ML20086C373
| ML20086C373 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 06/30/1970 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| References | |
| 02347, 2347, NUDOCS 8311230220 | |
| Download: ML20086C373 (97) | |
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- 11. B. Robinson tinit No. 2 Steam Pipe Break June, 1970
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TABLE OF CONTENTS Page Main Report: Steam Pipe Break. . . . . . . . . . .. . . . . . . .. . 1 Addendum No. 1: Failure Analysis of Piping to Safety Relief Valve . . . .. . . . . .. .. . . . .. . .. 1-1 Addendum No. 2: Design Analysis Report . . . . . . .. .. . . . . .. 2-1 Appendix A . . . . . . . .. . . . . . .. . . . . .. 2-8 Exhibit I. . . . . . . . .. . . . . . .. . . .. . . 2-10 Exhibit II . . . . . . . . . . . . . .. . . . . . . . 2-16 Exhibit III. . . . . . . . .. .. . .. . . . .. . . 2-17 Addendum No. 3: Transient Aspects of a Steam Pipe Rupture at the 11. B. Robinson Unit No. 2 . .. .. . . . . . . 3-1 ii
LIST OF TABLES Page Table 1-1. Chemical Analysis Results 1-11 Table 1-2. Summary of Spectrographic Analyses of Pipe and Weld 1-12 Table 1-3. Wall Thickness Variation at Pipe Failure Location 1-13 Table 1-4. Comparison of Mechanical Properties A-106 Grade B llot Rolled Pipe 1-14 Table 3-1. Raw Reactor Vessel Cooldown Data 3-11 Tabic 3-2. Computer Temperature Record 3-14 h L - - - - - - -
LIST OF FIGURES Page Figure 1-1. Safety Valve Connection 1-15 Figure 1-2. Adjacent Pipe and Insulation Cut by Steam and Water Jet 1-16 Figure 1-3. Pipe Struck by Safety Valve 1-17 Figure 1-4. I.D. Extrados of 3SV4 Exhaust 1-18 Figure 1-5. Piping System 1-19 Figure 1-6. Weld Area of Safety Valve 2SV-1 1-20 Figure 1-7. End View of Fracture on Valve End 1-21 , Figure 1-8. Side View of Fracture and Valve Body 1-22 Figure 1-9. End View of Fracture on Pipe End 1-23 Figure 1-10. Portions Cut from Pipe and Valve, After Sectioning 1-24 Figure 1-11. Corresponding Fracture Surfaces and Cross-Section From Pipe and Valve Ends 1-25 Figure 1-12. Light Micrographs of Etched Cross-Section of Secticn A Near Center 1-26 Figure 1-13. SEMs of Fracture Surface on Pipe End Section 1 1-27 Figure 1-14. Macro-Etched Cross-Section Through Deformed Valve End in Necked-in Region 1-28 Figure 1-15. Fracture Surface and Cross-Section Near Region of , Final Rupture 1-29 Figure 1-16. SEMs of Fracture Surface on Pipe End , Section 4 1-30 Figure 1-17. Light Micrographs of Cross-Section Through Section 7 1-31 Figure SEMs of Fracture Surface on Pipe End. , Section 7 ' ' 1-18. 1-32 Figure 1-19. Corresponding Sections From Pipe and Valve Ends 1-33 Figure 1-20a. liardness Traverse Valve End Section 2 1-34 Figure 1-20b. liardness Traverse Valve End Section A 1-34 - Figure 1-21. Sections Cut From Pipe End portion Before and After Tensile Testing 1-35 , iv a 2 - - - __-
List of Figures (continued) Page Figure 2-1. Sketch of Nozzle Modification for Main Steam Relief Valves 2-6 Figure 2-2. Sketch of Nozzle Modification for Steam Dump Valves 2-7 Figure 3-1. Steam Line Break Equivalent to 430 lb/sec at 1000 PSIA 3-20 Figure 3-2. Steam Break Comparision With No Stored lleat, Coolant Pump Coastdown, and 3 Charging Pumps 3-21 Figure 3-3. Pressurizer Water Level 3-22 Figure 3-4. Steam Break Comparision with 3 Charging Pumps and Coolant Pump Coastdown 3-23 Figure 3-5. Steam Break Comparision with 1 Charging Pump and No > Coolant Pump Coastdown 3-24 .
- Figure 3-6. Keff vs Core Taverage at 2250 PSIA with 1 Stuck Rod 3-25 Figure 3-7. TR-413 RCS Ilot Leg Temperature 3-26 Figure 3-8. TR-410 RCS Cold Leg Temperature 3-27 Figure 3-9. PR-444 Pressurizer Pressure 3-28 Figure 3-10. LR-459 Pressurizer Level 3-29 Figure 3-11. LR-477 Steam Generator Level Cold Calibration 3-30
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r TNCIDENT REPORT H. B. Robinson Unit No. 2 Steam Pipe Break on Tuesday, April 28, 1970 in accordance with the final stages of hot functional testing of the H. B. Robinson Unit #2 plant, secondary system safety valves were being tested. At approximately 3:19 p.m. the section of stcam pipe connecting the "C" main steam line to safety valve SVl-4-C f ailed. The failure was a 360 circumferential break which allowed the safety valve to be blown off. The escaping steam caused injury and burns to seven men in the immediate area. As the rupture was a non-isolable break, en uncontrolled cooldown of the plant occurred. No nuclear fuel was in the reactor at the time of the incident. Plant evolutions prior to the incident were: The final stages of the hot functional test program were in progress and specific. ally the steam generator secondary system safety valve lift set pressure was being checked. Verification of safety valve set points and adjustments as
- necessary were being performed using a pneumatic test device which allows the set points to be verified without having the system pressure up to lift pressure. The plan't was operating at 533*F and 2225 psi primary system pressure with a secondary system pressure of 900 psi and a constant level being maintained in the steam generator.
Eight of a total of twelve valves had been tested completing work on main steam line A and B. Work had just started on testi,ng the first of four valves on main steam line C, The test supervisor had notified the control operator located in the control room that tests were completed on steam line B and tests were being initiated on steam line C. It has been confirmed by reviewing data sheets recovered from the scene and by interviewing the men performing the test, that the pneumatic test device had been installed on the first safety valve of steam line C and that air pressure was being increased to relax the valve spring force, to make a determination of the valve set pressure. The test supervisor who was stationed by the valve being tested had a telephone ard earphones to provide a direct telephone contact with the plant control room.
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Incident Report At approximately 3:19 p.m. a loud noise was heard followed by a shower of steam, insulation, scaffolding, metal parts and construction debris in the vicinity of the safety valve on steam line C. The men in the vicinity of the valve were either knocked to the deck by the explosion or were forced to lie down due to lack of air to breathe. The rapid release of steam displaced the air from the area above the pipe requiring the men to stay in a position near the floor. They made their way out of the area and down the steps away from the immediate scene of the incident on their own power. These men were transported to a local lbrtsville, South Carolina, hospital by ambulance and treated for burns and injuries. One man received treatment and was released. The other six were admitted to the Byerly Hospital for treatment of their injuries. Several people including craft workmen, construction supervisory personnel and operations and test personnel were in close proximity to the incident and observed the incident from several vantage points. A summary of the interviews of several people who witnessed the incident reveals the following information: The initial noise or sound was immediately followed by a secord louder noise or sound. The initial steam accumulation in the area of the break spread in an almost horizontal plane, followed by the formation of a vertical column of steam which rose an estimated 150 feet into the air. Upon inspection following the incident, it was observed that there was an area of localized cutting of insulation on a nearby line on a horizontal plane from the break. This damage suggests an initial crack in the south southwest quandrant of the pipe. Steam apparently was directed horizontally in this direction for a brief period of time prior to a complete severence of the pipe and expulsion of the total valve assembly f rom the area by the force of the steam jet. Such a sequence of events is also suggested by the reported two stages of sound, the appearance of the fracture, and the direct' ion of travel (f the separated valve. The valve was propelled in a northeasterly direc-tion against the structural members supporting the steam lines, rebounded back toward the turbine building. The valve came to rest on the turbine building mezzanine floor approximately 12 feet to the east, 6 feet to the south and 18 feet below its original position. In addition to striking the supporting structure, the valve struck and carried away an angle brace and dented and moved the stack from the auxiliary boilers, causing its supports to bend and break away. ) 2
) Incident Report Although the valve spring was being tested, there is no conclusive evidence to show that this valve actually reached its set point and
" popped". The men testing the valve do not believe that it did pop.
Evidence to support this opinion included fresh grease on the disc which showed no evidence of having been exposed to hot steam. It has been stated that at least one other similar valve in steam line "A" had experienced a definite pop during testing the morning of the day of the incident. Other valves being tested had experienced varying degrees of valve seat lift from a " simmer" or weeping to a definite lif t or pop. The set point of this valve which had been set at the vendor's shop at 1140 psi was being checked and had reached about 40% of the test actuator pressure (50 psi out of 120 psi) which is required to overcome the spring force, and achieve " popping". The valve " pops" open because ~ the exposed area of the seating pad is greater than that of the valve port and, therefore, a larger force is applied to the biasing spring. Both the main steam stop valve and bypass valves were closed for one day prior to the incident. Steam line A was blown down for a brief period of time (estimated to be 2-3 minutes) the morning of the incident. Steam line B was thoroughly blown down as well as being used to supply steam to the steam driven auxiliary feed pump for testing the day of the incident. Steam line C was blown down briefly (cstimated to be 1-2 minutes) the day of the incident. From the piping configuration it is possible that condensate can collect in the line upstream of the stop valve. The amount of this water in the lines at the time of the incident is not known. The failure was a 360' circumferential break in an almost horizontal plane. It occurred in the pipe adjacent to the weld between the valve body and the branch from the main steam line. This branch line was approximately six inches long and it was a six inch diameter schedule 80 pipe consisting of A106 Grade B carbon steel machined to a schedule 40 weld preparation for the weld to the valve body. The valve body had a similar weld preparation. The break was located adjacent to and just below the field weld. A sketch of the configuration of the pipe and plant layout is attached. This weld and pipe had been subjected to a " cold hydro" test at 1356 psi and subsequently been operated at elevated temperature and pressure for approximately nine days prior to this incident. Temperatures varied from 520 - 540*F and pressure from 800 - 1000 psi during this nine day period. A detailed metallurgical examination of the broken pipe was performed by the Westinghouse Research Laboratories at Churchill. Westinghouse, Ebasco Services, B. F. Shaw Company, Carolina Power & Light Company and Southwest Research Institute as consultants to Carolina Power & Light Company, participated in this program. A report is included.as Addendum No. 1 to this report.
