to the initial integrated leak rate tests.

(2) When the test is terminated, building pressure shall have stabilized and shall not be increasing.

4.4-1 DEC 2 2 1975

as measuring

b. Test accuracy shall be verified by'supplementary means, such or by im the quantity of air required to return to the starting point measurements.

posing a known leak rate to demonstrate the validity of the test shall C. Closure of containment isolation valves for the purpose of the valves be accomplished by the means provided for normal operation of without preliminary exercises or adjustment.

4.4.1.1.4 Frequency of Test After the initial preoperational leak rate test, two integrated leak rate between each major tests shall be performed at approximately equal intervals at 10 year intervals. In shutdown for inservice inspection, to be performed at each 10 year addition, an integrated leak rate test shall be performed interval, coinciding with the inservice inspection shutdown.

4.4.1.1.5 Conditions for Return to Criticality

a. If Lt is not greater than 50 percent of the value permitted in 4.4.1.1.2, to criti local leak rate testing need not be completed.prior to return cality following a periodic integrated leak rate test.

percent of the

b. If Lt is greater than 50 percent and not greater than 100 to criticality will be perfornied value permitted in 4.4.1.1.2, return leakage into the penetration conditioned upon demonstration tUat local leakage above room, measured at full design pressure, accounts for all If this cannot be demon 50 percent of that permitted by 4.4.1.1.2.

the reactor shall be strated within 30 days of returning to criticality, shut down.

c. If Lt is greater than 100 percent of the value permitted by 4.4.1.1.2, the unit shall not be made critical.

4.4.1.1.6 Corrective Action and Retest or 4.4.1.1.2, the If repairs are necessary to meet the criteria of 4.4.1.1.1 local leak rate integrated leak rate test need not be repeated, provided that the leak measurements are made before and after repair to demonstrate integrated rate reduction achieved by repairs reduces the overall measured leak rate to an-acceptable value.

4.4.1.1.7 Report of Test Results Containment'integrated leak rate test and subsequent 2 The results of the initial periodic tests shall be the subject of a summary technical report which shall 2 1 test.

be submitted to the Commission within 90 days of completion of the 4.4.1.2 Local Leak Rate Tests 4.4.1.2.1 Scope of Testing The local leak rate shall be measured for each of the following components:

DE 2, 1975 4.4-2

a. Personnel hatch

b. Emergency hatch

c. Equipment hatch seals

d. Fuel transfer tube seals

e. Reactor Building normal sump drain line

f. Reactor coolant pump seal outlet line

g. Reactor coolant pump seal inlet line

h. Quench tank drain line

i. Quench tank return line

j. Quench tank vent line

k. Normal makeup to Reactor Coolant System

1. High pressure injection line

m. Electrical penetrations

n. Reactor Building purge inlet line

o. Reactor Building purge outlet line

p. Reactor Building sample lines

q. Reactor coolant letdown line 4.4.1.2.2 Conduct of Tests

a. Local leak rate tests shall be performed at a pressure of not less than 59 psig.

b. Acceptable methods of testing are halogen gas detection, soap bubbles, pressure decay, hydrostatic flow or equivalent.

4.4.1.2.3 Acceptance Criteria The total leakage from all penetrations and isolation valves shall not exceed 0.125 weight percent of the Reactor Building atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.1.2.4 Corrective Action and Retest

a. If at any time it is determined that the criterion of 4.4.1.2.3 above is exceeded, repairs shall be initiated immediately.

b. If conformance to the criterion of 4.4.1.2.3 is not demonstrated within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following detection of excessive local leakage, the reactor shall be shut down and depressurized until repairs are effected and the local leakage meets the acceptance criterion as demonstrated by retest.

4.4.1.2.5 Test Frequency Local leak detection tests shall be performed annually, except that:

a. The equipment hatch and fuel transfer tube seals shall be additionally tested after each opening.

b. The personnel hatch and emergency hatch outer door seals shall be tested at four-month intervals, except when the hatches are not opened during that interval. In no case shall the test interval be longer than 12 months.

4.4-3 U12 2 1975

Isolation Valve Functional Tests 4.4.1.3 valves shall be remotely-operated Reactor Building isolation such Quarterly, required to fulfill their safety function unless stroked to the position The latter valves shall be S .. r.-tical during unit operation.

operation is -0u P* . . . .

tested during each refueling shutdown.

