ML19316A237
| ML19316A237 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/29/1973 |
| From: | Patricia Hall, Mathews D, Miller A DUKE POWER CO. |
| To: | |
| References | |
| PROC-730329, NUDOCS 7912030352 | |
| Download: ML19316A237 (50) | |
Text
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EO-287 l THERMAL MOVEMENT MONITORING PROGRAM ' , _ _ ,
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FOR OCONEE 1-2-3 REACTOR COOLANT SYSTEM -~
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y,,p,- ..- ) w :,. ) i DUKE POWER COMPANY . ENGINEERING DEPARTMENT MECHANICAL DESIGN SECTION, MECHANICAL DIVISION CHARLOTTE, NORTH CAROLINA f E b E b $..d.Y-du[ C$yT]jl {}{{g FILE NO. 05-278 s 3 'y . e DUKE POWER x r l 000R B3$b1 I i
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THERMAL MOVEMENT MONITORING PROGRAM FOR OCONEE l-2-3 REACTOR COOLANT SYSTEM DUKE POWER COMPANY DESIGN ENGINEERING DEPARTNENT MECHANICAL DESIGN SECTION, hi'HANICAL AND NUCLEAR DIVISION CHARLOTTE, NORTH CAROLINA FILE 05-278 Duke Power Conoany 0uali tv Assurance Statement: This cccument includes engineering design of items related to nuclear safety, in ac:ordance with established procedures. Its quality has been assured.
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__ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . . . TABLE OF CONTENTS S,ection Title Pajun 1 i INTRODUCTION AND PURPOSE lNTER-COMPANY DIVISION OF RESPONSIBILITY 2 ll Ill DETERMINATION OF MONITORING POINT LOCATIONS 3 INSTALLATION METHODS FOR MONITORING POINT 4 IV HARDWARE SYSTEM TEST SEQUENCE AND DESCRIPTION 5 V 6 VI ENGINEERING LIMITATIONS AND PRECAUTIONS Vil DATA RECORDED 7 Vill COMPARISON OF ANALYTICAL AND TEST RESULTS 8 IX CONCLUSIONS 9 X APPLICABILITY OF TEST TO UNITS 2 & 3 11 APPENDIX - 4 m e-- = , e - -- -+
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- l. INTRODUCTION AND PURPOSE As the result of discussions in the 124TH Duke-ACRS Meeting held on August 13, 1970, Duke Power Company agreed to monitor the thermal expansion movement of the Unit 1 Reactor Coolant System #50. The purpose of the monitoring program is two-fold:
- 1) To determine the movement of the system piping during the hot functional test period as described in Section V of this report.
- 2) To review recorded data to assure that. thermal stresses and movements within the system are compatible with analytical predictions.
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. . l II. INTER-COMPANY DIVISION OF RESPONSIBILITY , Responsibility for the implementation of the monitoring program is shared by the Duke Power Company Design Engineering, Constructime and Steam Pro-duction Departments as outilned below.
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- 1) Design Engineering Department
- a. Establish need for program as described in 1. of this report end Appendix A
- b. Citablish monitoring point locations based on analytical results ;
and as described in Section lli of this report
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- c. Establish installation methods of monitoring devices as described in Section IV of this report
- d. Establish Engineering limitations and precautions as described in Section VI of this report l
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- e. Establish data to be recorded during test as described in Sec-tion VII of this report
- f. Make engineering comparison and evaluation of analytical versus actual test data as described in Section Vill of this report
- g. Define acceptability of results as described in Section IX of this report
- h. Describe applicability of test results with respect to Unit 2 and 3 as described in Section X of this report
- i. Preparation of report for AEC review.
