ML19296B778
| ML19296B778 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/19/1980 |
| From: | Gonyeau J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML19296B772 | List: |
| References | |
| LER-79-027-01T-01, LER-79-27-1T-1, NUDOCS 8002210493 | |
| Download: ML19296B778 (67) | |
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PRAIRIEISLAND NUCLEAR GENERATING PLANT Red Wing, Minnesota ts a UNITS 1 AND 2 %3 I I MINN E APOLl1 e ST. PAUL
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PRAIRIE ISLAND UNIT 1 STEAM GENERATOR TUBE BREAK LICENSEE EVINT REPORT UPDATE FEBRUARY 19, 1980 NORTHERN STATES' POWER CdMPANY . MINNEAPOLIS MINNESOTA
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8002210 $.
TABLE OF CONTENTS . Page Table of Contents i List of Figures 11 List of Tables 111 - A. Introduction 1 B. Regulatory and Offsite Agency Notification 1 C. Operational Sequence of Events 2 D. Operational Parameter Behavior - Short Term 4 E. Operational Parameter Behavior - Iong Term 4 F. Natural Circulation Cooldown/ Training 5 G. RCS Leak Rate and Minimum Water Inventory Calculation 5 = H. Radioactive Releases 8 I. Offsite TLD Measurement 10 J. RCS Iodine Behavior 11 K. Steam Generator Inspection / Corrective Actions 11 L. Recovery 14 M. Planned Actions 14 m N. Ccrnparison of FSAR and Actual Break Results 14 O. Recmmendations 16 P. SunTnary 17
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LIST OF FIGURES Figure Title Page 1 Pressurizer Ievel vs. Time 18 2 Pressurizer Pressure 19 3 Nuclear Power 20 4 Tave 21 5 Tc 22 6 AT 23 7 Steam Pressure 24 8 Pressurizer Level and Level Setpoint 25 9 Pressurizer Pressure 25 10 BCS Icop Pressure 26 11 Cold Leg 'Ibnperature 26 12 Hot Ieg Tenperature 27 13 Wide Range Steam Generator Level 28 14 Narrow Range Steam Generator Ievel 28 15 Steam Flow /Feedwater Flow - 11 SG 29 16 Steam Flow /Feedwater Flow - 12 SG 29 17 Turbine Steam Header Pressure 30 131 18 1 Activity Behavior as a Function of Time 31 19 11 SG Outlet Eddy Current Inspection Program 32 20 11 SG Inlet Eddy Current Inspection Program 33 21 12 SG Inlet Eddy Current Inspection Program 34
- 22. Sketch of Icaking and Adjacent Tubes 35 11
Figure Title Page 23 Overhead View - Failed Tube Area 36 24 Side View - Failed Tube Area 37 25 Front View - Failed Tube Area 38 LIST OF TABLES
- 1. Pressurizer Ievel Time Behavior 7
- 2. Gaseous Releases via Air Ejector 8
- 3. Gaseous / Iodine Releases via 'ID Aux. Feed Punp 9
- 4. Gaseous / Iodine Releases via Steam Dump 9
- 5. Liquid Releases 10
- 6. TID Gamma Radiation Exposure 10
- 7. Unit 1 Steam Generator History 13
- 8. Octnparison of FSAR and Prairie Island Break Results 14
- 9. Canparison of FSAR and Actual RCS Activities 15 APPENDIX I update report on release calculation and paraneter behavior lii
A. INTBODUCTION On October 2,1979, a tube break occurred in #11 steam generator of the Prairie Island Nuclear Generating Plant, Unit 1. % is event is being reported in accordance with Technical Specification Section 6.7.B.1 (c) in that an abnormal degradation of the reactor coolant pressure boundary occurred. W is report describes the event and subsequent analysis in detail since the break, due to mechanical wear, is expected to be of interest to the nuclear industry. It should be roted that all engineered safety systems functioned as designed and the plant operating staff acomplished safe reactor shutdown, steam generator isolation, and RCS cooldown in an expeditious manner following existing operating procedures. W is report addresses the following topics:
- 1. Regulatory and Offsite Agency Notification
- 2. Operational Sequence of Events
- 3. RG Parameter Behavior - Short Term
- 4. RCS Parameter Behavior - Iong Term
- 5. Natural Circulation Cooldown
- 6. RG Isak Rate Determination
- 7. Radioactive Releases
- 8. Offsite TID Measurements
- 9. RCS Iodine-131 Activity Behavior
- 10. Steam Generator Inspection and Cause Identification
- 11. Corrective Actions
- 12. Cmparison of FSAR analysis and actual break results
- 13. Recmmendations Additional information is provided in Appendix I. W is update report corrects typographical errors in the original report.