4.4.1.4 Annual Inspection of the of the accessible interior and .ex erior surfaces A visual examination rmed annually and its components shall be perf containment structure andleak rate test, to uncover any evidence of deterioration prior to any integrated structural ir tegrity or leak-tightness which may affect either the containment's by cor of any significant deterioration shall be accompanied The discovery non-destructive tests in accord with acceptable procedures, rective actions prior to the conduct of testing where practical, and inspections, any integrated and leak to the Commission within 90 local rate test. Results of the inspection days of 6ompletion.

shall be reported I21 3 2

A, IA 1 S Reactor Building Modifications affecting the Reactor replacement of components Any major modification or by either an integrated leak rate test shall be followed Building integrity meet the acceptanqe as appropriate, and shall or a local leak rate test, 4.4.l.2.3. respectively.

criteria of 4.4.1.1.4 and Bases a

pressure of 59 psig and Reactor Building is designed for 0 an internalto initial operation, the con The Prior steam-air mixture temperature of 286 F.

tainment is strength tested at 115 percent of design pressure and leak rate tested prior to The containment is also leak pressure.

tested at the design pressure. of the design These at approximately 50 percent satisfies initial operation pressurization rate from Reactor Building tests verify that the leak the specification.

the relationships given in during unit life of a periodic integrated leak rate test The performance the containment, in a current assessment of potential leakage from provides of the containment.

would pressurize the interior case of an accident that of the integrity of the containment a realistic appraisal In order to provide without pre conditions, this periodic.test is to be performed under accident isolation or leak repairs, and containment liminary leak detection surveys manner. The test pressure of 29.5 psig be closed in the normal valves are to high to provide leak rate test is sufficiently for the periodic integrated duplicates the preoperational of the leak rate and it an accurate measurement a relationship for The specification provides leak rate test at 29.5 psig. potential leakage at of air at 29.5 psig to the relating the measured leakage leak rate test is normally of the periodic integrated 59 psig. The frequency these tests can best refueling schedule for the reactor, because keyed to the shutdowns.

be performed during refueling is based on frequency of. periodic integrated leak rate tests The specified leaks in the First is the low probability of three major considerations.

~42,2, 1975 4.4-4

its significance to the load-carrying-capability of the structure. The sheathing filler will be sampled and inspected for changes in physical appearance.

Wire samples shall be selected in such a manner that kiLil the third inspection, wires from all nine surveillance tendons shall have been inspected and tested.

4.4.2.2 Inspection Intervals and Reports For Unit 1, the initial inspection shall be within 18 months of the initial Reactor Building Structural Integrity Test. The inspection intervals, measured from the date of the initial inspection, shall be two years, four years and every five years thereafter or as modified based on experience. For Units 2 and 3 the inspection intervals measured from the date of the initial structural test shall be one year, three years and every five years thereafter or as modified based on experience. Tendon surveillance may be conducted during reactor operation provided design conditions regarding loss of adjacent tendons are satisfied at all times.,

A quantitative analytical report covering results of each inspection shall be ,

submitted to the Coummission within 90 days of completion, and shall especiall; 2i address the following conditions, should they develop:

a. Broken wires.

b. The force-time trend line for any tendon, when extrapolated, that extends beyond either the upper or lower bounds of the predicted design-band.

c. Unexpected changes in corrosion conditions or sheathing filler properties.

4.4.2.3 End Anchorage Concrete Surveillance

a. The end anchorages and adjacent concrete surfaces of the surveillance tendons will be inspected. In addition, other locations for surveillance will be determined by information obtained from design calculations, pre stressing records, observations, and deformation measurements made during prestressing.

b. The inspection interval will be approximately one-half year and one year after the operation of the unit and will occur during the warmest and coldest part of the year.

c. The inspections made shall include:

(1) Visual inspection of the end anchorage concrete exterior surfaces.

(2) A determination of the temperatures of the liner plate area or con tainment interior surface in locations near the end anchorage concrete under surveillance.

(3) Measurement of concrete temperatures at specific end anchorage concrete surfaces being inspected.

4.4-7 D 2 975

crack patterns.

(4) The mapping of the predominant visible concrete The measurement of the crack widths, by use of optical comparators (5) or wire feeler gauges.

The measurement of movements, if any, by use of demountable mechanical (6) extensometers.

compared with those to which

d. The measurements and observations shall be in normal and abnormal load prestressed structures have been subjected measuremehts and observations at conditions and with those of preceding the same location on the reactor containment.

e. The acceptance criteria shall be as follows:

are favorable in compari If the inspections determine that the conditions close inspections will be termi son with experience and predictions, the If the in the schedule.

nated by the last of the inspections stated or movements, normal cracking inspections detect symptoms of greater than to determine the cause.

an immediate investigation will be made

f. Results of the inspection shall be reported to the Commission days of completion.

within 90

!i I21 4.4.2.4 Liner Plate Surveillance The liner plate will be examined prior to the initial pressure 4.4.2.4.1 test in accessible areas to determine the following:

a. Location of areas which have inward deformations. The shall be measured and magnitude of the inward deformations recorded. These areas shall be permanently marked for future reference and the inward deformations shall be measured between the angle stiffeners which are on 15-inch centers. The measurements shall be accurate to + 0.01 inch. Temperature readings shall be obtained on both the liner plate and outside containment wall at the locations where inward deformations occur.

b. Locations of areas having strain concentrations by visual examination with emphasis on the condition of the liner surface. The location of these areas shall be recorded.