- 2) Construction Department
- a. Installation of monitoring point hardware at locations described in Section ill and in accordance with methods described in Sec-tion IV of this report
- b. Assist Steam Production Department in the surveillance and record-Ing of data as requested
- 3) Steam Production Department
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Develop a procedure for monitoring the thermal expansion of the Reactor Coolant System as required by Design Engineering , Appendix A
- b. Coordinate and schedule with the Construction Department as necessary for the installation of monitoring hardware and data-recording
- c. Documentation of monitoring program results in accordance with Oconee Nuclear Station Test Procedure TP i A 60014 5 (See Appendix A)
- d. Transmittal of monitoring data to the Engineering Department for i cvicw anil approval and subsequent preparation of Thermal Monitoring R e p. . t
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Ill. DETERMINATION OF MONITORING POINT LOCATIONS
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Monitoring points in general are located at positions undergoing measurable expansion movements shown on Figure 1 and 2 in accordance with B & W analy-tical predictions. Specifically, monitoring point locations are defined on Drawing PO-187 (Figure 3) and as listed herein. 1 1 Monitoring Point Data Required Attachment Location & Quantity b i Movement of RC pumps in On support lug located 180* the X, Y & Z directions from pump 1A1 and 1A2 dis-charge at Elevation 813'-0" (one required x's 2 pumps) 2 OTSG X & Y movement at On east side centerline of top of RC inlet nozzle weld joint at nozzle (one required) 3 OTSG X & Y movement at On outboard side of either feedwater header leg of feedwater inlet piping i' on centerline elevation of butt weld to feedwater ring i header nozzle (one required) 4 OTSG Y movement at upper On 1 1/2" vent and level tube sheet sensing connection at top of tube sheet at Elevation 842'-2 3/4" on IA generator ~ (one required) 19 Pressurizer surge line On bottom, top, west and east nozzle Y & Z movement centerlines of butt weld to at RC System hot leg con- nozzle. (Four required x's nection on pipe assembly I nozzle.) A-33 20 X, Y & Z RC System hot On north, south, east and west leg movement at top of centerlines of tangent point straight run - 180' Ell weld. (Four required x's one tangent point weld) 21 X Y & Z RC System cold On reessure tap connection leg movement. (Long and of ti.s 1A1 and 1A2 RC pump short RC pump discharge discharge line. (One required line) x's 2 lines) 4
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l IV. INSTALLATION METHODS FOR MONITORING PolNT HARDWARE A description of the hardware and method of attachment of monitoring points hardware is shown on Drawing PO-187 (Figure 3) except for location number 1 which does not require attachments to the piping or vessels. At monitoring location 1, a 1/2" hex nut was welded to a 1/4" plate and bolted to the sup-port lug on the pump casing. All other monitoring locations as' described in 111 above require attachments to the piping or vessels. The welding of attachments directly to piping or vessels was not permissible. Heavy gauge stainless steel banding was used primarily on piping or vessels for the attachment method using a 1/2" stainless steel hex nut welded in position on the band. Target studs protruding thru insulation having a polished end for micrometer readings were held in place by a threaded con-nection with the 1/2" hex nuts. In addition, target studs are marked with location identification if monitoring is ever conducted in the future. Target studs were removed af ter the monitoring program and properly stored.
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V. SYSTEM TEST SEQUENCE AND DESCRIPTION in an effort to meet the Duke Design Engineering Department quality Assur-ance Inspection for harigers, restraints and suppressors as described in FSAR IC.3.7 and obtain data suitable for the preparation of this report, the Oconee Nuclear Station Steam Production Department prepared and executed ' test procedure TP 1 A 600 14 5, " Pipe and Component Hanger Hot Deflection < and Inspection Tett." This procedure is made a part of this report as Appendix A. Duke's Design Engineering Department reviewed and accepted the test proce-dure prior to its initial use. i 1
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VI. ENGINEERING LIMITATIONS AND PRECAUTIONS Any restrictions or deflection limits imposed on the Reactor Coolant System or its hangers, restraints and supports are included in test procedure TP 1 A 600 14 5 and were established in accordance with manu-facturer and Duke Design Engineering Department requirements. Approximate predicted movements for the Reactor Coolant System as listed on Drawing PO-187 (Figure 3) were established as limiting criteria for system movements. Steam Production Department personnel were to check and obtain Design Engineering approval if any of the predicted movements were exceeded during the system heat-up prior to escalating system temperature. 1
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, , ) Vil. DATA RECORDED l The extent of information recorded is described in Test Procedure TP *
- 600 14 5 (Appendix A).
Data recorded either meets or exceeds the information required for this report and Ic in accordance with manufacturer and/or Duke Design Engi-neering Department requirements. In an effort to make this report more meaningful and brief, only the information required to confirm and satisfy the purpose of the report is tabulated in Appendix C along with the predicted analytical Infor-mition for comparative purposes.
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Vill. CDMPARISON OF ANALYTICAL VERSUS TEST RESULTS A comparison of analytical uavements versus the actual measured movements is made for each point of Interest in Appendix C. The comparison is pre-sented in the form of Percentage Difference such that the overall accuracy of readings versus analytical predictions can be reviewed and determined. Review of the data in Appendix C Indicates that the horizontal movement of Reactor Coolant Pump 1A2 exceeded the approximate limiting criteria established on Drawing P0-187 (Figure 3) in one direction. Steam Produc-tion Department contacted the Design Engineering Department once the i variation was noted. Design Engineering made a simple stress calculation, determined that the variation was acceptable, gave the Steam Production Department the maximum limiting criteria and advised that the system
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A review of the variation at monitoring point I has been conducted with results as described in MDS-JR-52 (Appendix D).
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IX. CONCLUSIONS l
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Based on the information presented in Appendix C and as summarlzed in Section Vill of this report, thermal expansion movements of the Reactor Coolant System exhibitled average % Differences to predicted movements as follows:
% Difference % Difference % Difference X(North-South Direction) Y(Vertical Direction) Z(East-West Direction) 31.57% 6.02% 15 98%
for an overall average % Difference of 17.86%. These % Differences would be even smaller if the dimensions being examined were of greater magnitude; l however, % Differences experienced are acceptable based on the small dimensions being evaluated. The following conclusions are made as the result of this study.