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Incident Report A subsequent review of the quality assurance records of the subject weld shows that it had been made in compliance with the design specifications. A detailed review of the design of the pipe and valve installation is included as Addendum No. 2 to this report. The personnel on duty in the control room prior to the incident included a control operator and a shift foreman. These Carolina Power & Light Company personnel were operating the plant in accordance with the hot functional test program with the primary system at 533*F and 2225 psi on automatic pressure control. Three reactor coolant pumps were in service and one charging pump was operating in automatic control with letdown through the 45 gpm letdown orifice and feeding A train of the mixed bed demineralizer. Fbkeup to the system was being controlled automatically to the volume control tank via thc primary water pump taking suction from the monitor tank. The monitor tank was serving as a temporary water source as the primary water tank was out of service for modification. The secondary system pressure was approximately 900 psi with main steam isolation and main steam bypass valves closed. All three steam generator blowdown lines were valved to the blowdown tank and throttled to achieve system temperature and pressure control. All three steam generator levels were being maintained at approximately 70 percent level. The motor driven auxiliary feed pumps were lined up to take suction from the condensate tank and feed the steam generators as required to maintain proper level. The main feedwater and feedwater bypass valves were closed and the main feedwater pumps were secured. The steam driven auxiliary , feed pump was also secured and its discharge valves closed. No feedwater was being added to any steam generator at the time of the incident. At the time of the incident a loud noise was heard by the control operator followed by a rapid decrease in pressurizer level and pressure. In addition, level decreased rapidly in the "C" steam generator and reactor coolant system temperature began to rapidly decrease. Action was taken by the control operator to secure all three reactor coolant pumps. Two additional charging pumps were placed in service and letdown was secured to minimize the ef fects of pressurizer level and pressure decrease. . Pressurizer heaters were manually de-energized prior to reaching the automatic heater cutof f set point on pressurizer low level. Even though the pressurizer level decreased off scale, it is not believed that pressurizer steam bubble expanded out of the surge line. Within two to three minutes all available operators had reported to the control room in person as the extremely loud noise produced by the steem blowdown rendered communications by the plant intercom system undesirable. Two boric acid transfer pumps were started to provide makeup to the charging pumps in addition to the make up from the monitor tanks which was already in service. ) 4
Incident Report The level in steam generator "C" decreased to zero as indicated on the wide range level recorder in approximately one hour. The indicated wide range level reached zero at approximately the same time that there was a marked reduction in noise level coming from the steam blowdown. This indicates a good correlation between the measured steam generator 1cvel as indicated by the instrumentation and the actual condition of the level. The overall transient caused the reactor coolant system to cooldown approximately 213*F over a one hour period. During the course of the plant cooldown, detailed in-plant inspections were conducted to check for any other possible problems or piping system interferences. None cere found. The details and discussion of the plant transient along with the engineering evaluation of the effects of the transient on the system and components as well as the steam break protection ef f ectiveness is included as Addendum No. 3 to this report. A debriefing of the operating personnel involved in the incident was held on Thursday, April 30, 1970. This substantiated the material in the body of this report. In addition, some comments are worthy of note in regard to the qualifications of the operators to handle major plant emergencies. The operator remained fully cognizant of the plant conditions throughout the incident by monitoring the control board instrumentation. _In addition, the control operator felt that the information made available to him during his training was adequate to prepare him to handle this incident. There was an emergency procedure available in the control room, which the operator was familiar with; however, no specific reference was made to the written procedure during the course of the recovery. Frequent reference was made to detailed system flow diagrams which were available in the control room to verify the logic of the recovery operation. Recovery of the plant was accomplished in a carefully planned orderly manner. Pressurizer level was restored to normal, no-load, operating level which resulted in regaining a coolant pressure of approximately 2050 psi. Normal charging and letdown was established. In preparation for starting the first "C" reactor coolant pump, "A" and "B" steam generator temperature and pressure was reduced and reactor coolant pressure was reduced to minimize the thermal and pressure transient. The steam generator temperature and pressure was) reduced by intermittent use of bottom blowdown and by feedwater addition, and by steaming through the main steam isolation bypass valves. Reactor coolant pressure was reduced to approximately 1250 psi by use of the auxiliary spray system. ) 5
Incident Report In addition, a detailed survey of the reactor coolant and secondary systems was conducted to assure that there were no damaged equipment, pipe or equipment interferences or other abnormalities. A review of all available plant temperature instrument readings was made which included measuring selected surface temperature readings of the steam generator and coolant loops to be satisfied that there were no large temperature dif ferences between components and primary / secondary interface. Plant conditions prior to starting the "C" reactor coolant pump were: pressure 1250 psi, reactor vessel temperature 337*F, coolant loop temperatures (including bypass loop) were between 295'F and 322"F. Pressurizer level was normal. Reactor coolant pump "C" was started with no abnormalities noted. The resulting pressure transient was less than 50 psi and the maximum temperature transient noted was approximately 50*F. After plant temperatures were stabilized, the remaining reactor coolant pumps were started and the plant was cooled down utilizing normal operating procedures. h 6
Addendum N-. 1 J INTERIM REPORT ON FAILURE ANALYSIS OF PIPING TO SAFETY RELIEF VALVE AT CAROLINA POWER AND LIGilT, II. B. ROBINSON UNIT #2 I 1. Introduction Westinghouse Nuclear Fnergy Systems requested the Research Laboratories to conduct an analysis of the pipe connecting the safety relief valve to the main steam supply pipe system of 11. B. Robinson I Unit #2. This report summarizes the results obtained to date in the failure investigation.
- 2. Description of Failure A failure of the piping which connects a safety relief valve to the main steam line in Loop #3 of the 11. B. Robinson Unit #2, occurred while a check was being made of the set point of this relief valve, SVl-4C (also called #9 in sequence of check tests). The operating con-ditions of the steam line were 900 psi at a temperature of about 530 F.
The section of pipe which failed was insulated to a point approximately 3" above the f ailure which minimized the temperature gradient across the failed area. The failure occurred in a reduced section of a 6" pipe which connacts the relief valve to the 26" steam line (Loop #3), Fig. 1. I As mentioned in the incident report, the pneumatic test device was connected to this valve whict loads the spring so as to balance the seating press.ure on the valve disc. The operator was opening the valve regulator to balance the spring force when the failure occurred. The f ailure pattern, surrounding damage, and trajectories of the valve and exhaust chute indicate that the pipe first opened up on the West side, directly opposite the exhaust (which is called East), emitting steam in a fan jet which sharply cut the insulation on an adjacent pipe, Fig. 2. 1-1 L__ _ .
The pattern of the insulation cut and fracture face suggests that the jet had a f an angle of about 80* wide and was directed about 45' up from the horizontal. The safety valve apparently rotated as the pipe tore across its section. The exhaust chute broke off and was found under the Loop #3 steam pipe. The broken end of the safety valve apparently rotated and struck the auxiliary boiler stack due East, Fig. 3, and left three gouges on the stack as the valve continued in its trajectory. These three gouges conform to three regions of shear lips on the fracture face at the West position which are described later. After striking several other components, the valve came to rest on the deck below. An inspection of the exhaust elbow revealed longitudinal markings on the inner surface of the extrados. These may have been caused either by construction, by handling, or by debris being swept through with sufficient force to bare the metal, Fig. 4. An inspection was made of the remaining eight safety relief valves (SV), the three motor operated valves (RV), and the three pressure relief valves (PRV), Fig. 5. All the safety relief valves on Loops #1 and #2 had been checked for set points and several had been lifted three times in the process, in accordance with the procedure of the valve manufacturer for tests when the opening exceeds a simmer. Since it was suspected that such testing could strain the modified 6" pipe connecting the valve to the 26" main steam line, these pipes were inspected, particularly in the region where the 6" pipe is reduced in cross section and on the side directly opposite to the exhaust pipe. Figure 6 shows this area on Safety Valve SVl-1B. None of the pipes showed any deformation marks or surface cracks. It should be noted, however, that a strain of several percent is required to break the ( surface scale of this hot rolled pipe. The welds at the junction of the valves and pipes were suspect and were, therefore, inspected. The welds were sound and the only minor fault was a continuous weld crater (about .005 to .02" deep) attthe end of the weld, Fig. 6. The depth of the crater was about the thickness of the scale in most cases and is considered acceptable fabrication practice by Code Standards. 1-2
- 3. Failure Analysis Sequence The valve body and the mating section of the failed 6" pipe which joined the valve to the 26" diameter steam supply line were received at the_ Westinghouse R&D Center on the evening of April 30, 1970.
On Friday, May 1, representatives of the Research Laboratories, NES, and Ebasco Services, Inc. met to define the preliminary failure analysis . program. The analysis procedure essentially followed in this investi-gation is listed below.
- a. Photograph failure area of pipe and valve
- b. Cut failed area from valve 2-5/8" from back surface of holding test ring *.
- c. Radiograph f ailed section cut from valve and failed pipe segment
- d. Conduct visual examination of fracture surfaces
- e. Obtain replicas of entire fracture surface and conduct fractographic examination of both fracture surfaces using electron microscopic examination of the fracture replicas
- f. Obtain chemical analysis of pipe and weld metal for C, Mn, P, Si, S, 0, N, andfH as well as spectrographic 2
analysis for residual' impurities
- g. Lay out and section for metallographic examination
- h. Examine fracture areas with scanning electron microscope
- 1. Examine metallographic sections of both pipe and valve sections of failure. Record structure, mode of' fracture, and' thickness variations
- j. Obtain microhardness' survey of weld, base metal, and HAZ
- k. Obtain tensile data on pipe section at RT and 500*F Valve body minus ring removed containing weld was immediately i shipped to valve manufacturer for examination.
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- 4. Examination of Failed Component
- a. Visual Examination The end view of the valve end fracture (designated VE) Fig. 7, shows the development of the fracture. Two heavily deformed areas (A and B) are obvious in the West Direction. These were protrusions on the fracture surface which were severely damaged when the valve impacted an adjacent pipe (Fig. 3). An adjacent area, in the NNW direction, also showed damage. Proceeding from West to East along the fracture surface are shear lips, which project from the taper of the I.D. (see cross-section sketch of pipe / valve joint, Fig. 1) down and toward the weld on the 0.D. surface. At approximately the North-South axis of the valve end section the shear lip abruptly changes orientation. The fracture extends from the end of the 1.D. taper away from the weld at the 0.D.
At the East location, the shear lips terminate in a fracture area which runs from the I.D. taper to the edge of the weld on the 0.D. Another view of the valve end failure is shown in Fig. 8. The end view of the corresponding pipe end fracture (desi-nated PE) is shown in Fig. 9. This fracture surface is rather heavily oxidized from the steam which emitted from the pipe for approximately an hour during the blowdown. However, depressions in the pipe end fracture were observed at locations 1 ani ll '(near the West direction) u _- which correspond to the damaged protrusions in the same location on the valve end fracture. Visual examination indicated the fracture initiated at the West direction (at one or both locations, A and B, Fig. 7) and proceeded around the pipe in both directions, terminating at the East area. - This agrees with the observations on damage to the surrounding structure described in Sec. 2 of this report, which indicates'that failure initiated in the West location of the valve, and terminated at the East. Visual examination also indicated that the fracture had every appearance of a ductile shear rupture. 1-4 I
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- b. Radiography The valve end and pipe end fractures were radiographed to determine if other cracks, voids, or weld porosity could be observed l
l in the areas adjacent to the failure. Type M film was placed on the I.D. of the ring sections and radiographs were obtained using a 220 kv X-ray unit. Approximately eight radiographs were obtained on each ring section. One indication of a defect approximately 3/16" long and 3/32" wide was observed in the valve body in the SW location. This defect did not extend into the weld. One acceptable radiographic indication was also observed in the weld in approximately the North orientation perpendicular to the circumference of the weld. Neither of these indications were associated with the fracture and no other evidence of defects was observed.
- c. Chemical Analysis The specified material for the pipe was ASTM A-106 Grade B carbon steel. The valve body was cast steel ASTM A-216 WCB, and the weld wire MIL-7018. Check chemical analyses were performed on the pipe and the weld deposit. Since there was no evidence that the valve itself l was associated with the fracture, check chemistry was not obtained for l the valve body itself. Samples were machined from the West and East l
l locations of the pipe and the weld. The results of chemical analysis l for C, Mn, P, S, and Si are listed in Table 1. Also included are the specification limits and the vendors heat analysis. The chemical analysis data show excellent agreement between the vendors heat analysis and Westinghouse check analysis. The composition was within specifi-4 cation for all elements checked, except that the manganese content was slightly higher than the specification limit of 0.90. The qualitative spectrographic analysis listed in Table 2 also did not reveal any compositional anomalies. 1-S
- d. Metallographic and Fractographic Examination Replicas were obtained from the entire fracture surface of both the pipe end and valve end fractures prior to sectioning for further examination. The fractures were then sectioned as shown in Fig. 10. Various sections were then examined in a scanning electron microscope, and mounted for standard metallographic examination. Photo-graphs of the indicated sites of fracture initiation, (valve end Section A and the corresponding pipe end Section 5) are shown in Fig 11.
Macrophotographs of the fracture surfaces are shown in Figs lla and b. Cross sections taken through the weld and pipe end fractures perpen-dicular to the plane of the fracture are shown in Figs. 11c and d. These sections were macroetched to reveal the weld and heat affected zone. The cross section photographs show that extensive necking had occurred in the fracture area, indicative of a ductile shear fracture. Furthermore, the fracture plane was some distance away from the weld HAZ. Figure lld clearly illustrates that the failure is located at the end of the machined I.D. taper. Figures 12a and b are higher magnifi-cation micrographs of the valve end fracture at Section A showing extensive internal void formation, again indicative of a high strain failure. Rather extensive deformation of the grains can also be seen from these photomicrographs. Figure 13 shows photographs of the fracture of Section 1 (pipe end) (opposite Section B of the valve end) obtained with the scanning electron microscope. Section 1 is immediately adjacent to Section 5 and within one of the two areas where the fracture apparently initiated. At the mid-wall location the fracture was predominately intergranular, Fig. 13b, changing to predominately ductile shear near ( the 1.D. and 0.D. locations (Figs. 13a and c). The shape of the dimples on the fracture surface indicate that at this section the fracture initiated near the center of the wall and proceeded toward both the I.D. f and 0.D. Exaatnation of the fracture surface from the other location suspected to be the origin of failure (Sections 6 and 11) showed a pre- { dominately ductile shear failure but the dimple orientation was not as k 1-6
l 1 l l i l well defined as those shown in Fig. 13. Thus it was not possibic to i clearly define the direction of fracture propagation in this area. The results obtained to date do not permit a clear determination of whether f racture initiated on the 1.D. , 0.D. , or mid-wall locations. Photomicrographs of valve end Section B, the other damaged area near West, are shown in Fig. 14. The macrophotograph of Fig. 14a shows the portion of the surface which was damaged by impact, and also reveals that the fracture occurred some distance from the weld HAZ. Figure 14b shows several longitudinal cracks in material apparently associated with inclusion stringers. A corresponding section frcm the pipe end of the fracture taken approximately 2" from the fracture plane (Fig. 14c) shows several of these inclusion stringers which were generally observed throughout the structure. It does not appear, however, that the presence of the stringers, orientated parallel to longitudinal direction of the pipe, are responsible for any significant strength reduction. . The area where final separation occurred (pipe end, Section 4) is shown in Fig. 15. The cross section micrograph shows evidence of extensive deformation by bending at the 0.D. -surface. Scanning electron micrographs of the fracture surface of this section (Fig. 16) show - essentially a dimpled ductile rupture with the fracture propagation in this section being from 1.D. to 0.D. This observation further supports the conclusion that fracture initiated at the West location, proceeded to East, with the final termination of the fracture being at the East location. At approximately the NNE location, the fracture penetrated the heat affected zone, as illustrated in Fig.17, which shows a , globular pearlitic microstructure near the fracture tip. Again, evidence of extensive plastic strain near the fracture is evident from the micrographs. Fractographs of this section!(Fig. 18) show the fracture to be primarily intergranular near the I.D. and ductile shear near the 0.D., with the shape of the dimples indicating that the direction of-fracture propagation was from I.D. to 0.D. The intergranular areas of' f 1- 7 ' , u
J l fracture arise from the fracture path following the boundary between the pearlite colonies and the ferrite grains. A mixed mode fracture of this type is not unexpected in A-106 material. Figure 19 shcws matching cross sections (valve end Section C and pipe end Section 2) which again show the extensive necking near the fracture and the fact that most of the fracture path is removed from the heat affected zone. No area on the fracture showed evidence of any fatigue damage.