4.4.2.4.2 Shortly after the initial pressure test and approximately one year after initial startup, a re-examination of the areas located in Section 4.4.2.4.1 shall be made. Measurements of the inward deformations and observations of any strain con centrations shall be made.

exceeds 4.4.2.4.3 If the difference in the measured inward deformations 0.25 inch (for a particular location) and/or changes in strain The concentration exist, an investigation shall be made.

action.

investigation will determine any necessary corrective 4.4-8 L 2 2 '1975

4.4.2.4.4 The surveillance program shall be discontinued after the one year after initial startup inspection if no corrective action was needed. If corrective action is required, the frequency of inspection for a continued surveill..nce program shall be determined.

4.4.2.,4.5 Results of the surveillance shall be reported to the Com mission within 90 days of completion.

203 "'

2 1 .,

Bases Provisions have been made for an in-service surveillan e program, covering the first several years of the life of the unit, intended to provide suf ficient evidence to maintain confidence that the integrity of the Reactor Building is being preserved. This program consists of tendon, tendon anchorage and liner plate surveillance.

To accomplish these programs, the following representative tendon groups have been selected for surveillance:

Horizontal - Three 1200 tendons comprising one complete hoop system below grade.

Vertical - Three tendons spaced approximately 1200 apart.

Dome - Three tendons spaced approximately 120 apart.

The inspection during this initial period of at least one wire from each of the nine surveillance tendons (one wire per group per inspection) is con sidered sufficient representation to detect the presence of any wide spread tendon corrosion or pitting conditions in the structure. This program will.

be subject to review and revision as warranted based on studies and on results obtained for this and other prestressed concrete reactor buildings during this period of time.

REFERENCES (1) FSAR Section 5.6.2.2 4.4-9 L 221 975

4.4.3 Hydrogen Purge System Applicability Applies to testing Reactor Building Purge System.

Objective To verify that this system and components are operable.

Specification 4.4.3.1 Operating Tests An in-place system test shall be performed annually. This test shall consist of a visual inspection, hook-up of the system to one of the three reactor buildings, a flow measurement using flow instruments in the portable purging station and pressure drop measurements across the filter banks. Flow shall be design flow or higher, and pressure drops across the filter bank shall not exceed two times the pressure drop when new. Fan motors shall be operated continuously for at least one hour, and valves shall be proven operable. This test shall demonstrate that under simulated emergency conditions the system can be taken from storage and placed into operation within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

4.4.3.2 Filter Tests Annually, leakage tests using DOP on HEPA units and Freon-112 (or NJ equivalent) on charcoal units shall be performed at design flow on the filter. Removal of 99.5% DOP by each entire HEPA filter unit and removal of 99.0% Freon-112 (or equivalent) by each entire charcoal absorber unit shall constitute acceptable performance.

These tests must also be performed after any maintenance which may affect the structural integrity of either the filtration system units or of the housing.

4.4.3.3 H2 Detector Test Hydrogen concentration instruments shall be calibrated annually with proper consideration to moisture effect.

Bases The purge system is composed of a portable purging station and a portion of the Penetration Room Ventilation System. The purge system is operated as necessary to maintain the hydrogen concentration below the control limit.

The purge discharge from the Reactor Building is taken from one of the Penetration Room Ventilation System penetrations and discharged to the unit vent. A suction may be taken on the Reactor Building via isolation valve PR-7 (Figure 6-5 of the FSAR) using thu existing vent and pressurization connections.

a.

4.4-10 Dt.i 29 1975

/4.13 FUEL SURVEILLANCE Applicability Applies to the fuel surveillance program for fuel rods of Unit 1.

Objective To specify the fuel surveillance program for fuel rods.

Specification 4.13.1 Visual Inspection Two (2) Oconee Unit 1 fuel assemblies will be designated for visual inspection. These same assemblies will be inspected during each of the first three refuelings of Unit 1. Underwater viewing devices will be used to determine that the fuel rods have maintained their structural integrity.

4.13.2 Dimensional Examination Measurements of the length and outside diameter will be made on selected peripheral rods of the following fuel assemblies of the first core of Unit 1 both prior to operation and at the times specified:

a. One assembly after the first cycle.

b. Four assemblies after the second cycle.

c. Two assemblies after the third cycle.

4.13.3 Results of the fuel surveillance program shall be submitted to the,, 2° Commission within 90 days of completion of the program. 1 9 Bases This fuel surveillance program provides substantiating information for the first core in the present generation of B&W reactors. It provides for examination of fuel rods at the end of the first, second, and third cycles of Unit 1 to determine if fuel rods have maintained their integrity and to determine the extent, if any, of dimensional changes in diameter and length.

4.13-1

~I-~

1975

c. Quorum The chairman plus two members shall constitute a quorum.

d. Responsibilities The committee shall have the following responsibilities:

1. Review all new procedures or changes to existing proc dures determined by the station Manager or his designate to affect ope ational safety.