- 1) All expansion movements were near the actual predicted movement, indicating actual thermal stresses of the system are in acenrd-ance with the design.
- 2) Dif ferences between predicted movements versus actual moveuents are small enough that they can be accounted for by the possible inaccuracles of the monitoring technique and the effect of piping wall thickness manufacturing tolerances on system flexibility.
- 3) The system behaved as predicted by stress analysis techniques except in the case of Reactor Coolant Pump IA2 movement thereby confirming analytical predictions and actual thermal stresses in the system.
- 4) The variation in the movement of Reactor Coolant Pump 1A2 can be explained in that the pump Inlet piping probably has more flexibility than expected. In addition, the movement experienced between 450*F and 500*F was about twice that of other comparable 50*F temperature increases which would Indicate that the pump seismic restraint might possibly be stiffer than expected in the hinged joints. This
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restraint was examined by Design Engineering and determined to be free of any restrictions to the proper movement. 9
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6 s IX. CONCLUSIONS
- 5) The variation in the movement of Reactor Coolant Pump IA2 is totally acceptable as described in HDS-JR-52 (Appendix D) and any stiffness in the seismic restraint will improve with cycling of the system.
The pump returned back to the original cold position with a good degree of accuracy confirming these conclusions.
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, , X. APPLICABILITY OF TEST TO UNITS 2 & 3 The Reactor Coolant Systems for all three Oconee units are identical in arrangement, in design conditions and in all physical parameters such as piping wall thickness, materials of construction, support and restraint location, etc., that would have a bearing on the thermal stress analysis. Thus, the results of the thermal expansion monitoring test on Unit I are totally acceptable and indicative of results that would be expected from monitoring Units 2 and 3 Since the purpose of this monitoring program was academic in nature, only the Unit i Reactor Coolant System will be monitored. _ 9
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O O APPENDIX A OCONEE NUCLEAR STATION TEST PROCEDURE TP 1 A 600 14 5
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4f..: COi! TROL COPY . . li PIPE & COMPONENT IWiGER HOT DE??ECTION & INSPECTION TEST 9
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OC0hTE. NUCLEAR STATION . 1
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TP 1 A 600 14 5
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- PIPE AND COMPONENT HANGER HOT DEFLECTION AND INSPECTIO!I* TEST -
l 1.0 Purpose ,
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To verify the predicted load and. travel indications on the following hangers and adjust those not within the required margins: Reactor Coolant pump constant' load supports (two per pump). OTSC Auxi-
., liary Feedwater header hangers (three per header). Pressurizer spray ,
-< s line (two hangers). Points listed on P.O. Dwg. 187 (1) (2) 1-21, RC
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pumps LOCA restraints, piping supports, and OTSG supports. *- i
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2.0 References 2.1 Drawing No. 1016470, (B&W) Typical Pump and Motor Support Criteria. 2.2 Dra wing No. 129303E, (OTSG No.1) Drawing No. 146403E (OTSG No. 2) 9l
Cross sections through Steam Generator (B&W). ,,
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- 2.3 Grinnell Co. Instruction Manuals #52, Model R, Fig. 80V-Type E.
Constant Support (RC Pumps) Figure 98 No. 6 and Pig. 98 No. 8 Support Hangers (OTSG Aux. FW Headers). ,
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2.4 B&W TS/1/A/600/4.
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2.5 B&W Drawing #163356, Reactor Coolant Piping Arrangement Plan. 2.6 Dwg. No. P.O. 187 (1) -(2) .
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3 2.7 Eng. Letter - File OS-18, OS-27, - Dated 11-1-71. 3.0 Time Required Portion of a shift prior to Hot Functional Test, during Hot Functional 3
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Test and following cooldown from Hot Functional Test.
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4.0 Prcrcquisite Tests 9 None
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5.0 Test. Equipment *
. l 5" outside precision calipers.
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6" Scale '. . Inside Micrometer 3" - 4" range.
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Outside knerometer 3" - 4"* range. , Hydraulic Jack with 3" - 4" total length, including travel. ;-
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6.0 Limitations and Precautions 6.1 No adjustments to be made to the load supports except as authorized _
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by the Station Superintendent. ,
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6.2' 1101/01 - Plant Operating Limits and Precautions. : 6.3 Stop RCS heatup if the hanger deflection indicating pin is approaching
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or has contacted its travel hard stop.
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6.4 Stop RCS heatup if piping or components come in contact with seismic
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restraints, beams, etc. ' '
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7.0 Required Plant Status
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7.1 RCS in cold condition. :
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8.0 Prerequisite System Conditions , ,
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8.1 Alltravelstopsonhangersmu'stberemoved-friortosystemsheat-up.