- c. Dimensional Measurements Dimensions of the pipe end near the fracture were obtained from mounted metallographic specimens, and the results are listed in Table 3. These data show the taper length and width of the reduced section to be fairly uniform (within about 10%), hence machining eccentricity of the weld preparation does not appear to be a factor in the failure,
- f. Hardness and Tensile Results Hardness data were obtained on representative sections of the fracture. Figure 21a shows the hardness traverse obtained from the center of valve end Section 2. The hardness ranges from about 165 DPH in the pipe removed about 1-1/2" from the fracture up to 216 DPR in the heavily deformed area near the fracture. Hardness values for the valve end Section A are shown in Fig. 21b. The hardness values are normal for these materials.
Tensile specimens were machined from Sections 8, 9, and 10 of the pipe end failure and tested at room temperature and 500*F at a strain rate of 3 in/in/hr. In addition, a severely notched specimen was also tested at 500*F to determine the notch sensitivity. The notch consisted of a 0.010" slot with 0.005" radius for a depth of 1/8" on each side of i a 3/4" section of full thickness pipe (about 0.4"). A comparison of the 4 results of these tests with other data for ASTM A-106 Grade B pipe material is given in Table 4. Photographs of the fractured tensile specimens are shown in Fig. 21. The comparison shows that the pipe . material is normal in strength and ductility and that the ultimate I i { 1-8 l
l l l strength is not reduced by the notch. It is quite probable, however, l that local yielding occurs at a much lower stress than that reported for the yield strength which is for an extension,of about 0.002". The fracture surf ace of the notched specimen enables a comparison with that of the pipe fracture. It can be seen that the oblique shear fracture of the test specimen duplicates that of the pipe. The surface appearance is dimpled shear in both the test specimen and the major portion of the pipe failure.
- 5. Analysis of Results The results of the investigation availabic to date indicate the following:
a) The fracture initiated near the West location and proceeded around the pipe in both the clockwise and counterclockwise directions, terminating at the East location, b) All areas of the fracture showed evidence of extensive plastic deformation, c) The I.D. fracture path coincided exactly with the end of the machined I.D. taper. d) The fracture path was not associated with the weld HAZ although it did penetrate the HAZ in several locations. e) No significant materials deficiencies were found with respect to chemical analysis, microstructure, pre-existing defects, or mechanical properties, f) No evidence of fatigue damage was observed in any area of the fracture. 1-9 l
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J
, l j
1 Conclusion l While this is an interim report, sufficient information is available to identify the mode of failure. The lack of any material deficiency and the extreme plastic strain on fracture surface indicate that failure was caused by ; overloading. The fracture surfaces and component trajectory dictate that f ailure initiated on the side opposite the exhaust pipe and that ;i. bending and tearing occurred about an axis perpendicular to direction of the exhaust pipe. The peak stress must reach the ultimate to initiate failure. ' ' The preliminary stress analysis of the static condition does not account for the overload. There are mechanisms which may be postulated for generating the required increas in stress. The investigaticn of the \ sources of the overload is given in Addendum No. 2. Further investi- , gation is in progress on the failed pipe and on a sample from another valve location in an attempt to more precisely locate the origin of ** failure. ,
'I s
x.-- ene e c F l e ?- 4 k 1-10 ,
'> l 1
.\
.N
1 n. P L \ s I Table,1-1. Chemical Analysis Results l G, ~ 4 {, Pipe A-106 Grade B k (- Pipe End \ l'ipe End, Youngstown S&T Co. Element "W" "E" Heat #45601
~
Spec. [* O.25 ,* h'j
,s 0.26 C.30 max 0.23 arbon 3
0.26 '0.26
- Manganeuc
'A 0.86 . 'O.86 0.29 to s 0.80 1.06
' 'N Phosphorus 0.002 0.02 ' .0'.048 max 0.008 a* Sulfur 0.019 0.026 0.058 max 0.024
~~
0.022 Silicon 0.22 0.21 0.10 min 0.21 Oxygen 0.0003 -- Nitrogen 0.006L 0.0053 -- Hydrogen 0.0002 -- ( 4
, s
' Weld Wire MIL-7018., i
~ +
3'
.- Valve End ~ Valve End
'k , - Element S j '
"E"_> Spec. Heat #930G
^*
's ' 0.059 O.057 d, 0.12 max 0.07 Carbon 0.061 0.066
.. Manganese < 1.03 <-
l'.00 0.90 max 1.01
-(* , Phosphords 0.013 0.013 0.03 rux ' '
O.01 y* Sulfur 0.019 0.01J O.04smax 0.015 Silicon. 0.48 0.43 0.80' max s O.39 0xygEn - 0.037 1 *- -
~~
Nitrogen 0.0153 ' ' O.0112 ' Hydrogen 0.0002 " % ;- - y- .. c > x,
~
,"l
. "\ r, , , . .
. .- g .. * - Cr - 0.01
?- .
Ni . ' ' EC Ni - 0.01
%" f( '
Mo - 0.01 t** "V - 0.01
'+
,- s 3
e . v *
'tO +
+
* ~~
y- 5 ? E 1-11 g
\,g
'E .
s ,., ,
'#' ^ '
\ k g ((
_"e *
\v .g #
- / a J .p . ,e
. 1 # '
_[__._____ _______.lu.-_... _ . _ _ .-. ..-]
Table 1-2. Summary of Spectrographic Analyses
- of Pipe and Weld ,_
Specimen Specified Analysis W Designation Material Fe Si Mn Al Cu Sn Cr Ni V Pipe End-East ASTM A-106 >10 0.16 1.1 0.04 0.02 <0.01 0.04 <.02 <0.002 Pipe End-West Grade B >10 0.19 0.8 0.04 0.02 <0.01 0.04 <.02 <0.002 Ti As W Zr Co Pb g Mo
<0.01 ND <0.2 ND <0.01 ND ND 0.01
<0.01 ND <0.2 ND 0.01 ND ND 0.01 y ,
Fe Si Mn Al Cu Sn Cr Ni V U Valve End-East' MIL-7018 >10 0.34 - ND 0.02 <0.01 0.34 <.02 0.01 Valve End-West >10 0.34 - ND 0.02 <0.01 0.04 <.02 0.01 Ti As W Zr Co Pb Al Mo
<0.01 ND <0.02 ND <0.01 ND ND 0.005
<0.01 ND <0.02 ND <0.01 ND ND 0.005
- Qualitative - Accuracy + 100%.of Amount Present t
I l s
Table 1-3 Wall Thickness Variation at Pipe Failure Loca, tion I I
- -as 16--
l 1 l Measurements taken from Photographs , of the sections of the Pipe End as , shown in Fig. 1. g N I l 4 l*- e- --e
. ~ ~ .
Pipe End No. & Direction A B A-B C 4 (E) .445 --- --- --- Taper cut 13 (ESE) .453 .197 .256 .453 Deformed 13 (NNE) .441 .142 .299 .433 7 (N) .433 .126 .307 .457 12 (NNW) .433 .126 .307 .441 2 (NW) .401 .110 .291 .386 Necked area 1 (WNW) .394 --- --- --- Taper cut. 5 (W) .401 .118 .283 .425 6 (W) .417 .138 .279 .464 11 (WSW) .413 .126 .287 .453 14 (SSW) .394 .126 .268 .394 l - I l 1-13 1 1 l l 1
-. .__ - _.. = - - _ -. . _ _ _ - - _ _ - . .-.. - - .
r Table 1-4. Comparison of Mechanical Properties A-106 Grade B llot Rolled Pipe l 'l Ultimate Yield Red of i Strength Strength Elongation Area Temp. . Source ksi ksi % % *F 75.7 48.5 38 II) 55.4 RT Mill Test 73 37 35 II) 57 80 ASM llandbook
" " 72 30 26(1) 55 500 check Test:
Pipe Section( ) 75.4 50.5 33.9(2) 58.3 80 74.6 42.6 32.7(2) 50.1, 500 I Notched Pipe ( ) 81.7 51.3 -- - 500 l i I (1) '" gauge length ( ) Gauge Section 0.375 x 0.4 x 1" ( ) Pipe Section 0.4 x 3/4 x- 1" Notch Section 0.4 x 1/2, notch 0.010 width x 0.005" radius 5 l 4 4 1-14
1979-1 j i
- 29-t/2" >
\
EXHAUST CHUTE I 5.375"l0 9. 0" 0D --> SAFETY 7 VALVE Sy y 6.065"lD I l/ WELD l BACKUP RING )
-_ __ . _ _ . _ FAILURE AT END OF II N, TAPER 26" OD Nm STEAM PIPE 5.76t"1D PIPE h
WELD JOINT Figure 1-1. Safety Valve Connection 1-15
- i i
l \ l i i -l l
} [VW.}
{! g e
;>r }
Il L .1 - c av
??l <
y 1 3 i iw Y.; ;
) ,
~ ~ "
/..w~1es '
~ i .. f,M p '
J3; '*?dt.- . .~y w Figure 1-2. Adjacent Pipe and Insulation Cut by Steam and Water Jet 1-16
~~
g:yb[:'(, 2
. ? , f_,'K;7:}.
4 f<>.::%: .g;3 r.. 4, .. ; _.e , 1,2
- f;~ka:;-.
s :j ll ' i
~j )' l 1J . l , ,
r
. ,- r 4 .,j s~ )
-n:w. , " -
;-...r ' . (('N 1, t Figure 1-3. Pipe Struck by Safety Valve 1-17 1
+
4 L
\
U : k, , :t,,/ .. 9.,, i Figure 1-4. I.D. Extrados of 3SV4 Exhaust E I 1 l , 1-18
1979-2 N FAILURE POINT 3sy_l y\ RV 2 -3 -4 DRAIM CONTAINMENT I N\ RV 2SV-1 -2 -3 -4 i DRAIN LOOP 3
% LOOP 2 RV iSV-1 -2 -3 -4 RELIEF SAFETY n VALVE VALVES LOOP l DRAIN A B C PRV-1325 REL EF VALVES D
Figure 1-5. Piping System l l 1-19
y-t dDM.,
-.o
?~
.;;. 7, . ,
1,s .. .% : s N' ',ly c e- o,cy.; ;o. g+% ,
. . ;y ,i yc
- s ;*h s:l5 A,As '- j:
?mm;& Rv~,w..,
n: n
.l 1
=
'_.,y, M.,y o, _~, ,, ._;
\,
Figure 1-6, Weld Area of Safety Valve 2SV-1 1-20
\ l N p-" p . B. >
.- A- ^
m l H ~-
, , g G - J: -_ -
c u. :
, ^ '
~
._l
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E l- i .t;u re 1-7 !!nd l'iew of' Fract ure on \'alve lind (\'li) 1-21
l l l l l l l l l
- l. .
,n%f=p tyg y;pw.,
' f[ y;'
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m.. l Q: 3 ,
%n 1 $..<
j . t t .n ., i f- . 4 t i i l l
._, _ ..m.2_ MrdOL Figure 1-8. Side View of Fracture and Valve Body l l-22
W
=-,, ,9 lill,111}111P11111l111J11
, , IIltlll[pl1{f11q111{l11l111 4
10 6" 9 2' 3 8 l N 13 4 7 14 -J l E Figure 1-9. End View of Fracture on Pipe End (PE) 1-23
/
I
./
11 a B gtg3;
,. s W +- *E yalve End k' '
C . l g W' i D F J
+
N t 12 13 2 1 5 4 W+ +E pipe End 6 14 11 s 10 9 8 Figure 1-10. Portions Cut From Pipe and Valve, After Sectioning for Further Mechanical and Meta 11ographic 2 valuation, Note Section Identification 1-24 7
*F*
?l#, Valve End
&,&gw:e>@g:
.i o,g*
V,pf,QC'
._W I.D. 10(',j%
'.l74'; Section A
.?M
'f?b;2 3 dI :1M[ff ' '
Cross-Section x
& )1,u '.:" fi
. ,..v ,.