2. Review station operation and safety considerations.

3. Review reportable occurrences and violations of Techni al Specifica tions and make recommendations to prevent recurrence. I G /3i

4. Review all proposed tests that affect nuclear safety or radiation safety.

5. Review proposed changes to Technical Specifications and safety-related changes or modifications to the station design.

e. Authority The Station Review Committee shall make recommendations to the station Manager regarding Specification 6.1.2.1-d.

f. Records Minutes of all meetings of the committee shall be-kept at the station, and copies shall be sent to the station Manager, Vice President, Steam Production, and the chairman of the Nuclear Safety Review Committee.

6.1.2.2 Nuclear Safety Review Committee

a. The Executive Vice President and General Manager shall appoint a Nuclear Safety Review Committee having responsibility to verify that operation of the station is consistent with company policy and rules, approved operating procedures, and license provisions; to review important pro-posed station changes, and tests; to verify that abnormal occurrences and unusual events are promptly investigated and corrected in a manner which reduces the probability of recurrence of such events; and to detect trends which may not be apparent to a day-to-day observer.

b. The activities of the Nuclear Safety Review Committee shall be guided by a written charter that contains the following:

Subjects within the purview of the committee Responsibility and authority Mechanisms for convening meetings Provisions for use of specialists or subgroups 6.1-2 Lt 2 1975

f. Meetiig Frequency:

The committee shall meet at least three times per year at intervals not to exceed five months and as required on call by the chai-man. During the period of initial operation, this committee shall meet at least once per calendar quarter.

g. Quorum:

The chairman or vice-chairman plus three members, or appointed alternates, shall constitute a quorum. No more than a minority of the quorum shall have direct line responsibility for station operation.

h. Meeting Minutes:

Minutes of all scheduled meetings of the committee shall be prepared and shall identify all documentary materials reviewed. These minutes shall be formally approved, retained, and also promptly distributed to the Executive Vice President and General Manager; Senior Vice President, Engineering and Construction; Senior Vice President, Production and Trans mission; Vice President, Design Engineering; Vice President, I 3/2 i/ v Steam Production; and station Manager. A copy of these minutes shall be kept on file at the station.

i. As a safety review to the normal operating organization, the committee shall review the following:

1. Proposed tests and experiments, and results thereof, when these con stitute an unreviewed safety question defined in 10CFR50.59.

2. Proposed changes in equipment or systems which constitute an unreviewed safety question defined in IOCFR50.59, or which are referred by the operating organization.

3. All requests to the NRC/DRL for changes in Technical Specifications or license that involve unreviewed safety questions as defined in IOCFR50.59.

4. Violations of statutes, regulations, orders, Technical Specifications, license requirements, or internal procedures, or instructions having safety significance as determined by the NSRC.

5. Reportable Occurrences as defined in 6.6.2.1 of these specifications. I //

Special reviews or investigations as required by the Vice President Iz 6.

-President, Steam Production, or the station Manager.

6.1-4 BL" '4'2 c.'197 5

6.2 ACTION TO BE TAKEN IN THE EVENT OF.A REPORTABLE OCCURRENCE I i:,

4 /:

6.2.1 Any reportable occurrence shall be investigated promptly by the station Manager.

6.2.2 The station Manager shall promptly notify the Vice President, Steam Production, of any reportable occurrence. I The Station Review Committee shall review a written report which shall describe the circumstances leading up to and resulting from the occurrence and shall recommend appropriate action to prevent or minimize the probability'of a recurrence.

6.2.3 The Station Review Committee report shall be submitted to the Nuclear Safety Review Committee for review of any recommendations.

Copies shall also be sent to the station Manager and the Vice President, Steam Production.

6.2-1 2 ,? 1975

6.6 STATION REPORTING REQUIREMENTS 6.6.1. Routine. Relorts The following reports shall be submitted to theDirector, Office of Inspection mnd Enforcement Region II, Atlanta, Georgia.

6.6.1.1 Startup Report A summary report of unit startup and power escalation g shall be submitted following (1) receipt of an operating lice.ns , (2) amendment to the facility license involving a planned increase in p wer level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the unit. Startup reports shall be submitted (1) within 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) nine months following initial criticality, whichever occurs first. If a startup report does not cover all three events, i.e., initial criticality, completion of the startup test program and re sumption or commencement of commercial power operation, supplementary reports shall be submitted at least every three months until all three events are completed.

6.6.1.2 Annual Operating Repor, Routine operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to April I of each year.

The initial report shall be submitted prior to April 1 of the year following initial criticality.

Each annual operating report shall provide the following:

a. Operations Summary (1) A narrative summary of operating experience during the report period relating to safe operation of the facility, including safety-related maintenance not covered in 6.6.1.2.a(2e)

(2) For each outage or forced reduction in power- of over 20 percent of design power level where the reduction extends for greater than four hours.