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9.0 Test Method ' 9.1 Cold and hot pipe hanger position buttons will t:. observed before - - . . . . _.
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The WHITE button is Cold ~ position, and the
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and during system heatup. RED button is hot position. Position indication readings will be
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taken at 50*F increments during heatup and cooldown. Pipe and e component deflections will be read and recorded at 50*F increments during heatup and cooldown. Seismic restraints will be checked at
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200*F, 350*F and 500*F during heatup and cooldown to insure piping. and compor.ents do not come hard against a restraint. 1 10.0 Data Required
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! 10.1 Initial Cold Positions Vs. RCS Temperature.
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600/14-5
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10.4 Final Cold Positions Vs. RCS Temperature following Hot Functional
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11.0 Acceptance Criteria .- Position indication readings during and after test stay within the travel range and do not contact the hanger stops. Piping and components do not come in contact with seismic restraints. Piping and component deflections
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' 12.0 Procedure 12.1 Record all hanger positions on enclosed data sheets prior to Hot Functional Test.
+12.2 Record all hanger positions,every,50* change of RC temperature
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during heatup for Hot Functional Test. 12.3 Item #6 on PO Dwg.187 (1) (2) to be taken as following:
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1.a. Cold Position o
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- 2. The reading A, B, & C to be taken with a 6" scale at the - l designated point, with the block away from the OTSC & firm against .e support plate.
- 3. The "D" reading to be taken at 500*F with the inside Micrometer
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at the designated' points marked on the blocks, then by the use of the jack the block is firmly pressed against the stationary block on the OTSC. 12.4 The RCP LOCA restraints, piping supports & OTSG supports to be checked as following:
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- a. Cold Position
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.,c. 350*F -
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' The data sheets to be checked off as 0.K. or against HARD STOP. -
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NOTE - If against HARD STOP, notify Shif t Supervisor to stop RC heatup and notify Test Coordinator.
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12.5 Upon attaining 532*F adjust load adjustment bolt if hanger is not at hot position until indication is on hot position button. 12.6 Record all hanger positions every 50*F change"of RC temperature
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during cooldown after Hot Functional Test. . f-13.0 Enclosures 13.1 5 ketch of Grinnell Constant Support, Figure 80V-Size 52. 13.2 Data Sheet - RC Piping , 13.3 Support Schematics. 13.4 Data Sheet .- Pump & Miitor Supports
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1325 Data Sheet - RC Piping Restraints ,
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13.6 Data Sheet - OTSG Lateral Supports 13.7 Data Sheet - Hangers
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APPENDIX C OCONEE NUCLEAR STATION UNIT 1 REACTOR COOLANT SYSTEM COMPARISON OF DESIGN VS ACTUAL MOVEMENTS I 1 i l l C-I l l
_ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . P Considerations Utilized in Comparison Study
- 1. Reactor Coolant System Coefficient of linear thermal expansion for carbon steel pipe between ambient temperatures 100*F and 500*F = 0.034 inches /
foot.II)
- 2. B & W maximum Reactor Coolant System movements are shown on Drawings 128625E and 128626E, Figures 1 & 2. These movements were converted to approximate movements at 500*F by linear extrapolation.
3 For measurements made perpendicular to piping and vessels, the radial expansion of same is added or subtracted, as the case may be, to the thermal movement for converting to centerline movements.
- 4. Feedwater System coefficient of linear thermal expansion for carbon steel pipe between ambient temperatures 100*F and 350*F = 0.020 inches /footf 2) 5 The directions (North), (South), (East), (West) and (Up) located beside the X, Z, Y coordinates in the following tables indicate the direction of movement of the particular point.
References:
(I} ANSI B31 7 (1969), Table A.5., Coefficients of Thermal Expansion. (2) ANSI B31.1.0 (1967), Appendix B, Thermal Expansion Data. C-il
!
l
..
.
Monitoring Point Location # 1 - Pump 1A1 R. C. System Actual Measurement (Inches) ' Press. (psi) Temp. (*F) X(North) M Z(East) 0 100 2.653 0.597 1.878 500 200 2.545 0 964 1.673 1032 350 2.475 1 378 1.283 2150 500 2.376 1.872 0 752 i Design Movement @ E Pump 100*F to 500*F 0.394 1.060 0.880 l. Actual Movement @ E Pump 100*F to 500*F 0.277 1.275 0 990(I} Difference Between Design & Actual 0.117 0.215 0.110 Movement
% Difference 34.8% 18.4% 11.8%
1 (I)l.126 Difference - 0.136 radial expansion = 0 990 C-1
_, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . Monitoring Point Location # 1 - Pump 1A2 R. C. Sys tem Acutal Measurement (Inches) Press. (psi) Temp. (*F) X (South) Y(Up) Z(East) 0 100 1.975 0.202 2.027 500 200 2.062 0.320 1,822 1032 350 2.325 0.790 1.572 2150 500 2.741 1.277 1.262 Design Movement @ [ of pump 100'F to 500*F 0.394 1.060 0.880 Actual Movement @ E of Pump 100*F to 500*F 0.630(y), 1.075 0.765 Difference Between Design & Actual Movement 0.236 0.015 0.115
% Difference 46.1%( l.4% 14.0%
-
-
.