}{'l ;l;'~}p;$4 ;&, Fracture Surfaco 0.D.
h -Q':' ', , 8 Section A i$>a$h.;: (~ ,
- d. Cross- '" - * " - - ~~ - A Section l
Fracture Surface I.D. b. nn--- _w r. O . b. 1.D. s i . . x, - f' i i: 1 L-_._ -
,, . . _ w. a. 2 mad O.D.
Pipe End Necked-in Region Section 5 Approx. 3X Figure 1-11. Corresponding Fracture Surfaces (a,b) and Cross-Section (c,d) From Pipe and Valve Ends Through Necked-in Region, blacro-Etched to Reveal Weld and lleat Affected Zone 1-25
,y ;: ; - r -~ ,;- - (7 l f *j,3 l' l [i l ,
6 . fi I ' 4 .,, Valve End t $ { Necked-in Region 1
.? f
- lj.
j'. : i: , I4, 500X l ff'9 W ' ',
'l M
/ :
yc ( *J f 4
, r'
\
)
F 9 Fracture Surface 200X S t on < h. n er h Vo s To d l During Failure by Excessive Tensile Strain 1-26
l o)) h9, e
.e'44 n ' v '4 ~ ~'
.e> a. " Dimpled" Fracture
~g ..,..,.
- ip , Surface Predominately f Ap;j!g.#fS r
ii ' 4k
- y, m-
.;; T-;
' ,w i
, y- Fracture Propagation )
?-
{
. . }W' .
.8 %
\
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- b. Predominately
; -' . Intergranular Fracture e i r- <
Y ,P $ b t
$.d Qi l.]: & 1 ,
e. h!??-l$g, 1 N[?-l- f{D - - .h e r, N .d yi f Y ,, %, Center SL? k, ,; , w y '~' gp
- 7 c. " Dimpled" Fracture g , , ' r, Surface Predominately I V 4*.,f ir [,' Ductile Shear Failure j
m * , W$,
? < w,
,~w
&,.&,(U x4y,.a .sr W
%%/ '
4
~~ / e, Fracture Propagation
'es -
1000X 4 '
^
w v [A s+ f _ Near 0.D. Figure 1-13. SEMs of Fracture Surface on Pipe End Section 1, Necked-in Region 1-27
.i l z e ini, Su tion H
., ~ . , , pa 'n I.D. g. # O .1) .
(' yj'& \ Q _:, ' Y'N }' U [ ' ~, ',w k ,, *yf',W Y
]NI ; &,'(-hb Y *?
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,a .l '&i, Tc0, - f., , :; j
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'd
, j <
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,' f ,p s / ' . ~ l , G. "eV J Portion
. , /j .,. >
.?u
>a;
; %. ., ) .. '; 4.i
< . i s ,, 'e
,i ?
machined e
, # f , ', <g.&' j' ' . ;j ../ / ' ,, '
n.>
/
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out for chemiccl
,f
; K's; J *l';
s' 17 <) . [ a na l y s i ., y, 1 . g: ./ of weld
- > y- <
.f ,
-V -
metal 1 A ,,. s i ' s. / s [ ,s,eh ,,;9y~ ~ [ I 'pp- ' ~6]
- p. hp ' , , ,
$}t p.. ;;
n% / g, , j y;: ,* #M
. 6' ,g ,')p'Q.
' [
h) fI g. % ,: ., ji 3-h, bi . rj ,
- f J' q -
gp 7-f g 4% wi, > c .: : t ! [' - ~ d i. t ' ', c s k r ,.;. 3
' .i4 a.
i i y,, .
- % % ?+ Approx. 3X l
T ?:1 r sfm ,,,/w/ T., '42 o g?- D
, ; Portion deformed by
@j impact damage
[1 N\'sa, y ,. , l .
~
V- s asifan-4."g i .M' i c. L p
- d, 'b b.,. %
7
\ Wh&g.f-A
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\
i WMl -
.p_gl Q, %
< n 4.w%y>j.
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- I 9.
w z. iL ' '#'wM.i;X Nhl: l f = >e (l. 5 Qn V nr . -s f',,h %; wi,r 2.
- g. @Yp-QQ Section through PIPE END at a I
- 2 17 C distance of 2" from fracture I by .
'k '~f y - W.- !
h 500X
" ^
h g ., f'Ns;g
'/ ' hy -f kr) e , (La j
e .
,; s y wr .: s G
l V. . : Q:> ~;)+.. t
,- '~. ~_
- z. ).
$$L s ..; kb D;_ m ,,.
l' i cu re 1-1l. %e ro-Et ched Cross-Sect ion Through Deformed V.\llE END in Necked-in Region (a), l. i gh t Micrographs Show That hial f rack l'ropagat ion Occurred Along St ringers of inclusions (b) hh ich a re Dis t ribut ed Throughout the l'i p e (c) 1-28
l._-_-___---. 0.D. I.D. l l i Fracture Surface of
'I ~,[ pipe end portion l
,L' i -
.cy; 3
.I c,- ; East ' '[4[4 13
+
., W.5 1 ,
~
- k. .:2
, 'm
. cj:s l Cut for Section 4 l
*x .}
a d
#!M i r s. "
. 1 1
1 i
; 3 ., i l
i ! Approx. 2X
,,33,c, l I
I l 0.D. I.D. Section 4 Cross-Section
; l 1 .
l
.- , t 6 i '
( ;f ' c y, j Approx. 3.5X Figure 1-15. Fracture Surface and Cross-Section Near Region ) of Final Rupture, Note Considerable Bending of Tube in O.D.-Direction 1-29
l.D. . Near I.D.
- Fracture Propagation
,4 -
' e ay .g>-
5i " . . v Center Frac',ure Propagation
- hear 0.D.
0.D. L MNkg
)
Figure 1-16. SEMs of Fracture Surface on PIPE END, Section 4, the Region of Final Rupture 1-30
,l ll e
c f r x 2 o 5 o l e _ I . r u Jb. , - ' X t c _ 0 0 . u s 4 Y D r
. t s
g p]%: %g '
~- V 0 s r r o
'- a c
'k m5 ei
,'e X, ' . ,. k*
NM t
\ -
ee r t
/
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=
ui t l i,',
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iv ob T i ' t
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, 4 . - cG r e T 4' - -
a S e
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/
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' } / A Q l X
0 oI r r s a 0 kg 5 'i
'., 5 h s
'
- j
- % T a a ' k - f' n oe
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7k' i n t o cZ
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4
, s k.
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, [
a ( s c 4, oe
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s
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t
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ot A
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- O gt oa
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N M ne t P h gp X ii 0 LL x o 5 rX . p3 7 D. p 1 I A - 1 e e
" f.
c r 5 a u h.- - Qy, .A - ; f r i g l }y ,l
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s F f ,- - b' ;sWj q e r u 3 ?d t O , c
$' I l;.
a
. : r k
C. ' f L m o r f D. T,- "2 0 l l ll,
_ _ _ _ _ _ - - -_ - _ __ ._ -_. - _- - -.- .- .. .. . ....... . - .~. __ _ _- i l I.D.
-- y, i
' ^
j ,. l ,P q Intergranular fracture l I l 1 - i l s.c , C
- r. .
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- a. Near I.D.
1000X i x!p ' ' 'y ,' ' A ' ,
, =;; ,, Predominately ductile shear failure
, [.
(l *l l y <* sf
,.r
- p. ..,9 ' .
s i, f C -
.. .q .
m E *'
.~ ,b
- b. Near 0.D.
0.D. ) Figure 1-18. SEMs of Fracture Surface on PIPE END, Section 7 (approx. NNE) 1-32 1
, -s.,-
, r.0 ~:$
- .j' [ ily? Valve End
-? ". .f:,
t,<y .4 ' ,
- c. w, K I.D. ,
.! q pi - 0.D.
l r; l,
.y.
3
\. - ,N
.~",
,:s m '-
9 f.,Q Section C
, . ;4. %C- . f).. , l y ;;~ . .q,. l 4 ,24 ^.
,. , .-;. r, ,; ;_
. ;p .-
4 .e .:
, +. . ,
~y >
-- g e. ,
*o- .r . .
., . f:=
[ *,. dSNGe s ^
.,a v'Tih. -
sg;p Ki i
- +;.f ,.7 - + ,
.?-by1,, .e.
)[ C,5l4t "hU r
-ly~ Pipe End l
. i,s%
I.D. i} f aDt3 0.D. AL i t sti i Section 2 i l i, i
..b~ .
- l. ~h oa -
Approx. 3.5X Figure 1-19. Corresponding Sections From PIPE and VALVE ENDS, Macro-etched to Reveal Weld and IIcat Affected Zone i i l l 1-32 l
i l l l deformation OO O OM O
'l 1
l 110 I l 2tl, I l I65 g l 164 l 34 118 212
/44 VICKERS DPli IIARDNESS sur 1-20a. liar ess Traverse valve End )
20/ 2N E09
' I 222
- [l G 9 Q Q $0OO bO l 9
/
\- 233
/66 /9/
I L /9/ j95 , 237 2IO , 200 e. Figure 1-20b. Ifardness Traverse Valve End Section A ) 1-34
- -_ .a
p 3/4" ,j s ~- Sections cut I from PIPE END 2" .. l portion i Section No. 8 9 10 , W:b'.[}P f.
. ;j )
x erf= . 3 . W. b' After tensile testing 7, /->' , ' ,y ,
- ".,. e. y ' ,. ,
;A
{l -l +li , 4
);M~s a
g { T'4hj(+e[
~
'[P W7Me- %,
k
. <%. s.'
Notched specimen
+
Unnotched specimens Cage section 0.375 x 0.4 x 1" Test temperature: 500*F RT 500*F Figure 1-21. Sections Cut From PIPE END portion Before and After Tensile Testing
)
1-35
ADDENDUM 2 H. B. ROBINSON UNIT NO. 2 STEAM PIPE BREAK DESIGN ANALYSIS REPORT
- 1. S CO.'E The following report contains the stress analysis of the failed valve nozzle for steam conditions at the time of the incident.
Additional stress analyses were performed on all other steam system relief valves in the plant at design static and dynamic conditions. A summary is included of the quality control performed following the incident on the 12 main steam line safety valves by the Crosby Valve & Gage Company. i Modification to the nozzle connections of the 12 Crosby main steam safety valves and 7 other relief valves are in progress.
-11. CONCLUSIONS The nozzle failure is a ductile failure as stated in Addendum No. 1.
However, no factor or condition detarmined by way of the post incident investigation has been objectively assigned at this time as the cause for the failure. i i 2-1
III DISCUSSION A. STRESS ANALYSIS OF FAILED VALVE N0ZZLE l 1. Analysis at Incident Conditions l Based upon information presented in the main body of this report. the incident occurred during static conditions with Main Steam pressure at 900 psig saturated and 50 psig on the diaphragm of the air lift l device being used to check the setpoint of the valve. The primary and secondary stresses to which the nozzle was subjected at the time of ; the incident are indicated below: Summary of Stresses Prior to Failure - @ 900 psig saturated static pressure. Primary Stresses I Longitudinal Pressure Stress 4660 psi Deadweight Stress 960 psi Total - Primary Stress 5620 psi The code for " Power Piping" (USAS B31.1.0) allows 15,000 psi plus 20 percent for-short term loading of ASTM A106 GR.B at temperatures up to 650*F. Secondary Stresses Bending Stress due to eccentricity ( ) 2360 psi Bending Stress due to change of thickness ( ) 6600 psi Axial Stress due to temperature gradient ( ) 500 psi Total - Secondary Stress ) 9460 psi Total - Primary + Secondary Stress 15,080 psi Note (1) - Computed using flange formula for stress at end of weld neck flange. Note (2) - Computed for long cylinders with step change from .280" of Sch 40 to .432" of Sch 80. , Note (3) - Based on a 2*F. Temperature gradient across the insulated nozzle. Note (4) - Secondary stresses have been added although they may be of opposite sign. The mill tests of the actual nozzle material reported a yield strength of 48,500 psi and an ultimate strengtit of 75,700 psi. Results of laboratory tests (Addendum 1) substantiated these strength values. Mechanical test results givea in Table 4 of Addendum 1 on specimens from the failed nozzle material at 500*F give a yield strength of 42,600 psi
)
l and ultimate strength of 74,600 psi. 2-2 _ , __ __ -4
- 2. Analysis Under a liypothetical Valve " Pop" Condition Crosby Valve & Gage Co.pany advised that the quantity of steam passing through valve seat would be in direct proportion to valve lift. The tabulation below assumes that 15 percent instantaneous total steam flow passed through the valve for a 10 percent valve lift (5 percent more than expected).