1/The term "forced reduction in power" is defined as the occurrence of a component failure or other condition which requires that the load on the unit be reduced for corrective action immediately or up to and including the very next weekend. Note that routine preventive maintenance, sur veillance and calibration activities requiring power reductions are not covered by this section.

6.6-1 Entire Page Revised 2 9~ 1975

1__ý (a) the proximate cause and the system and major component involved (if the outage or forced reduction in power involved equipment malfunction);

(b). a brief discussion of (or reference to fep~os of) any reportable occurrences pertaining to the outage or power reduction; (c) corrective action taken to reduce the probability of recurrence, if appropriate; (d) operating time lost as a result of the 6utage or power reduction (for scheduled or forced outages,2/ use the generator off-line hours; for forced reductions in power, use the approximate duration of operation at reduced power);

(e) a description of major safety-related corrective maintenance performed during the outage or power reduction, including the system and component involved and identification of the critical path activity dictating the length of the outage or power reduc tion', and (f) a report of any single release of radioactivity or unusual radiation exposure specifically associated with the outage which accounts for more than A0 percent of the allowable annual values.

b. Changes, Tests and Experiments A brief description and the summary of the safety evaluation for those changes, tests, and experiments carried out without prior Commission approval pursuant to the provisions of IOCFR50.59.

3/

c. Reporting of Radioactive Effluent Releases Data shall be reported to the Commission in a form similar to that shown in Table 6.6-1 and shall include the following:

(1) Gaseous Releases (a) Total radioactivity (in curies) releases of noble and activation gases.

(b) Maximum noble gas release rate during any one-hour period.

(c) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed.

2/The term "forced outage" is defined as the occurrence of a component failure or other condition which requires that the unit be removed from very service for corrective action immediately or up to and including the next weekend.

3/ Shall be reported on a semi-annual basis.

6.6-2 Entire Page Revised 2 17

(d,) Percentage applicable limits released.

(2) l*ddi ne R, Icases (a) Total 1-131, 1-133, 1-135 radioactivity- (in curies) released.

(b) Total radioactivity (in curies) released, *by nuclide, based on representative isotopic analyses performed (c) Percentage of limit.

(3) Particulate Releases (a) Gross radioactivity (ý-y) released (in curies) excluding back ground radioactivity.

(b) Gross alpha radioactivity released (in curies) excluding back ground radioactivity; (c) Total radioactivity released (in curies) of nuclides with half lives greater than eight days.

(d) Percentage of limit.

(4) Liquid Releases (a) Gross radioactivity (ý-y) released (in curies) excluding tritium "and average concentration released to the unrestricted area at the Keowee Hydro unit.

(b) The maximum concentration of gross radioactivity (B-Y) released to the unrestricted area (averaged over the period of release).

(c) Total tritium and alpha radioactivity (in curies) released and average concentration released to the unrestricted area at the Keowee Hydro unit.

(d) Total dissolved gas radioactivity (in curies) and average con centration released to the unrestricted area at the Keowee Hydro unit.

(e) Total volume (in liters) of Keowee Hydro liquid waste released.

(f) Total volume (in liters) of dilution water used prior to release from the restricted area.

(g) Total radioactivity (in curies) released, by nuclide, based on representative isotopic analyses performed.

(h) Percentage of limit for total activity released.

Entire Page Revised 6.6-3 DL, 2 21975

(5) Solid Waste (in cubic feet).

(a) The total amount of solid waste packaged Estimated total radioactivity (in curies):

(b) destination if shl'Jed off site.

(c) Disposition including date and (6) Environmental Monitoring reporting period, the (a) For each medium sampled during the following information shall be provided.

1. Number of sampling locations.

2. Total number of samples.

are found to be sig

3. Number of locations at which levels nificantly greater than local backgrounds.

4. Highest, lowest, and the average concentrations or levels of radiation for the sampling point with the highest average and description of the location of that point with respect to the site.

(b) If levels of station-contributed radioactive materials in en vironmental media indicate the likelihood of public intakes in excess of 3 percent of those that could result from continuous exposure to the concentration values listed in Appendix B, Table II, Part 20, estimates the likely resultant exposure to individuals and to population groups, and assumptions upon which estimates are based shall be provided. (These values are com parable to the top of Range I, as defined in FRC Report No. 2.)

(c) If statistically significant variations in off-site environmental concentrations with time are observed and are attributed to station releases, correlation of these results with effluent releases shall be provided.

d. Personnel Exposure and Monitoring A tabulation (supplementing the requirements of 10 CFR 20.407) of the number of personnel receiving exposures greater than 100 mrem in the to reporting period and their associated man-rem exposure, according duty function, e.g., routine plant surveillance and inspection (regular duty), routine plant maintenance, special plant maintenance (describe maintenance), routine fueling operation, special refueling operation (describe operation), and other job-related exposures.

e. Fuel Examinations includ Indication of failed fuel resulting from irradiated fuel examinations, examinations ing results of eddy current tests, ultrasonic tests, or visual completed during the report period.