II)0.766 difference - 0.136 radial expansion - 0.630 (2)During Hot Functional Test, the target stud was accidently bent during installation of the mirror insulation thereby voiding the initial cold reading. The target stud was realigned for remaining readings and the cold reading was taken again af ter the unit was cooled bac.k down to 130*F.
- Value exceeds limiting criteria listed on P0-187 Rev.1 Figure 3 C-2
.
._ _____- .
Monitoring Point location # 2 - OTSG Elevation 849'-8 1/2" R. C. System Actual Measurements (Inches) Press. (psi) Temp (*F) X (South) Y(Up) 0 100 1.234 0.982 500 200 1.181 1.590 1032 350 1.065 2.408 2150 500 0.953 3.375 Design Movement @ E of Pipe 0.094 2.212 100*F to 500*F - Actual Movement @ E of Pipe 100*F to 500'F 0.074(2) 2.393 Difference Between Design & Actual 0.020 0.181 Movement
% Difference 23.8% 7.9% __
.
.
Original readings taken from target studs were erroneous as confirmed by readings taken between tube sheet and seismic stop. The reidings recorded above were taken between the tube sheet and seismic stop and on the South side. (2) Difference 0.281 minus 0.207 radial expansion = 0.074. c-3
.-
__ _ -. __ _ _ _ _ _ _ _ _ _ _ _ _ - . Monitoring Point Location # 3 - OTSG - F.W. Header Norrie R. C. System F. W. System Temp. (*F) Actual Measurements (Inches) Press. (psi) Temp. (*F) X (North) Y(Up) 0 100 100 2.033 0.338 500 200 2.013 0 553 1032 350 T.985 0 997 2150 500 350 1.920 1.517 Design Movement e E of F.W. Pipe 0.142 1.154 100*F to 350*F Actual Movement at E of F.W. Pipe 0.113 1.179 100*F to 350*F Difference Between Design & Actual 0.029 0.025 Movement
.
% Difference 22.7% 2.1%
.
1 c-4 ) l l _.
_ _ _ _ _ _ _ _
. .
Monitoring Point Location # 4 - OTSG Upper Tube Sheet R. C. System Actual Measurements (Inches) Press. (psi) Temp. (*F) Y (Up)
'
0 100 0.682 - 500 200 1 315 1032 350 2.110 2150 500 2 960 4 Design Movement 100*F to 500*F 2.236 ! Actual Mosement 100*F to 500*F 2.278 Difference Between Design & Actual 0.355 Movement
._
% Difference 1.9%
i
.
C-5 _ ~
.. . . .
Monitoring Point Location #19 - Pressurizer Surge Line Nozzle R. C. System Actual Measurement (Inches) Press. (ps i) Temp. (*F) y b } b p) Z b)(East) i 0 100 1.839 1 776 500 200 1.491 1 994 1032 350 1.084 2.242 2150 500 0.630 2.535 I
.
Design Movement 100*F to 500*F 1.204 0 704 Actual Movement 100*F to 500*F 1.194 (2) 0.774 I3)
'
Difference Between Design & Actual 0.010 0.070 Movement
% Difference 0.8% 9.5%
l l
-l I
,
)
i U)Y and Z readings reported above are for the top and west monitoring points. Bottom and east readings were discarded as they were less meaningful . (2)Dif ference 1.209 minus .- 0.015 radial expansion = 1.194 (3I Di f ference 0.759 plus 0.015 radial expansion = 0.774
-
)
i
!
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.
.
.
Monitoring Point Location #20 - 180*E11 Tangent Point at Top of Straight Riser
. C. System Actual Measurements (inches)
Press. (pal) Temp. (* F) X Horth) Y6Jp) Z 66ast) 0 100 2 559 1.024 2.841 500 200 2.548 1 720 2.770 1032 350 2.515 2.609 2 700 2150 500 2.485 3.515 2.646 Design Movement 100*F to 500*F 0.035 2.250 0.265 Actual Movement 100*F to 500*F 0.023(I) 2.491 0.246 (2) Difference Between Design & Actual 0.012 0.241 0.019 Movement
.
% Difference 41.4% 10.2% 7.4%
O) 074 difference - 0.051 radial expansion - 0.023
.
> ( ) 195 difference + 0.051 radial expansion = 0.246
.