Summary of Stresses - With 10 Percent Valve Lif t and 900 psig steam (Saturated) Primary Stresses Longitudinal Pressure Stress 4660 psi Deadweight Stress 960 psi Exhaust Thrust Stress ( ) 6140 psi Total Primary Stress 11,760 psi The total primary stress for the '10 percent valve lift condition is less than the yield strength of 42,600 psi at elevated temperature of 500*F. ) Note (1) - Valve " pop" action is a significant blowing of the safety valve with distinct . audibility and visual evidence of steam relieving the valve. Although this can occur with less than a 10 percent valve lift, a 10 percent valve lift has been assumed to insure conservative review. Characteristic valve lifts for a simmer condition with audible and especially visual evidence of escaping steam needs only 0.001" lif t or 1/10 of 1 percent lift. Note: Full valve lift for this valve is 1.125" (Refer to Exhibit 1) . Note (2) - Based on thrust 900 psi being equal to (900/1140)l/2 times thrust at 1140 psi. Secondary Stresses Bending stress due to eccentricity 4940 psi Bending stress due to change of the thickness 6600 psi Axial stress due to temperature gradient 500 psi (Estimated) Total Secondary Stress 12,040 psi g Total - Primary + Secondary Stress 23,800 psi 2-3
B. REDESIGN OF SAFETY VALVE N0ZZLE A redesign of the 12 main steam line safety yalve inlet nozzles was instituted to increase the conservatism of the stress level and to decrease possible struss L.tensification factors. The extent of the revis2on included increasinh wall thickness to 8" Sch 160 and also enlarging the nozzle diameter to udsrease thickness differences between the nozzle wall and valvs bode woll, The revised design is included as Exhibit II. The result of d stress en,1yets Iwr the revised nozzle under valve setpoint conditions which include a f ull capacity blow are shown balow. Summary of Stress - Valves SVl-1, byl-2, SVl-3 and SVl-4 for each of three loops-New Nozzles 8" Sch 160 Primary Stresses @ 1140 psig Steam Pressure (saturated)
' ' ' Longitudinal Pressure Stress 1900 psi o Deadweight Stress 210 psi \
Edtchquare Stress 1450 psi Exhaust .Tbt ust Stress 10,100 psi Total Primary Stress 13,660 psi The code for " Power Piping" (USAS B31.1.0) allows 15,000 psi plus 20 percent for short term loading of ASTM A106 GRB at temperatures up to 650*F. Wind stress has not been included since it is less than earthquake stress. Wind and earthquake are not considered simultaneous loads.- ~ Secondary Stresses Secondary stresses are considered negligible since physical configuration of the new design has eliminated major thickness difference between the valve body and nozzle. C. STRESS ANALYSIS FOR REMAINING MAIN STEAM RELIEF VALVES Based on the approach taken for the modification of the Safety Valve nozzles, the seven relief valve nozzles have also been modified as shown on sketches in Section D. h 1. Analysis for the 3-6" Power Relief Valves, RV1-1, 2 and 3 2-4
The results of a stress analysis for the revised nozzle connection to the above valves at the setpoint conditions 1140 psi (Maximum pressure level) 6" Scb 80 Nozzles Primary Stresses at 1140 psig steam pressure (saturated) Longitudinal Pressure Stress + 3540 psi Weight Stress - 90 psi Earthquake Stress + 1960 psi Exhaust Thrust Stress + 7660 psi Total Primary Stress 13,070 psi
- 2. Analysis for 8 Inch Steam Dump to Atmosphere Valves, PRV-1325A, B, C and D The results of a stress analysis for the revised nozzle connections to the above valves using the maia steam line design pressure 101 psig is tabulated below:
8" Sch 100 Nozzles Primary Stresses at 1085 psig Steam Pressure (saturated) Longitudinal Pressure Stress + 3160 psi Weight Stress - 80 psi Earthquake Stress + 1670 psi Exhaust Thrust Stress + 10,730 psi Total Stress 15,480 psi D. SKETCHES OF N0ZZLE MODIFICATIONS TO MAIN STEAM RELIEF AND STEAM DUMP VALV.ES The revisions co the relief and steam dump valves are shown on attached Figures 1 and 2. Figure 1 shows the revised nozzle connection to power operated relief valve RVl-1, RVl-2 and RVl-3. Figure 2 shows the revised nozzle connection for main steam relief valves PRV-1325A, B, C and D. l E. POST INCIDENT QUALITY ASSURANCE ON MAIN STEAM LINE SAFETY VALVES l Exhibit III summarizes the Quality Assurance performed by the Crosby Valve & l Gage Company on the 12 Main Steam Line Safety Valves. The tests performed by Crosby enabled them to recertify the adequacy of all 12 valves for intended service. , ) 2-5
C A R O L I N .L PO W E R ( L I G H T CO. H . E>. R G E I N . '. N U N I T N o. 2 SKETCH OF 14 0 ~2 Z l. E M O D I F I C A T I.O N FOR MAIN STE AM RE LI E F VALVES FULL SCALE D W6. N o. SK-9 9 4 3 - M. 4 6-30-70 P. A R M E L V.A L V.E 3 0 D Y p 70 0 D-RVl-1 R V I - 2. r 'mhw ,
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j l APPENDIX A t i 1 Postualtion of additional conditions that would be required to cause failures
}
1 l Full capacity steam blcw at steam condition at cLne of incident. i The Postulation of a full capacity steam blow is made for the purpose of exploring a possible mechanism for the nozzle failure. It is emphasized, ! however, that this is a highly improbable occurrence in view of all the information presented in the main body and in the Service Report Appended ] to Addendum #2 of this report. To achieve a full blow, either of the j following conditions must have been satisfied.
- 1. The set point of the safety valve was less than 1000 psig (normal l setting is 1140 psig). The 50 psi which was reported during a check of l
the set pressure in the valve lif t device is equivalent to an additional 100 i psi steam line pressure. Thus, if this were the case, the 900 psig reported in the steam line could have caused the valve to open. However, this would require that either the safety valve set pressure as certified by Crosby's j factory report was incorrect, or the set pressure decreased during shipment, handling or after installation.
- 2. The air pressure regulator was set above 120 pai. At this setting the equivalent line pressure would have-been greater than 1140 psig enabling the valve to fully blow. However, the air pressure regulator is a sensitive device requiring manual adjustment. Therefore, a deliberate action would be required to change the 50 psi as reported to allow 120 psi to enter I
the valve lift device. I Because the Crosby representative performing the valve check had paused during the test to check steam line pressure just prior to the incident' the equivalent steam line pressure could have been no greater than that necessary to bring about a balance valve condition. This condition would I ' have caused a simmering which would have been noticeable, but was not detected. 2-8 um.
ilowever, despite the imponderable improbabilities required for a full blow the following stresses would be imposed on the nozzle wall with a fully open valve seat. Summary of stresses in original nozzle under full blow conditions: Primary Stresses Longitudinal Pressure Stress 4660 psig Deadweight Stress 960 psig Thurst 40940 psig l Total Primary Stresses 46560 psig Secondary Stresses Bending Stress Due To Eccentricity 19600 psig Bending Strew Due To Change In Thickness 6600 psig Axial Stress Due To Temperature Gradient 500 psig Total Secondary Stresses 26700 psig Total Primary plus Secondary Stresses 73260 psig Should the total Stress of 73260 psig be coincident with the stress intensification factor due to a pipe wall imperfection, stresses may l result which exceed the ultimate strength of 74600 psig for this material. ) . 2-9
EXHIBIT I CROSBY VALVE & GAGE COMPANY WHENTHAM, MASS O2093
- 617 / 3 84-3121 J
r C FI O S B Y J T June 12, 1970 Mr. R. Pulito Ebasco Services Inc. 2 Rector Street New York, New York 10006
Subject:
Steam Safety Valves Purchase Order No. NY-434117 Crosby Order No. 708061 H. B. Robinson Nuc1 car Power Plant Carolina Power & Light Company Hartsville, South Carolina
Dear Sir:
Enclosed herewith are three (3) copies of Service Report covering the testing of the safetv valves on the above order, and report covering subsequent investigation and rework of the valves. Very truly yours, CROSBY VALVE .& GAGE COMPANY
/ ,
l, (/CW J.L-[Corcoran Vic,e President-Engineering
)
6 JIC /glp Encs. 2-10
~__
SERVICE REPORT Date: June 10, 1970
- 1. Customer Ebasco Services Inc.
2 Rector Street New York, New York
- 2. Location Carolina Power & Light Company H. E. Robinson Nuclear Power Station Hartsville, South Carolina
- 3. Persons Contacted Mr. J. LaBlang, Westinghouse Operations Manager Mr. G. Traylor, Westinghouse Shift Foreman Mr. L. Mazingo, Maintenance Foreman, C.P. & L.
- 4. Crosby Representatives Mr. R. A. Wright Mr. A. P. McCarthy
- 5. Valve Data 6 - 6xRx8 HCU-65W 6 - 6xQx8 HCU-65W
- 6. Comments Arrived at the H. B. Robinson Plant at 8:00 A.M. Tuesday, April 28, 1970. We met with the above mentioned people and discussed what would be required for the testing of the secondary safety valves.
These valves had been checked on Monday, April 27, 1970, at the operating pressure of j 1020 p.s.i.g. and the following noted. Valve No.10 6xRx10 - Slight visible leakage. Valve No. 7 6xQx8 - Intermittent leakage. Valve No. 10 6xRx10 - Intermittent leakage. The whispering valves were gagged and they closed off tight. At no time while checking Valve No. 10-3 at 1020 p.s.i.g. was there any evidence that this safety valve was mounted on a pipe nipple which would rupture at a pressure less than we were presently sustaining. At approximately 11:00 A.M. April 28, 1970, we began our testing on the No.1 or "A" Miin Stream Line. Both the "A" and "B" Main Strema Lines were tested with no incidence or problems.- 2-11
Service Report (Continued) June 10, 1970
- 6. Comments (Continued)
We began our testing of the "C" Main Stream Line with Valve No.10-3 set at 1140 p.s.i.g. There was no change in our established procedure. The air motor was installed and we backed off to what we felt would be a safe distance, should the valve pop. At about 50 p.s.i.g., on air motor, Mr. McCarthy paused to allow for a final check of the line pressure. The line pressure was reported to be 900 p.s.i.g. and at about the same instant there was the pipe rupture. As stated before, there was no forewarning, or anything out of the ordinary to suggest that what happened was going to happen. I feel the entire incident, as it occurred, was a complete surprise to all of us involved. I was unable to determine the exact time of the accident as my wrist-watch was blown from my hand at the same time I was blown off the "B" Main Stream Line back toward the turbine platform. The test results attached are recopied from the original data taken prior to the failure. R.A. Wright / gip (Typed 6/12/70) ) 2-12 _. _ _______________m._ . _ _ . _ _ _ _ _ _ .