B. REGUIATORY AND OFFSITE AGENCY FDTIFICATION Notifications associated with this event were prmpt. We anergency Director declared a site emergency at 1430. Offsite notifications are not required for a site emergency; however, the Dnergency Director, being cognizant of the FSAR analysis, deemed the event significant enough to alert the offsite agencies of the potential for possible offsite consequences. All offsite contacts required for a general emergency were alerted. W e following are several of the offsite agencies contacted: (a) Minnesota Department of Dnergency Services (1432) (b) NRC Dnergency Response Center, Bethesda, and the Region III I & E Director (1433) (c) NSP General Manager of Power Production (1445)
% e Governor of the State of Minnesota was notified of the event at 1432.
he NRC I & E Site Inspector heard the announcement of the reactor trip at 1424 and was in the control rom within one minute of the event. A Minnesota Department of Health Dnergency Response ham was dispatched and arrived at the site environs at approximately 1700. he Region III I & E Dnergency Response Team arrived on the site at approximately 1830.
'Ihe site emergency was terminated at 2200 on October 2.
C. OPERATIONAL SEQUENCE OF EVEtES We following is a chrcnological sequence of events that occurred fran the first indication of the steam generator tube break until the leakirs tube was identified. See Appendix I, Figures 1-23 for short term paraneter behavior with explanation of behavior. Date Time Event Oct 2 1414 High Radiation alarm on 1R15 (air ejector discharge gaseous radiation nonitor) This alann came in and cleared twice before staying in. 1420 Overtenperature 6 T 'Ibrbine Runback signal due to decreasing pressure. 2 actual runback occurred. Approximate time high leakage started. 1421 Iow Pressurizer pressure (< 2139.9 psig) 1421 (approx) Canmenced load reduction 1422 Iow pressurizer level (< 18.3%) 1423 Started second chargiry pump (#11) 1424 (approx) Started third chargiry pump (#13) 1424:09 Reactor trip for "Iow Pressurizer Pressure"
- (< 1900 psig) 1424:14 Safety injection occurred due to "Iow Pressurizer Pressure (< 1815 psig) 1124:33 Minimum RCS water inventory; RCS pressure begins increasire 1426 11 Reactor Coolant Pump stopped 1427 12 Reactor Coolant Pump stopped 1432:29 11 Steam Generator level increased above the "Io Io Ievel" setpoint (13%) on the narrow rarge after having gone offscale low after the trip (It is normal for SG Ievel to go offscale low on a trip; recovery in this case was much nore rapid than usual)
Date Time Event Oct 2 1438 SI Reset 1441 Icop A MSIV closed - 1456 Pressurizer Level returned on scale 1456 Stopped 12 SI pump 1456-57 Began depressurization of the RCS using the pressurizer PORV. (The valve was cycled 6 to 8 times to reduce pressure to required value) 1502 Pressurizer level reached the high level setpoint (> 55%) 1506 11 SI Pump stopped 1507 Pressurizer Relief Tank rupture disc relieved 1515 RG pressure at 910 psig (sane as 11 SG pressure) Leak apparently stopped 1550 Ccrnmenced normal cooldown Oct 3 0640 RHR placed in service to continue cooldown to cold shutdown 1300 BCS at cold shutdown Oct 6 1640 Ccinpleted draining of RCS Oct 7 0250 Identified leaking tube 'Ihere are several considerations that need to be made during a recovery fran an event as this. For the cooldown:
- 1. Since the leaking steam generator cannot be used for cooling, RG coolcbwn will tend to be slower. Also, there will be some delay in cooling the leaking steam generator to the point where entry is possible.
- 2. With a leak next to the tubesheet, there will be scme draining of steam generator water into the RCS. Adequate boron must be injected. In this case, a second boric acid tank was used to raise the boric acid concentration. In 11 steam generator, a sufficiently high boron con ntration was measured to ensure no problems with RCS dilution during the cDoldown.