6.6-4 Entire Page Revised D 2 1975

6.,6.2 'Non-Routine Reports 6.6. 2-. 1 Reportable Occurrences

a. Prompt Notification with Written Fo0l]owu0 wit.ion 24. hours of The types of events listed below shall be reportced. transmiission or fhCIi discovery (by telephone, telegraph, mailgram, and Enforce;-;. nt, Region I1, or hiis to the Director, Office of Inspection within tLxý'o weeks to the Director, designate) with a writtenfl follox.,up repo-rt IT (c*)y to the Director, Office Office of Inspcction and 3>-forcemnnt, Region USNRU).

of Management Information and Program Control, required, when Protective System to trilp,.as (l) Failure of the Reactor limiti; the setpoint specified as the a monitored parameter reaches Technical Specifi.cation)s.

safety sysiem setting in the parameter or or affected systems w.heicn any (2) Operation of the unit condition for operation i less operation subject to a limit"ing of the ]i.miting conservXative aspect conservative' than the least the Technical Specirficat iols.

in condition for operation established in fuel claddin g, reactor cbolant (3) Abnormal degradation discovered containment.

pressure boundary or priTiary rcC, 't with prcdj.ctea Val Of di, (4) Reactivi-ty ano)Tal ies involving teY than or steady-state con".Ii ens erea reactivity balance under . 'cf Ci~t--iVUt .flU[)yC J% kLK/z a c~L-luuatcýC" specif<ct i 'ns; equal to specifiLed in the technical margin less conservative than short-terlm reactivity increases correol.pfld to a renct-or per o2 reactivity of less thn ,5 seconds, or if subcritica] aln unplanned Ak/M.; or any urpianned criticality.

insertion of uore than 0.5Z or or tore componcnts which prevents (5) Failure or mal]function of one functional require fulfillment of th.-

could prevent, by itself, the the to cope with accidents analyzed in ments of systems required Safety Analysis Report.

or could inadequacy which prevents (6) Personnel error or procedural requirements of prevent. by itself, the fulfil.*,enlt of the functional Safety a*cidents analyzed in the systems required to cope with Analysis Report.

that, as a direct natural or ran-made events (7) Conditions arising from of safety unit shutdown, operation result of the event, require by Technical Specifi measures requi-red systems, or other protective cations.

or in the transient or accident analyses (8) Errors discovered in the Analysis as described in the Safety methods used for such analyses that have or the Technical Specifications Report or in the bases for operation in a manner less conservative could have permitted reactor than assumed in the ana*lyses.

Entire Page Rexvised 6.6-5

b, .hi , Written Reports the subject of written reports The types of events listed below shall be and Enforcement, Region II, within to the Director, Office of Inspection (Copy to the Director, Office of Manage 30 days of discovery of the event..

USNRC).

ment Information and Program Control, instrument or engineered safety feature (1) Reactor protection system t*an those

.stablislhcd be less conservacive settings which are found to the fulfill but which do not Prevent by the technical specifications of affecteS system..

ment of the functional requirements in a degraded n de perlatited by a (2) Conditions leading to operation li-inting or s.hutdoown required by a limiting condition for operation condition for operation.

or implemeutation of administrative (3) Observed inadequacies in the could cause operation of a unit Which procedural coni-rols durin-, in the, Reactor Protective provided reduction of degree of redundancy Feature Systems.

System or Engineered Safety Environmental ,onitorii,g 6.6.2.2 picocuries per 1-131 co-ncentrationrs of 10

a. If individual milk samples shov one w..eek A$lvising the be submitted within liter or greater, a plan shall annual doses willI to ensure the plant rel.ated NRC of the proposed action of 15 mrne/yr to the thyroid of any indi be within the design objective vidual.

a calendar quarter show average concentr7atJions

b. if milk samples collected over greater, a plan shall be submitted within 30 liter or of 4.8 picocuries per the plant related the NRC of the proposed action to ensure days advising mrem/yr to the doses will be within the design objective of 15 annual thyroid of any individual.

radioactivity period, a measured level, of

c. If, during any annual report associated with gaseous other than those in any environmental medium value, a ten timt.s the control station radioiodine releases exceeds advising the NRC be submitted within one week written notification will an evaluation of any of this condition. This notification should include necessary to factors, or other aspects release conditions, environmental explain the anomalous result.

radioact ivity period, a" measured level of

d. If, during any annual report with gaseous other than those associated in any environmental medium control station value, a releases exceeds four times the radiKodine advising the NRC of notification wil be submitted within 30 days of any Wxritten should include an evaluation this condition. This notification necessary to environmental factors, or other aspects release conditions, explain the anomalous result. D*o*1975 Entire Page Re0vised 6.6-6