i CROSBY VALVE & GAGE COMPANY Wrentham, Massachusetts SLBNARY VALVE INCIDENT REPORT H. B. Robinson Nuclear Power Station Carolina Power & Light Company Hartsville, South Carolina Wednesday, April 29, 1970 - Hartsville, South Carolina Crosby Representative, U. D. Greenlaw, arrived at site for evaluation of incident and personal contact with injured Crosby Engineering personnel. Valve No. SV-10-3 was examined on deck of plant. Discharge piping was still bolted to valve body, but riser pipe which had been welded to drip pan was missing. Valve spindle rod was broken where the rod protrudes from the top of adjusting bolt. The air lift assembly had been separated from the valve and damaged extensively. The broken portion of the spindle was with the air lift assembly. The remainder of the valve appeared to be in satis-factory condition except for some damage to the inlet pipe attached to the valve,and the valve bonnet. Thursday, April 30, 1970 - Hartsville, South Caroli *na Disassembled valve by releasing all spring tension and removing discharge piping. Valve internals were in excellent condition with the exception of the disc insert which was not free to move in valve disc holder. Factory installed lubrication was noted on bearing and guiding surfaces. Soft powdery rust deposits were observed on the bottom eurface of the valve body on the outlet side indicating that the valve had possibly been leaking. Green paint was noted on the inside o.f the flange of the discharge piping attached to the valve outlet. C ro sby-Wrentham Valve SV-10-3 was returned to the Factory for repair and was examined more fully by W. D. Greenlaw, J. L. Corcoran and Crosby Inspection personnel. The top of the valve bonnet was found to run out from the bottom end indicating that the casting had been cubjected to a force sufficient to cause yielding. All other parts were inspected and found to be in acceptable condition except for the spindle and disc insert as noted earlier. As a precautionary measure the body to bonnet bolts and nuts were replaced. In accordance with Ebasco instructions the valve body, the new valve bonnet, new disc insert and nozzle were subjected to the following non-destructive tests. Bonnet - Magnetic Particle. Disc Insert - Radiography and Liquid Penetrant. Nozzle (to the degree possible) - Radiography and Liquid Penetrant. Body - Radiography of valve inlet (pressure containing portion) and Magnetic Particle (complete) . 2-13 l l
Crosby Valve & Ga'ge Company Wrentham, Massachusetts Summary - Valve Incident Report (Continued) Crosby-Wrentham (Continued) The nozzle and disc insert met non-destructive tests. Several valve bonnets had to be magnetic particle inspected in order to select a bonnet which fully met the magnetic particle requirements af ter a minimum amount of blend gr'ading. The valve body inlet was found to have a linear indication (radiography) on the lower end of the valve inlet in one location. This indication was carefully explored by blend grinding. Liquid penetrant and magnetic particle inspection was used to facilitate indication removal and determination of type of defect. Probing showed the indication to be a linearly disposed random unconnected indication and for the most part comprised sand inclusions and in one instance a small void in the base metal. Removal of the indication was confirmed by radiography and the area repair welded. The weld repair area was checked in process by liquid penetrant and the final weld by radiography. The valve body was prepared for hydro testing to 2100 p.s.1,g. in accordance with the original specification. No leakage was found on hydrostatic test. The valve was then completely assembled with a prorated spring and tested on steam to assure satisfactory valve performance. In addition to the standard steam operational tests, air lift tests were conducted to - determine valve lif t at various pressure differentials of valve set pressure to boiler steam drum pressure. The valve was set at 600 p.s.i. for these tests using the spring designed for that pressure. Pressure Differential % Air Pressure Valve Lift Comments 30% 80 .001 Simmer 30% 80-84 .090 Pop 20% 55 .001 Simer 20% 55-57 .100- Pop 107. 25 .002 Simmer 10% 25-27 .120 Pop The test was run in two (2) modes, 1) by increasing air pressure after the valve had started to simmer to cause the valve to more fully open and 2) by holding the air pren-sure constant at the simmer pressure and raising the system or steam pressure to cause the valve to lif t more fully. The D/4' valve lift is 1.125. From the data, simmer-could be defined as a slight valve opening distinguished by stem lift, ~ steam escaping - and audible sound. Pop action under-air lift-shuulation would be defined as a valve lif t af ter simmer where perhaps 10% of 'the actual total valve ~lif t might be expected and the sound level and quantity of steam released considerably greater than at the shamer j point. We also observed that a typical time transient to reach 50% 'of specified air pressure - was approximately 10 to 20 seconds and 15 to 20 seconds to -increase air pressure to cimmer pressure of.the valve. The valve on a- simmer test or a simer and pop test would - normally be open for several seconds. A complete cycle starting from zero air pressure. , could be- performed in approximately one - (1) . minute. - It should be noted 'in reviewing . the foregoing'thatithis was under factory test conditions and. procedures. l i 2-14
Crosby Valve & Gage Company Wrentham, Massachusetts Summary - Valve Incident Report (Continued) Crosby-Wrentham -(Continued), Following satisfactory completion of steam tests, the valve body was remachined to the new approved weld prep. The valve was then tested for set pressure and seat leakage on air in accordance with the specifications and factory test procedures. May 21, 1970 Crosby Representatives, J. L. Corcoran and W. D. Greenlaw, met with Ebasco in New York to discuss the incident and determine further course of action. W. D. Greenlaw was requested to inspect another valve at C.P. & L. The weld preparation and the return of the eleven (11) valves was discussed and a program of rework and retesting established. May 22, 1970 - Hartsville, South Carolina Crosby Representative, W. D. Greenlaw, with the assistance of Ebasco personnel disassembled one of the valves which had allegedly " popped" (air lif t tests) during the valve set pres-sure testing. The valve was found to be basically in excellent condition. Some factory installed lubrication was found on the spindle point to spring washer bearing surface. No lubrication was found on nozzle or guide ring threads. No detennination could be made of possible effects of steam flow in valve body versus rust condition since the interior of the valve body exhibited some general rusting throughout. Further inspection of other valves at the site revealed that since the start of the testing all of the weatherhoods had been removed and were not replaced and as a result all of the valve interiors had teen subjected to environmental conditions and foreign material attributed to the construction activity in the area. Therefore, a conclusive opinion as to relative condition of one valve to another was not possibic. The eleven (11) valves returned for repair and rework were disassembled. The valve bodies were tachined to the new weld prep. Eight (8) inlet pipe and weld extensions were removed and are being forwarded to Ebasco in New York. The valves were then reassembled and set on air for correct set pressure and tested for seat leakage. All valves satisfactorily ! passed all tests. i W.D.Greenlaw/ gly - June 12,1970 l l Q 2-15
l l Exhibit II Carolina Power & Light Company H. E. ROBINSON - UNIT NO. 2. Main Steam Safety Valve Inlet Nozzles l i
. / D.eo \ -
VALVE
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CRobbY Dw4 N& $ 4$404$ (FJll $W i-i , $V l-2) 5 A F $4 04 fe (70( $V j-3, SV l-4) L 37l* 26* tv 8'- 5tH 40 irg-.lY- - 6.990cm. No22LF %
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l w. soio EXHIBIT III l l EDASCo SERVICES INCORPORATED
- HATERI ALS ENGINEERING AND QUALITY COMPLI ANCE l QUALITY COMPLIANCE REPORT l E.O. NO. WELC 9699.44 REPORT NO.
J j cusar oR rRonc' Wes tinghouse Electric Corporation rea m ac mo /secc mo. l JLR_ Robinson Unit No. 2 HNDOR. MANUFACTURCR OR CON TR AC TOR P,0, No, Crosby Valve &_ Gage Co Fr0-8347 June 12, 1970
SUBJECT:
QUALITY ASSURANCE TESTING PERFORMED ON THE 'lWELVE (12) N0ZZLE SAFETY VALVES RETURNED TO CROSBY VALVE AND CAGE C0 FROM H B ROBINSON UNIT NO. 2 PLANT The writer made several visits to the Crosby Valve & Gage Company to witness required tests and to evaluate Crosby's Quality Control Records of the twelve (12) Nozzl; Safety Valves returned from the H B Robinson site. Valve, Crosby's S/N 10-3, Tag SV1-4C was returned the 6th of May. This valve was involved in the subject failure. Nondestructive testing was started on the 7th of May on the original body, seat nozzle, an'd seat disc. A new yoke was provided due to excess run-out and alignment of the original yoke. The valve inlet, seat disc, seat nozzle was radiographed. All accessibic areas of the valve inlet, valve bowl and new yoke were magnetic particle inspected. The nondestructive tests were satisfactory on the seat disc, seat nozzic, yoke and bowl. The radiographic and magnetic particle inspection of the valve inlet exhibited a rejected indication located in an area designated on radiographic film 4-5. The valve inlet was ground to eliminate this defect. Magnetic particle and liquid penetrant testing was used to assure the removal of the defect. Radiographic inspection was performed on the ground areas and no apparent defects were noted. The area was weld repaired to the requirements of Section IX of the Pressure Vessel Code and local stress relieved. The re-pair area was liquid penetrant inspected and radiographed. No apparent defects were noted. Valve, Tag SV1-4C, was re-machined for seat step dimension. It's seats were lapped, assembled with prorated test spring and given a steam test to assure valve performance. The tes t was satisfactory. The valve was dis-assembled and a new weld preparation was machined on the valve inlet. The body and seat nozzic was hydrostatically tested at 2100 psig. No visual leakage was noted. The valve was assembled with its rated spring, spring set and tested at the set pressure of 1140 psig. The valve seat leakage at 90% of set pressure was zero. Eleven nozzle safety valves were returned to Crosby Valve & Gage on the 25th of May. The valves were disassembled. The valve parts were given a . visual inspection. Their general appearance showed no apparent damage to the
) .
2-17
EBASCO S ElWIC ES INC0RF0 RATED internals. All valve seats were re-machi .ed to provide seat step clearance for re-lapping at assembly. The weld ends were inachined to Crosby's Drawing SA-F54045. The valves were assembled and set using air pressure to set the required spring tension in accordance with systems requirements. Seat leak-age inspection was performed at 907. of each valve's set pressure. No seat leakage was noted during a five minute test period. The valves were painted, and sent to the Shipping Department for , boxing and shipping. _q .) ' p$%u r-JWS: loc J W Singleton,* C Representative 2-18
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Addendum No. 3 TRANSIENT ASPECTS OF A STEAM PIPE RUPTURE AT THE H. B. ROBINSON UNIT NO. 2
- 1. Transient Description ,
On Tuesday, April 28th,1970, during hot functional testing of the H. B. Robinson Unit No. 2 plant, a steam pipe failure occurred on the con-necting line between the No. 3 steam generator main steam line and a safety valve. Subsequent investigations indicated that the break, which developed rapidly, was a complete circumferential failure of the connecting pipe. Since the safety valve was blown clear, it can be assumed that from the moment of the accident, unimpeded steam discharge took place through a break area of 0.176 ft . This is slightly smaller than the " credible" steam break pre-sented in the FSAR and repeated on Fig. 1 for reference. In the prcsence of a nuclear core, protection against the type of failure which occurred would normally be provided by the safety injection signal initiated from coincident . low pressurizer pressure and level. The available instrumentation on the plant recorded a sufficient number of plant parameters during the accident to enable a comparison to be made between calculational techniques and the actual tran-sient experienced. The significant difference between the accident which' occurred at the H. B. Rob- -l , inson plant and the type,of accident analyzed in the FSAR was- due to the timely
. intervention of the plant operators. The account of the accident shows that the actions taken .by the ' plant operators were.swif t and positive but this was
~
~ inconsistent with the FSAR in which no operator action was considered. The-
} . . operator's actions 1resulted in a transient behavior of the reactor coolant
'3-1
, . . . . . ... . ... . . . . . , , , . . ... . . , , , . ,a
pressure which diverged considerably frcm the one presented in the FSAR analysis . It is of particular significance to know that the s tandard calculations do yield a realistic answer when modified only by a knowledge of the input data concerning operator action, but not changed as a result of a knowledge of the course of the accident. To demonstrate this , the FSAR analysis was repeated for the actual break size which -occurred, with the only data changes made being those corresponding to a trip of all reactor coolant pumps at 30 seconds and the operation of all three charging pumps at 70 seconds af ter accident initiation. The data was also corrected to allow for an initial reactor coolant temperature of 533 F Figure 2 shows the results of the new analysis compared to the actual plant transient. The pressurizer pressure, and the hot and cold leg temperatures on loop 3 (the affected loop) are shown and compared with the recorded plant data. The effect of the operator action of starting all three charging pumps on pressurizer pressure is verified by the computer results. The pressure attains a minimum valve of about 1880 psi and then increases as the steam is compressed in the presarrizer by the returning water level. The -pressure contin-ues to rise as the incoming charging flow more than compensates for the shrinkage of the reactor coolant due to the temperature decrease. l i Figure 3 shows.the computed pressurizer water level and an estimated actual tran-sient water level based on the measured pressurizer pressure. These curves sup- ) port the conclusion-that'the liquid level did not recede ont of the~ surge.line e into the reactor coolant loop during cooldown. 3-2 l j
Most important of all, Figure 2 shows that the FSAR type of calculation gives a conservative estimate of the transient with regard to the time scal'e and temperature change and does show the effects of increasing system mass with the charging pumps, and minimizing system shrinkage by reactor coolant pump coas t down. A delay of a few minutes in implementing these actions, or failure to carry out either one, would result in emptying of the pressurizer and surge line, in which case the pressurizer pressure would fall fairly rapidly throughkhafety injection pressure set point of 1715 psi. It appears, therefore, that the timely actions taken by the operaters did preempt the initiation of a safety injection actuation signal and avoided the need for automatic boration and coolant makeup through the action of the safety injection system. Although there was no fuel in the reactor at'the time of the accident, the operators did follow the emergen~cy boration procedure by starting both boric acid transfer pumps and delivering fluid from the boric acid tanks - to the suc-tion of the charging pumps. Under normal reactor operating conditions these actions would have resulted in the boration of the reactor coolant to prevent ( ( the reactor from' returning to criticality, f J 1 Vie FSAR calculations do not include the effects of release of stored heat from -' die reactor vessel and internals, and from the coolant pipes and steam genera-tors to the reactor coolant. Since the main differences in computed and recor-ded - transien ts (see. Figure 2) are in time scale and temperature ' transients, i t, , seemed appropriate to investigate the effect of stored heat- releaseon the above analyses. Four cases were chosen corresponding' to heat capacities of 13-3 __ _ _ -. _ __ J
Addendum No. 3 to Incident Report l l i 200,000 Btu F and 77,000 Btu F , each associated with time constants of 5 minutes, and of 20 minutes. The lower heat capacity corresponds approxi-mately to the reactor vessel and the higher heat capacity to the entire Reactor Coolant System metalwork. The results are shown on Figure 4 where system Tavg and pressurizer pressure are compared. Note that the transients again include early operator action in tripping the reactor coolant pumps and establishing maximum charging flow. Although it is apparent that manipulation of stored heat sources and time constants would eventually yield a close approximation to the recorded T avg it is significant that the minimum pressure is now con-siderably higher than the transients of Figure 2. The results are still con-servative relative to the actual plant transient since, in the context of steam break analyses, a high reactor pressure is pessimistic as it delays or prevents the initiation of safety injection. Although the pressure " hang-up" is more severe when stored heat release is con- *' sidered in the analyses, the reactor coolant pump coast down and full charging flow are still absolute requirements to attain a minimum transient pressure which is higher than the low pressurizer pressure set point for safety injection. This is shown in Figure 5 which is a re-run of the Figure 4 analyses, with the exceptions that only one charging pump is in continuous operation and the reactor coolant pumps are not tripped. For all four cases the pressurizer pressure falls rapidly through the low pressure set point of 1715 psi subsequent to einptying of the pressurizer and surge line. 3-4
Since the events show that operator action can preempt the normal performance ! of the safety injection system, it was considered c.ecessary to determine what would have happened if the accident had occurred at end of core life. A cal-culation was made to determine the reactivity effects based on minimum shut down margins and- assuming that (1) the measured core temperature and pressures occurred and (2) the charging flow was unborated to maximize the nuclear effect . The minimum core shut down margins occur at EOL Cycle 3 and are 1.77% ak with the worst rod stuck out of the core atd 2.92% Ak with all rods in. Figure 6 shows the variation of Keff with core average tempera-ture for the case of one stuck rod and the plant pressure being controlled at 2250 psia. A return to criticality is attained at core average temperature of 493 F in this case. An estimate of the corresponding case with all rods in, indicates that a further cooldown of 41 F to 452 F would be required to reach
~
criticality in the latter case. Although it is clear from an examination of the actual plant data, as shown in Figure 2, that a return tc power would have l occurred in both cases if boration had not been accouplished, calculations show I that the core average powers would have been only 4% cad 5.6% of nominal (2200MWt) respectively, for the no-stuck rod and stuck rod cases. This is due to the - very low cora flows which exist at the lower system temperatures as a result of . reactor coolant pump coast down. Any return to power in the presence of a low core flow greatly increases the core average temperature, which in turn, attenu-ates the return to power. Therefore, it is concluded that little or no damage to a core would occur if this particular transient takes place with minimum reactivity shut down margins and -the operator fails to borate'in accordance with the applicable procedure. 3-5
Automatic plant protection by actuation of the safety injection system in the
- event of the break size which occurred is dependent upon the substantial reduc-tion in reactor coolant pressure, which the transient analysis indicates would exist assuming there is no large increase of charging flow. Alternatively, how-ever, protection can be attained by the operator adding concentrated boric acid with the charging system (via the boric acid transfer pump) which negates the need for safety injection.