-
C-7 i '
, - .
. . -
Monitoring Point Location #21 - l Al Pump Discharge R. C. System Actual Measurements (inches) Press. (psi) Temp. (*F) X (North) Y (up) Z (East) 0 100 0.836 0.834 1.146 500 200 1.102 I.150 1.230 1032 350 1.380 1.500 1.315 2150 500 I.734 1.910 1.432 Design Movement 100*F to 500*F 0.578 1.094 0.459 Actual Movement 100*F to 500*F 0.847 ) 1.076 0.286 Dif ference Between Design r, Actual 0.269 0.010 0.173 Movement
% Difference 37.8% 1.7% 46.4%
t j
.
.
(IIDi f ference 0.898 minus 0.051 radial expansion = 0.847 C-8 _ -
. .- -
. . .
Monitoring Point Location #21 - I A2 Pump Discharge
.
'
R. C. System Actual Measurements (Inches) Press. (ps i) Temp. (*F) X (South) Y OJp) Z (East) 0 100 0 775 0.664 1 553 500 200 0.902 1.112 1.653 1032 350 1.187 1.465 1 787 2150 500 1.494 1.871 1 982 Design Movement 100*F to 500*F 0.578 1.094 0.459 Actual Movement 100*F to 500*F 0.668(I) 1.207 0.429 Di f ference Between Des ign and Actua l 0.090 0.113 0.030 Movement
% Dif ference 14.4% 9.8% 6.8%
.
( }0if ference 0.719 minus 0.051 radial expansion = 0.668 C9 i
- to APPENDIX D OCONEE NUCLEAR STATION MDS-JR-52
.
I
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i i L _ _ . _ .
- - _ - - __ _
_ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- *,,,
**, -
OUKE POW 6R LOHi'ANY
- , ,
DESIGN ENGINEERING - MECHANICAL SECTION ENGINEERING J'JSTIFICATION REPORT
,
Date: February 1, 1973 Report No, MOS-J R- 52 To: E. D. Powell/J. E. Smith Originated By: P. N. Hall Station: Oconee 1 Approved By: [ L System: Reactor Coolant Sys 50 Sheet I of 1 Mech Sec Group File No: 05-27.4 Variation Reported by a rd Date : Mr. J. E. Smi th, ONS, 1-31-73 Description of Variation: The approximate established. acceptance criteria for thermal movement of Reactor Coolant Pump 1A2 were exceeded during the
~ ~
~
thermal monitoring program. Certify that the exce'ption is accepta'b'le.' Engineering Analysis Required: Analyze the actual movement with respect to establishe'd criteria and certify results to Duke Steam Production for purpose of clearing exception noted in monitoring data. Engineering
Conclusions:
Actual movement was reviewed with respect to its effect on the thermal stresses in both the RC Pump inlet piping and discharge piping. In both cases the resulting stresses are expected to be less than predicted by a factor up to 15%; therefore, the exception to the established criteria is totally acceptable. Engineering Disposition of Variation (List all drawings, specif kations and/or calculations revised by number, name and revision number): This item will be discussed in more detail in Design Engineering's report on RC System thermal moni toring' program. Engineering Regommendations: None required. S.K. Blackley, Jr., Chief Engineer Mechanical & Nuclear Division
.
By: PNH:rjd cc: D. G. Beam J. S. Davis R. E. Miller J. L. L. Ostertag
.
__ _ _ _ _ _ _ _ _ _ _ _ _ _ a ff '/ pf
' REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDSS DISTRIBUT]ON FOR INCOMING MATERIAL 2707297 REC: OREILLY J P ORG: PARKER W O DOCDATE: 08/28/79 NRC DUKE PWR DATE RCVD: 09/11/'i DOC TYPE: LETTER NOTARIZED: NO COPIES RECEIVED SUDJECT: LTR 1 ENCL 1 FORWARDING SUDJECT FACILITY"S SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE AND ENVIRON MONITORING REFT COVERING Tile PERIOD JANUARY 1, 1978 THROUGH JUNE 20, 1970.
(/d b(CN ' PLANT NAME: OCONEE - UNIT 1 REVIEWER INITIAL: XJM OCONEE - UNIT 2 DIGTRIBUTER INITIAL $ M OCONCE - UNIT 3
- &* DISTRIBUTION OF THIS MATERI AL IS AS FOLLOWS *+-****************
NOTES:
- 1. M. CUNNINO! LAM -- ALL AMENDMENTS TO FSAR AND CHANGES TO TECH SPECS ANNUAL ENVIRONMENTAL RPTS (OL STAGE).