6'.6.3 Special Reports Special reports shall be submitted to the Director, Office of Inspection anid En forcemenL, Region "i, within the time period specifi ed 'or each report. T'hiese re ports shalI be submitted covering the activities identif I d I)elow p1urstuint to the requirements of the applicable reference specification

a. Electrical System Degradation, Specification 3.7.

b. Excessive Liquid Waste Releases, Specification 3.9ý

c. Excessive Gaseous Waste Releases, Specification 3

• 10.

d. Inservice Inspection, Specification 4.2.4.

t

e. Reactor Vessel Specimen Surveillance, Specification 4.2.8.

f. Containment Integrated Leak Rate Test, Specificatiot 4.4.1.1.7.

g. Reactor Building Annual Inspection Report, Specification 4.4.1.4.

h. Tendon Stress Surveillance, Specification 4.4.2.2.

i. End Anchorage Concrete Surveillance, Specification 4.4.2.3.

j. Liner Plate Surveillance, Specification 4.4.2.4.

k. Single Loop Operation, Specification 3.1.8.

1. Fuel Surveillance Program, Specification 4.13.

6.6-7 Entire Page Revised DL6 2 2 1975

DU*E POWER COMIPANY TABLE 6.6-1 OCONEE NUCLEAR STATION REPORT OF RADIOACTIVE EFFLUENTS ONS-S/A-07 Year I.

(

TABLE 6.6-1 (CONTINUED)

DURE POWFR COMP'ANY OCONEE NUCLEAR STATION REPORT OF RADTOtCTIVE EFFLUENTS ONS-S/A-0S Year II. Airborne Releases____

r i t-s UJ~ Ja . F eb . Ma r . Apr . Ma v J u ne _ _ _A_ u __ __ __ N o v ._ D e c._ TO T A L 1

2. Total halo ens 3,._Total. nobtclae jcross radio-Curies Curios jOc.

___ _ _ __ _ _ ___1___ ___ __ _ ___ ___

ov

__ITOA

15. Total particulate gross al-oha

6. Maximumn noble gaýs release rate - ________ ____ ___

7.Perccnt of eaplicable. limit for:_ ___ _______ ___ ____ ___ ___ ______

a. nob le gases f~~4___________

T i II___

b. haloeens %____ _ _ _ _ _ _I_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

c.particulatcs ____ ____________ ___ ___

18. Isotopc reluased: -~Curies ____]I____ ___ ___ t

___ x- a-140 I _ _ _ _ _

Sr 90 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _14

__ _ _ _ _ _ _ I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

HaI_ l3 _ _ __ __ _ __ 4}__ _ _ __ _ __I__

1-131 _ _ _ _ _ _ _ _I_ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _

1-133a______4 I 1_____________1 5__ ____

TG3ases_1___r __ _ __{___.1 __

I K __ _ _ _ -__ _I__ _I__ _ _ _

Gee K _____ ___ T t____ ___ ________ ________ ___

I

__ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ _ _ _ _ _ _ __ _ _ _ _ __I_ _ _ _ _ _ __ _ __ _ _ _ _ _ _ _ _ _

_ ____ 1 _ _ 1_r--------r _ j _ _ _ _

(0

UNITED STATES NUCLEArv REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. i G TO FACILITY LICENSE" NO. DPR-38 CHANGE;NO.2 G TO TECHNICAL SPECIFICATIONS; AMENDMENT ýO. 1 G TO FACILITY LICENSE NO. DPR-47 CIHANGENO.2 I TO TECHNICAL SPECIFICATIONS; AMENDMHENT TO 1 ' TO FACILITY LICENSE NO. DPR-55 CHANGE NO.1 ,, TO TECHNICAL SPECIFICATIONS DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-269, 50-270, AND 50-287 Introduction By letter dated January 15, 1975, Duke Power Company (the licensee) requested a change in the Technical Specifications of Licenses No. DPR-38, DPR-47, and DPR-55 for the Oconee Nuclear Station, Units 1, 2, and 3. The proposed amendments would. modify the station reporting requirements and delete the definition of an abnormal occurrence.

Discussion The proposed changeswould be administrative in nature and are intended to provide uniform license requirements. In Section 208 of .the Energy Reorganization Act of 1974 "abnormal occurrences" is defined as an unscheduled incident or event which the Commission determines is significant from the standpoint of public health and safety. The term "abnormal occurrence" is reserved for usage by NRC. Regulatory Guide 1.16, "Reporting of Operating Information Appendix A Technical Specifications", Revision 4, enumerates required reports consistent with Section 208. The proposed change to required reports identifies the reports required of all licensees not already identified by the regulations and those unique to this facility.

The proposal would formalize present reporting and would delete any reports no longer needed for assessment of safety related activities.

Evaluation The new guidance for reporting operating information does not identify any event as an "abnormal occurrence." -T-he proposed reporting requirements also delete reporting of information no longer required and duplication of reported information. The standardization of required reports and desired format for the information will permit more rapid recognition of potential problems.