- 2. Effects of Transient on Plant Equipment The primary pressure boundary of the reactor vessel, steam generator, reactor coolant pump, reactor coolant piping, and pressurizer was exposed to a rapid temperature transient and a pressure transient due to the blowdown of the secondary side of the No. 3 steam generator. caused by the failure of the safety valve on the No. 3 steam line. The specific transient curves are shown in Figures 1 through 6 of Addendum No. 3. In addition, the steam side of the steam generator was boiled dry. The reactor internals were subjected to a thermal gradient due to the temperature transient.
The above conditions were reviewed by the equipment designer and Westinghouse and found to be acceptable for continued operation for the design life of the plant. A brief discussion of the evaluation for'the primary pressure boundary, the steam generator and reactor internals transient follow. Of the pressure and temperature transients experienced by the reactor coolant i equipment, the temperature transient is of more importance. The pressure transient with its rapid decrease from 2235 psig to 1860 psig a'nd slow increase back to 2035 psig actually reduced the primary stresses during the transient 3-6
(the primary stresses are propertional to pressure). The one cycle pressure ; transient is well within the design pressure of all of the reactor coolant system and the rate at which it occurred did not " hammer" the equipment, therefore, the general membrane stress was below the design stress. The local membrane and bending loads on nozzles are below those resulting from normal startup and cooldown since the equipment was subjected to a partial cooldown of 200*F in lieu of the 460*F normal heatup or cooldown range. The rate at which this thermal transient applied the load to the , nozzles is not a factor in the evaluation of local membrane and bending stresses. A secondary and peak stresses caused by the thermal transient need not be evaluated by the rules of Section III of the ASME Code for the emergency 4 I condition. the Thissmallpipebreakwasanemergencyconditionasd{finedb (S ~ m (9'YAd ' ) ASME Code Section($)w Paragraph N412 t(3)41n that it has a low probability of failure (less than 25 occurrences) and required shutdown of the plant to repair the failed pipe break. Since this accident is primarily a thermal ( t transient (which occurred for only one-half cycle) the Code recognizes that { a limited number of cycles (25) will have negligible effect on the fatigue usage factor of the equipment. l In addition to the basic reactor coolant equipment pressure boundary discussion above, Westinghouse performed a more detailed analysis on the steam generators since they not only' vere subjected to the reactor coolant transient but. Unit No. 3 was subjected to a complete blowdown without feedwater addition. 3-7
l l The major area of concern is the differential pressure across the tubesheet. ~ Westinghouse Tampa Division has a calculated fatigue usage factor for a i l single application of 2500 psi differential across the tube sheet as follows: f i ] Channel / Primary Shell Juncture .002 Tubesheet/ Secondary Shell Juncture .00008 { Tubesheet Center .00009 The maximum differential across the tubesheet of 2035 psig occurred af ter the blowdown was completed and the reactor coolant pressure had been raised to 2035 psig, therefore, the fatigue usage factor increase will be less than j indicated above. Hence, the CPL accident resulted in a negligible fatigue 1
; life usage in the tubesheet complex.
The reactor internals, while not a part of the reactor coolant pressure 1 4 boundary, have been reviewed for the effect of the transient on core structure components. A transient heat transfer calculation was perforced to ascertain. the relative motions, axial and radial, of the core barrel and thermal shield. The calculations indicated that the effects of joining members were significantly less severe than the acceptable design condition. Other pertinent areas, such as barrel-baffle bolts, were reviewed for possible detrimental. effects, and > were found to be satisfactory. Subsequent to cooldown of the plant, the reactor vessel internals were removed from the reactor vessel and were visually examined. No evidence of damage due to this transient was found and the internals have been - L accepted for power operation. l l 3-8 m
, - ,~ -
- 3. Data Recorded During the Incident The following figures and tables, Figures 7, 8, 9,10, and 11and Tables 1 and 2, represent incident transient data as recorded by the plant recorders and computer and a person stationed at the temperature indicator on the incore thermocouple panel.
Figure 7, page 26 is a copy of the strip chart recording from plant recor-der TR-413. This recorder records the wide range temperature signals from the RTD's installed in wells in the hot leg of each loop. Channel TE-413 was inoperative at the time of the incident. Figure 8, page 27 is a copy of the strip chart recording from plant recorder TR-410. This recorder records the wide range temperature signals from the RTD's installed in wells in the cold leg of each loop. Channel TE-420 was inoperative at the time of the incident. Figure 9, page 28 is a copy of the strip chart recording from plant recor-der PR-444. This recorder records a narrow range pressure channel from the pressurizer and shows the primary system pressure transient and recovery sequence up to the point of plant depressurization in preparation of ,re-starting reactor coolant pumps. Figure 10, page '29 is a copy of the strip chart recording from plant recor-der LR-459 which reccrds pressurizer level, rigure 11,page '30 is a copy of the strip chart recording from plant recor-1 der LR-477 which records secondary side water level in each of the three steam generators. LT-497, ste.am generator No. 3 level, was inking lightly and can only be seen during the last portion of the transient. Table 1, page 11 is a table of reactor vessel thermocouple temperature data as recorded by hand and wrist watch from the temperature indicator on the in-core thermocouple panel by an individual stationed at the panel. 3-9
i I Table 2, page 14 is a copy of the computer printout of the plant tempera-ture record. These data points had not undergone final calibration and were reading low at the time of the incident. 3-10
TABLE 3-1 H U EoB3ITxi CTr:l." EI ' *. ; J C l'l AlW _URIT llo. ,' l RAW ICACTCR YESSF.L C00110III l'ATA FIsoM TC - 36E (G - 15) - 5'T 11 CIlZI;T OF h/P8/To F HrA!in/See From Strip Chart -- Sko - 15: 20 495 - 15: 25 490 - 15!26 485 - 15: 26:30 480 - 15: 26:45 475 - 15: 26:55 470 - 15: 27: 20 465 - 15: 27:45 460 - 15: 28:05 455 - 15:28:30 450 - 15:29:15 h45 - 15:30:15 440 - 15:31:10 435 - 15:32 430 - 15:33 425 - 15:33:45 420 - 15:34:40 415 - 15:35:40 410 - 15:36:30 405 - 15:37:30 400 - 15:38:30 Changed Scale --- 390 - 15:39:10 PRESSURIZER PEESSURE 383 - 15:40:05 380 - 15:41:15 --- 2000 psia - pressurizer press. 375 - 15:42:20 --- 2025 psia - pressurizer press. 370 - 15:43:30 365 - 15:45: 12 360 - 15:46:27 355 - 15:47:55 350 - 15:49:30 345 - 15:51:20 340 - 15:53: 12 335 - 15:55:50 330 - 15:58:45 325 - 16: 03:00 320 - 16: 10:05 --- 2100 psia - pressurizer press. 319 - 16: 23:00 318 - 16:34: 20 317 - 322 - 17:07:00 322 - 17:30 320 - 17:42 319 - 17:50 318 - 17:50 316 - 17:03 1 3-11
TABLE 3-1 (Cont'd.)
) 39 -
18: 18 313 - 18:36 310 - 18:50
'd 1
3-12
I TABIE 3-1 (Cont'd.) REACTOR VESSEL T/C RAW DATA y INCIIENT of h/28/70 T/C CROSS CHECK AT END OF TRANSIENT T/C (16: 15:00) Time data taken fl - 313 f3 - 319 f5 - 320 f7 - 319
#9 -
320 fil - 320
# 13 -
320
#15 -
318
# 17 -
319 PUMP START TRANSIENT # 3 RCP 18:58 Hrs. - Time Start _S Min. 337 - 0 341 - 01/2 344 -
'O3/4 347 -
1 349 - 2 352 - 21/2. 353 -
- 2 3A 354 -
3 e-355 - 31/2 356 - 4 3A' 356 .- 6 3-13~ >
. . . . . .. . . . - . . . . . . . ~ _ . - - _ _ _ _ _ _ _
TABLF 3-2 Computer Temperature Record April 28,1970 1519 502.9 514.5 :' O.0 X 510.1 lg 1520 '19 8 . 6 499.7 :: 0.0 X 489.4 l 1521 494.2 495.1 :' O.0 X 474.9 lg 1522 438.3 500.2 :~ 0.0 X 469.'t i 1523 485.7 496.2 :" 0.0 X '+56.8 i 1524 486.9 484.8 :: 0.0 X 433.0
} 1525 487.2 480.2 :" 0.0 X 420.6 1526 482.2 483.4 :: 0.0 X r421.1 ',y; Q ,
1527 474.7 471.3 :: 0.0 X '*15.4
, Q 1528 463.8 452.2 :: 0.0 X 400.8 j 1529 454.1 443.6 :: 0.0 X 39's.7 .j 1530 446.2 446.4 : 0.0 X 396.7 ~ !O 1531 441.1 442.0 :: 0.0 x 389.8 .