(DISTRIDUTION CODE A007) FOR ACTION: DR Cl " ORD#4 DC**W/7 ENCL INTERNAL: sce ._u*W/ ENCL NRC PDR**W/ ENCL
& En*W/2 ENCL EED**W/ ENCL ENVIRO SPEC DR**W/ ENCL EFFLUENT TREAT SYS**W/ ENCL RAD ASSESSMENT BR*nW/ ENCL KASTNER**W/ ENCL EXTERNAL: LPDR'S WALHALLA, SC**W/ ENCL TERAn*W/ ENCL NSIC+*W/ ENCL ACRS CAT Bn*W/ ENCL r
.- (
y1 Dr *
.
DISTRIBUTION: LTR 20 ENCL 20 CONTROL NBR: 782480287 SIZE: 1P+1GP
'
c*0c****************************** THE END *************************A- ***- 012 800 Cp
*
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DirKn Powen COMPANY l' owr.w Iltist.n No 422 Sot: Tis Cutincu Srims:T, CliAug.oTTE, N. C. una.u w t LLI AM O. PA R K E R, J R. Vocr Patsf ormt Trit Pa o N t:Aata 704 5 +taan Paoouction 373-4063 August 28, 1978
#
Mr. James P. O'Reilly, Director L
-
" * - * *"= ' I 8 Region II Office of Inspection and Enforcement '
< U. S. Nuclear Regulatory Commission Suite 1217 ~
'
230 Peachtree Street, Northwest - Atlanta, Georgia 30303
'
Reference:
Oconee Nuclear Station :; Docket Nos. 50-269, -270, -287
Dear Mr. O'Reilly:
Pursuant to 10CFR50, 550.36 and Oconee Nuclear Station Technical Specifi-cation 6.6.1.2(c), please find attached data concerning radioactive effluents released from Oconee Nuclear Station. This report covers the period January 1 through June 30, 1978. V y truly yours,
~
%
j Nu :. d. k' William O. Parker, Jr RLG:vr Attachment cc: Director, Office of Inspection and Enforcement
- l i
782480287 s S
\l\
.#.
ge ATTACHMENT 1 RADIOACTIVE EFFLUENT RELEASES SOLID WASTE
.
,,
-
Pt 1 of 4 RADIDACTIVE EFFLUENT RELEASES SOLID WASTE Total volume of solld waste packaged (cubic feet) 26,073.8 Total estimated activity involved (Curies) 752. Disposal of materials shioped off-si te: All shipments to Chem-Nuclear Systems Waste Disposal Facility at Barnwell, South Carolina. DATE CUBIC FT. # OF CURIES 1/3/78 185 2.38 1/6/78 185 2.61 1/7/78 238 1.89 1/9/78 238 1.14E+1 1/10/78 130 1.45 1/12/78 377.4 4.16E-1 1/13/78 205 2.10 1/17/78 205 1.25E+1 1/18/78 235 1.50E+1 1/19/78 205 4.07 1/21/78 238 4.78 1/22/78 238 8.93 1/23/78 766.6 6.88E-1 1/24/78 205 1.86E+1 1/25/78 391.4 2.23
- 2. 8 C,E+1 1/27/78 185 1/30/78 357.4 2.82E-1 2/1/78 205 9 23 2/2/78 283 5.08 2/6/78 130 5.10E-4 2/7/78 185 3.74
'
2/9/78 166 , 1.40E-2 2/9/78 80 1.20E-4 2/9/78 80 1.29E-4 2/13/78 185 6.04
- Ptg2 2 of 4
.
,
-
DATE CUBIC FT- # OF CURIES
'
2/15/78 756.2 2.10E-1 2/15/78 60 8.24E-1 2/15/78 - 60 5.55E-5 2/16/78 185 3 02
-
2/17/73 185 3.49 2/20/78 657.8 3 32E-1 2/22/78 205 6.21 2/27/78 60 1.94E-2 2/27/78 60 6.78E-2 2/27/78 60 5 39E-2 2/27/78 60 2.25E-2 2/27/78 60 6.20E-2 3/1/78 185 7.29 3/3/78 185 8.44E-2 3/5/78 238 1.71E+1 3/8/78 589 4.17E-1 3/10/78 205 6.79 3/14/78 7.4 5.10 3/14/78 205 1.26E+1 3/15/78 637.8 3 38E-1 3/17/78 205 7.39 3/21/78 205 9.43 3/22/78 74 4.08E+1 3/27/78 205 7.82 3/30/78 7.4 ' 1.19 3/31/78 205 7.91 1 3/31/78 238 8.23 ) 4/4/78 167 8.10E-1 4/5/78 238 4.23E+1 4/5/78 468.4 5.96E-1 l
-4/7/78 238 7 29 j 4/7/78 205 1.04E+1 4/11/78 166 3.00E-2
~
4/11/78 166 7.00E-2 4/12/78 205 1.17E+1 4/14/78 14.8 1.02 4/14/78 205 9.89 4/18/78 205 7 73
. - . . .