During our review of the proposed changes, we found that certain modi fications to the proposal were necessary to have conformance with the desired regulatory position. These changes were discussed with the licensee and have been incorporated into the proposal.

We have concluded that the proposal as modified improves the licensee's program for evaluating plant performance and the reporting of the operating information needed by the Commission to assess safety related activities and is acceptable. The modified reporting program is consistent with the guidance provided by Regulatory Guide 1.16, "Reporting of Operating Information - Appendix A Technical Specifications",

Revision 4.

Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the change does not involve a significant increase in the probability or consequences of accidents previously consider~d and does not involve a significant decrease in a safety margin, the change does not involve a significant hazards consideration, (2) there is reasonable assurance that thehealth and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulctions and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Date: 2JL 22 1975

11

- - - 1

4. / 9; 7

' -Cd 1- i' 5

-- S Y*

, I

• ROUT~111'G AND TRAt-S!MITTAL SUP O me trico sym~bol or 1ocotfion) IN ITIF A I 'C WC IJU 01311 - f/con clirrncesDA DLZS omann f/signaturcs i.

Reoba -for final checks,~A~

Attached for )'our concurrence arc five packagecs ())rcsdc I Stati on, Quad Cities Station, Cooper, Pilgrim rind Calvert Cl iffs) of ni no frmi ORB AWN>Lc 3 'nC i C ,ýI

-F P

also revises, the entiire admni3strative conatrols sectiun 1t is roques ted that, in the interest of review con sist ency, those packages (and the 4 future. recporting, rcq~uirce..sonts pa!ckage-s) be0 assigned to one OELLI no Questio--ns. may be di rec ted to0 the PM,"for the1 parti cul ar 4Ž case or to Mie Fletcher,. coordinator for reportingj"W"

/

I) C.

Do NOT use UhS fwal as n OE 1LJ0 MnspprovAls, clcaanccs, anj SitrillaC actions'-~ C(~lC5

/.

TI"oTM (WCM0. MU CyCF or cauonh )

11 -3 -75 * ,Ij. '7 1)LZ i~7380 OPTIONAL FOPM 'I c4S-10-61594-I 6527-13 GPO $01-101 AUGUST 1067 GSA FPMR ( 41CFR) 100-11.206 1,

. . P

• 4

- - 4..-

I -

UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NOS. 50-269, 50-270 AND 50-287 DUKE POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FACILITY OPERATING LICENSES Notice is hereby given that the U.S. Nuclear Regulatory Commission (the Commission) has issued Amendments No. f G, . 0, ana I ato Facility Operating Licenses No. DPR-38, DPR-47, and DPR-55, respectively, issued to Duke Power Company which revised Technical Specifications for operation of the Oconee Nuclear Station, Units 1, 2, and 3, located in Oconee County, South Carolina. The amendments are effective January 1, 1966.

These amendments revise the provisions in the Technical Specifications relating to Reporting Requirements.

The application for the amendments complies with the standards arid requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments. Prior public notice of these amendments is not required since the amendments do not involve a significant hazards consideration.

For further details with respect to this action, see (1) the appli cation for amendments dated January 15, 1975, (2) Amendments No. 1,

O F F IC E - " .... ... .... ... .... ... ....... .... ... ,. .. . . . .. . . .. . . . . . . . . . .. . .. . . . . . . . . . . .. . ............... ............................. ...............

S U R N A M E 0 DA .(E . ................... ........................................

DA YrEO * " .......' .. .. .......... ........ .... m..............................................  !.... ...................................... m....................... ............................. .............................................. F .................................

Form AXC-318 (Rev. 9-53) A.E(CM 0240 *r u. 9,; aOVikmcNrINI PRINTINGI OFFICM 1 97AL-526-166

7

- 2 andil Ito Licenses No. DPR-38, DPR-47, and DPR-55, with Changes No. 2 6 2 1, andl , and (3) the Commission's related Safety Evaluation. All of these items are aviilable for public inspection at the Commission's Public Document Room, 1717 H Street, NW., Washington, D.C. and at the Oconee County Library, 201 South Spring Street, Walhalla, South Carolina 29691.

A copy of items (2) and (3) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attention: Director, Division of Reactor Licensing.

Dated at Bethesda, Maryland, .this DE C 2 2. 175 FOR TIHE NUCLEAR REGULATORY COMMISSION Original sga~d by R. A. Purpe.

Robert A. Purple, Chief Operating Reactors Branch #1 Division of Reactor Licensing

/ DRL:ORB#l "G ec :S OELDD dtd.11/.3/7 no-t*-e..R...7.5 . ............ui.......................

I..................

DA 2/

.3 /75 12/...5 _..J.zl ...............

,,,,I,, - 2. .R.9/

.... .).....0 . ........................................

U.. I N.....

Form AJ[r-318 (Reev. 9-53) A.EC]M[ 0240 "*u; 81 QOVKENM1%Pr PRINTrING OFF'ICi81 1974-526-166