4 1532 '436.5 427.9 :' O.0 X 374.4 l l { 1533 432.1 423.1 :" 0.0 X 368.2 , c3 gu 1534 T0439A T0459A T0'*06A T04' IGA 1 '
,. 1534 427.4 418.2 : 0.0 X 364.0
- 0.0 X 368.4 i 1535 422.6 '# 2 3. 3 ~
g 1 536 417.6 418.7 :" 0.0 X 363.8 1537 413.2 404.8 :~ 0.0 X 351.0
! 1538 408.9 400.2 :: 0.0 X 346.7 7.( g 1539 404.2 405.2 :: 0.0 X 352.1 4 1540 400.2 401.2 :" 0.0 X 3'e8. 4 j 1541 395.9 388.0 :~ 0.0 X 336.2 1@ 15'i2 391.0 384.1 :: 0.0 X 333.2 1 1543 338.2 389.6 3: 0.0 X 339.6 ] 15844 384.5 385.7 :" O.0 X 337.6
.N1 9- 1545 381.2 374.1 :- 0.0 X 325.9 f 1546 378.6 371.9 :: 0.0 X 323.5 4 . 1547 376.2 377.7 :: 0.0 X 330.1 1 543 373.5 374.6 :: 0.0 X 328.4 j; O 15't 9 370.7 372.1 :: 0.0 X 326.1 q 1550 368.0 360.9 :' O.0 X 316.2 jg 1551 365.6 358.7 :: 0.0 X 314.4
, _ _ _ _ s_, - _
,, 1552 363.4 30 5.4 - 0.3 X 321.7 1553 361.1 30 5.1 ' O.0 X 320.6 O 1554 35".7 351.9 : 0.0 X 319.9 1555 356.3 350.1 ': 0.0 X 309.1 1556 3 5't . 8 357.9 : 0.0 X 316.2
,9 1557 353.2 356.9 " 0.0 X 315.6 1 553 351.7 346.6 3 0.0 X 336.1
< 1551 350.1 345.3 ': 3.0 X 305.6 g 1600 34U.1 353.2 :" 0.0 X 313.4 9 1601 3:.6.6 352.4 3 0.0 X 313.2
.d 1502 3r 5,3 341,3 .- 3,3 x 3a),)
.) g 1GJ3 144.4 343.9 ~ J.3 ;'. D J.9
.f 1634 34 1.3 3:. ) . ~/
3.3 7. 311.3 1c.:5 3>.1.2 ! 4 !. . .' '
..) .\ !!!.6 g lo u t, 341.1 1./.5 -
- 0. J . 311.3
. . , 1637 P 0.2 333.4 :" 0.0 X D3.7 c' l 1608 339.2 337.3 3 0.3 x 334.5 g 1609 333.6 345.7 :" 0.0 X 312.9 i
t 1610 337.3 3f5.1 3 0.0 X 313.'t 6 1611 337.2 336.0 3 0.0 X 395.0
'1612 T0439A T0459A T040GA T0446A f O 1612 336.9 335.6 :" 0.0 X 305.2 1613 337.1 3'i'* . 86 3 0.0 X 314.6 Y 1614 336.7 343.9 0.0 X 314.7 O 5 1615 336.3 334.7 0.0 X 306.7 1616 336.3 334.9 :" 0.0 X 307.2 l '1617 336.3 342.9 3 0.0 X 315.6
- O 1618 336.2 341.3 :" 0.0 X 315.8 1619 336.0 332.7 :" 0.0 X 307.1 1620 335.8 330.9 :" 0.0 X 310.2 0 1621 335.3 328.1 :' O.0 X 310.4
.l 1622 335.3 337.2 3 0.0 X 329.4 1623 335.4 339.4 3 0.0 X 326.3 5 O 1624 334.9 331.6 :' O.0 X 315.3 i 1625 334.7 332.3 :" 0.0 X 312.7 3 1626 33'6.3 340.9 :" 0.0 X 317.9
_Q 1627 33't . 3 340.9 *: 0.0 X 316.4
',4 1628 333.9 332.1 :" 0.0 X 309.6 3 1629 333.7 331.9 0.0 X 311.8 i 9 1630 333.6 340.4 3 0.0 X 319.1 1 1631 333.4 340.6 0.0 X 313.4
'l 1632 333.7 331.9 0.0 X 310.0
- $ 1633 ~ 333.9 332.3 3 0.0 X 310.0 Q 1634 333.7 340.7 0.0 X 316.7 t
1 .. , , , . - - - - , . . . . . . . . . . . . . . . . . . . . . . . . .
, 1633 333.4 3'io.4 - 0.] X .,;1G . 2 O '1636 332.9 330.7 ~ 0.0 X 306.5
, 1637 332.1 33 ') . 3 0.9 x 306.7 j 1638 332.1 331.0 ~ 0.0 X 339.2
) @ 1639 332.5 339.1 : 0.0 x 322.1 i 1 640 332.5 338.7
- 3.0 X 322.1 1 1641 331.6 330.1 : 0.0'X 32a.7 O 1642 331.4 330.5 :~ 0.3 $ 321.3 1643 331.4 339.6 :- 0.0 X 330.5 j 1644 331.6 340.0 3 0.0 X 330.1 1@ 1645 331.6 331.4 : 0.0 X 320.7 1 1646 331.7 331.7 - 0.0 X 320.6 1647 332.1 340.7 :- 0.0 X 329.6
.ly O 1648 332.5 3'e 1.1 - 3.0 X 329.0 j 1649 332.3 332.5 0.0 X 320.6 i 1650 T0439A TJ W A T04;f.A TO'eMA
}@
i 1650 1651 332.5 332.3 332.7 ': 0.0 341.6 :" ').0 X 320.7 329.0 l 1 1652 332.5 337.1 A .0 X 32'e.6 l ? O, 1653 332.3 ,33 7.1 :- 0.0 x 324.4 i i 1654 332.3' 337.2 : 0.0 X 324.8 l
- 1655 332.5 337.4 3 0.0 X 324.8 dO 1656 332.1 4 1657 332.5 337.4 337.8 :- $v .0 a.0 X 324.8 X 324.8 1 1653 332.3 337.3 3 0.0 X 325.0 lg 1659 332.3 337.0 x 0.0 X 325.0 3 1700 332.5 337.8 3 0.0 X 324.8
{ g 1701 332.7 338.0 x 0.0 X 325.2 3 1702 332.5 337.8 0.0 X 324.8
.! 1703 332.5 337.8 0.0 X 324.6 lg 1704 332.5 337.6 x 0.0 X 324.1 1 1705 332.3 337.8 0.0 X 324.2
]g
{ 1706 1707 1708 332.3 332.5 332.3 337.8 x 337.8 337.8 0.0 X 324.2 0.0 X 324.4 0.0 X 324.1 1709 332.5 338.0 x 0.0 X 323.7 g , 1710 332.3 337.8 0.0 X 323.9
.y 1711 332.1 337.6 0.0 X 323.5 1712 331.9 337.8 3 0.0 X 323.5 4 1713 331.9 337.6 : 0.0 X 323.3
.] g 1714 331.9 337.6 x 0.0 X 323.2
I 1 715 331.7 337.4 :: 0.0 X 323.0 IO 1716 331 7 337.24 :: 0.0 X 322.8
. 1 717 331.9 337.8 :: 0.0 X 322.6 1713 331.7 337.6
- 0.0 X 322.2 jg 1 719 331.7 337.4 :: 0.0 X 322.4 j 1720 331.7 337.6 :: 0.0 X 321.9
- 1721 331.9 337.6 :" 0.0 X 322.1
)g 1 722 331.9 331.9 337.8 ::
337.8 :: 0.0 X 322.2 0.0 X 321.9
]4 1723 1 724 331.7 337.6 :: 0.0 X~321.9
{ g 1 725 331.7 337.6 :' O.0 X 321.5 j 1726 331.9 337.8 :: 0.0 X 321.9 i '1727 331.7 337.6 :: 0.0 X 321.5 0 1728 T0439A T0459A T0406A T044GA i 1 728 331.7 337.4 :" 0.0 X 321.9 4 1729 331.7 337.4 :: 0.0 X 321.5 g 1 730 331.4 337.4 :" 0.0 X 320.9 1731 331.6 337.4 :' O.0 X 320.9 1732 331.4 337.2 :" 0.0 X 321.1 g49 1733 331.4 337.2 :: 0.0 X 320.7 i 1734 331.4 337.4 :: 0.0 X 320.4 j 1735 331.2 337.2 :" 0.0 X 320.4
-; g 1736 331.2 337.2 :" 0.0 X 320.1 1737 331.2 337.4 :" 0.0 X 319.9
, 1~738 331.2 -337.4 :: 0.0 X 319.9 (j g 1739 331.2 337.2 :" 0.0 X 318.6 1 1740 331.0 337.2 :~ 0.0 X 317.7 1741 331.0 341.8 :: 0.0 X 321.7 331.0 341. 8 .:: 0.0 X 321.5 g 1742 1743 331.0 333.1 :" 0.0 X 312.7 i
1744 331.0 332.9 :: 0.0 X 312.4 1745 331.0 341.4 :" 0.0 X 320.6 d g 1 746 330.9 341.4 :' 0,0 X 320.7
~ }.h] : 1747 330.5 333.1 :: 0.0 X 313.4 S 1748 330.3 333.1 :: 0.0 X 313.4 g !
1749 329.9 340.7 :" 0.0 X 321.3
-C 1 750 329.9 341.1 :: 0.0 X 321.1
?? l 1751 3'29.9 332.1:: 0.0 X 313.1 3 O ; 1752 32 9. 7 331.9 :~ 0.0 X 312.9
$ ; .1753 329.7 340.5 :: 0.0 X 320.9 1754 329.6' 340.4. :: 0.0 X 321.1 ,
yQ i
.c O 1 .1755 329.4- 331.6 :: 0.0 X 312.4 0.0'X 312.7 @, ; 1756 3,29.4- 331.2 ::
9 1757' 329.4 .331.2 :s 0.0 X 312.6' O _ f 1758 329.4 339.6 :". 0.0 X 321.7,
, Iz . .s ..' xi.. *. . ; . . .'
ILJJ 329. 4 533.7 0.J . 312.9 TO 1001 329.0 330.9 - 0.0 X 313.4
.-f 1802 328.8 339.4 - 0.0 X 321.9 1803 323.4 339.2 ' O.0 X 320.9
; O 1 004 323.ft 339.5 : 0.0 X 312.2 .
1805 320.6 339.5 - 0.9 X 312.4 1806 T0439A T0459A T0f*06A TO'i'rGA l O 1806 323.3 339.2 ~ 0.0 X 320.9 11 1 807 327.9 338.6 :: 0.0 X 320.9 4 1803 327.6 329.2 - 0.0 X 312.2 0 1809 327.2 329.2 ~ 0.0 X 312.0 J 1810 327.0 337.6 ' O.0 X 321.6
'j 1811 326.0 33 7. r, .* 0.0 X 320.4 O 1 312 326.6 328.6 :: 0.0 X 311.8 j '
1813 326.2 323.1 : 0.0 X 311.4 1814 326.2 328.1 :: 0.0 X 311.2 y 9 1815 326.4 336.9 :' O.0 X 319.5 1 y 1816 326.4 336.9 - 0.0 X 319.1 s 1817 326.2 32E.1 : 0.0 X 310.5 E O; 1013 326.2 323.1 0.0 X 310.7 336.5 ' 1819 325.9 O.0 X 313.r, J 1320 325.7 336.3 :' O.0 X 313.9 . d O 1821 325.7 327.7 0.0 X 310.5 l i 1 822 325.7 327.7 :- 0.0 X 310.3 d IS23 325.6 336.2 :- 0.0 X 313.9
.i O 1 824 325.6 336.0 *: 0.0 X 310.4 ) 1325 325.7 327.2 M 0.0 X 309.6 i 1826 325.6 327.0 ~ 0.0 X 309.6
^ O , 1827 325.6 335.6 :: 0.0 X 317.5 1828 325.4 335.6 :" 0.0 X 316.9
..t 1829 325.2 335.4 x 0.0 X 316.6
?! O 1830 325.0 327.0 :" O.0 X 307.'e 1 i 1831 325.0 326.6 9.0 X 307.4 1 1832 324.8 335.4 : '..;.1 X 3:5.6
' O 1833 32'i.3 335.2 3 0.0 X 315.4 J 1834 324.6 326.'s : 0.0 X 397.6 Ji f835 324.3 326.4 : 0.0 X 307.6 i O 183G 324.C 335.1 M . 1.0 X 316.6
- 7. 1837 324.6 334.9 - ' .0 X 316.7
1,3 , 32's.4 320.1 d.0 ' 320.3 jl 'O 1 C33 32 !.. ! 326.1 :: 0.0 . 307.9 3 1U 0 32 3. ') 334.7 - 0.0 X 317.1 j 1 ;t,1 3:3.) 334.7 :: 0.a X 319.1
- j O 1@2 32;.9 325.9 :~ 0.0 X 310.5 10:.3 323.9 326.1 ~ 0.0 X 309.3 1044 TU439A 70459A T0406A T0f t'e6A 3
1044 323.7 334.7 :" 0.0 X 310.4 j 1L45 .32 3. 9 334.5 - d.0 X 310.4 J g 1846 323.7 334.3 :: 0.0 X 317.9 j 1847 1348 323.7 323.7 325.6 :- 325.6 : 0.0 X 309.2 0.0 X 309.1 334.1 - 0.0 X 317.1 j} g 1849 1050 323.5 323.5 334.1 : 0.4 X 317.7 j 1051 323.3 325.4 - 0.J X 308.1 1852 323.3 325.2
- 0.0 X 300.5 dO 1953 323.3 333.7 - 0.0 X 316.7 0.0 X 316.6
.3 1054 323.2 333.7 :-
324.4 - 0.0 X 307.2 j,1 g 1355 1856 323.0 322.8 324.2 '- 0.0 X 336.9 1 u 1057 322.0 333.4 :- 0.0 X 315.3 0.0 X 315.3
.}
4 d 1350 1859 322.6 32 2. 't 333.2 :: 32 4. 't :: 0.0 X 306.9 1900 371.3 334.5 :: 0.0 X 328.6
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