, _ Praa 3 of 4
*
- DATE- CUBIC FT. # OF CURIES
- 4/18/78 14.8 5.10E-1 4/20/78 14.8 2.55 4/20/78 238 4.76
' 4/21/78 170 4.25 4/21/78 14.8 1.70 4/22/78 512 8.00E-2 4/24/78 592 2 90E-1 4/25/78 205 .' 38
-
4/25/78 14.8 5 95 4 4/28/78 14.8 6.63 5/1/78 205 3 99 5/3/78 7.4 3.40E+1 5/4/78 537 2 1.03 5/5/78 238 1.09E+1 5/5/78 7.4 1.19 5/6/78 205 9.03 5/7/78 7.4 6.80E-1 5/7/78 238 2.24 5/12/78 238 1.73E+1 5/14/78 130 2.06 5/16/78 130 1.07E+1 5/18/78 238, 1.14 5/18/78 205 3.16 5/19/78 868 2.19 5/19/78 45 8.73E+1 5/22/78 238 3 06
'
5/23/78 238 1.04 l 5/26/78 238 4.00 5/26/78 576.4 1.78E-1 5/26/78 203 3.19 5/28/78 238 2.43 5/31/78 205 2.66 6/3/78 205 . 3.85 6/4/78 205 . 2.22 l 6/5/78 238 6.51 6/6/78 205 1.96 l l _
' .
,. .
- -
F 3 4 of 4
.
- DATE' CUBIC FT. # OF CURIES 6/8/78 205 3.32 6/9/78 81.4 3 04 6/9/78 205 2.50
; 6/9/78 238 2.05 6/12/78 205 7.37 6/12/78 238 1.16 6/14/78- 205 1.06E+1 6/15/78 130 <
1.03E+1 6/16/78 205 1.43E+1 6/16/78 621.6 7.03E-1 6/17/78 130 1.53E+1 6/20/78 205 1.40E+1 6/26/78 130 4.10E-1 6/28/78 205 3.63 6/28/78 205 1.74E+1 6/30/78 238 7.84 6/30/78 562.4 1.84 6/30/78 205 8.09 . TOTAL 26073.8 7.52E+2
,
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O e ATTACHMENT 2 RAD 10ACTI EFFLUENT RELEASES LIQUID RELEASES l
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- i. J l/,L 72 L.?, is CT! i' ..f=i-01 1. w 3- 91 2.543 'I 1. N E+ D
- 5. U 19.H ;;:0-*/ M i! !* SEL2a*E- UO!/ 'L 2.17 3-3 e 7.I42-3J 2.5- M , 2. 75-1v
;. m.4 : ..; 2 00i'.i1 CI 2 i "iL4/.*I~ UCI/ :L 1.775-07 1.52E-17 a. h 1.hE-J7 , ..
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- a. J:lAL f:L:43J O'J 7 E' :.l.E+'t s.odi+31 1.43E+32 3.2 - E+ 32
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- -l'8 7.5t:-02 i.G45-32 1.225-01 2. $ 5-11
;-i i3 3.835-03 3.145-13 5.975 '3 1.IdE-32
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. l ATTACHMENT 3 l RADICACTIVE EFFLUENT RELEASES AIRBORNE RELEASES
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- 2. TwfAL aALwJEns OsnIES s.935-02 0.93E-03 1.15E-02 7. ale-02
- 3. TwiAL /AnlICJLAiE Jksis JETA-wA:uA CanIE3 6.405-04 1.05E-0J 6. 99 E-03 5.335-03 4 TOTAL inITIL4 CJAIES 3.05E+0i 4.52E-04 9.1J5 01 l .2JE+02
- s. T0iAL /AnTICJLAi!
sad 55 ALenA A07IVI!( CJnIES 0. C. O. O.
- a. 4AXIm h aw?LE JAS itELEAsi nATE UCI/5E' l.60E+03 l . 60 E+03 1.60!+03 1.00E+03
- 7. rExCE i OF APPLICA3L2 .
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- 4. 154Torda TiELIA3E2 OGIE3 fad!IOLLAiE5 0a-131 2. 33E-05 2.c25-05 7.23E-05 1 225-04 5A-LA-4-0 1.095-0s 1.29E-05 2.625-05 5.00E-05 5n--O 2. l o E-07 2.25E-07 2.095-07 6.50E-0 7 04-434 6.7dE-06 3. 77E-Ce 3.0 6E-05 4.J2E-05 Sn-o, 2.92E-06 3.04E-06 2.82E-06 8.78E-06 0.-oo 7.84E-07 7.15E-05 4.725-05 1.19E-04 Ca-lJs 3.95E-06 I.98E-07 2.385-06 a.34E-06 Cs-43a 9.17E-Co 2.ooE-06 5.705-0s 6.895-05 An-o' l . I I E-0 7 8 .35 E-0 7 I . 70E-0 7 4.67E-07
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