- 3. Since large quantities of 12% boric acid were used for injection via the SI pumps to the cold legs, these lines must be flushed.
D. OPERATIONAL PARAMETER BEHAVIOR - Short Term he Prairie Island plant omputer has the " Post-trip review" option which accumulates plant key parameter data on a continuing basis and saves the data immediately preceding and after a reactor trip. In addition, sof tware has been developed which plots this data. Figures 1 through 7 show the behavior of the following parameters as a function of time: (1) Pressurizer Level (2) Pressurizer Pressure (3) Nuclear Power (4) Tave (5) Tc (6) Delta (6) T (7) Steam Pressure (secondary) It is worthwhile noting that pressurizer level went off scale before the safety injection occurred due to low pressurizer pressure at 1815 psig. Accident analyses in the past had assumed initiation of the SI signal due to the coincident low pressurizer level (5%) and low pressurizer pressure (1815 psig) signals. In May 1979, the SI actuation scheme had been modified, as a result of Wree Mile Island followup evaluations, to a 2-out-of-3 coincidence low pressurizer pressure actuation logic. With either schene, SI signal initiation would have had to wait until pressure had dropped below 1815 psig. %us, the safety analysis still bounds the plant response. Pressure durino the most severe portion of the transient was dr@ ping at a rate of about 100 psi / minute. Pressure recovery was rapid once the SI pumps started injecting. Se plots of nuclear power and delta T show the power reduction prior to the trip. %e increase in Tave would be expected to accmpany the load drop since the rods were at the withdrawal limit. Tc increased as expected after the load reduction and appeared to be levelling out at about 532F when the trip occurred. E. OPERATIONAL PARAMETER BEHAVIOR - Inng Term Figure 8 through 17 are recorder tracings showing the long term behavior of the following primary and secondary system parameters in order: (1) Pressurizer Ievel and Ievel Setpoint (0-100%) (2) Pressurizer Pressure (1700-2500 psig) (3) RG Ioop Pressure - Icops 11 and 12 (0-3000 psig) (4) Cold Leg Temperature - Iocps 11 and 12 (50-650F) (5) Hot Leg Temperature - Icops 11 and 12 (50-650F) (6) Wide Range Steam Generator Level - 11 and 12 ( 0-100%) (7) Narrow Range Steam Generator Level - 11 and 12 (0-100%) 6 (8) Steam Flow /Feedwater Flow - 11 Steam Generator (0-4.5 x 10 lb/hr) (9) Steam Flow /Feedwater Flow - 12 Steam Generator (0-4.5 x 10 lb/hr) (10) Turbine Steam Header Pressure (0-1500 psig) F. NA'IURAL CIRCULATION COOIIXEN/ TRAINING After the reactor trip and the safety injection, the reactor coolant pumps were tripped in accordance with I & E Bulletin 79-06C. Cooldown was acomplished using natural circulation until 2128 on 2 October when 12 Reactor Coolant Pump was restarted to allow cooldown using normal procedures. Natural circulation tests and. demonstrations have been conducted as part of the preoperational startup testing progran and requalification training exercises. In fact most operators have had an opportunity to cooldown the plant using natural circulation. Wus, the operators were aware of what to expect as parameter behavior. Cooldown rates on the order of 50F/ hour were obtained with a Delta T of approximately 30F. Cooldown was acomplished by directing steam dump to the condensor during the first phase of the cooldown. It should also be pointed out that the Prairie Island licensed operator requalification program normally includes simulator training in emergency proct ., and accident simulation. Steam generator tube rupture exercises, as well as a variety of other accident conditions, have typically been included in these training programs. G. RCS LEAK RATE AND MINIMUM PCS NATER INVEmDRY CAICUIATIONS Maximum Imak Rate his calculation is based on pressurizer level vs time from the post trip review data (Table 1 and Figure 1). Ieast squares fit of data was performed for those areas of the curve in which pressurizer level was decreasing uniformly, as follows: (1) 14h 20m 57s to 14h 21m 29s (32 sec.) Ieak rate = -0.163750 %/sec (2) 14h 22m 25s to 14h 22m 57s (32 sec.) Ieak rate = -0.168749 %/sec
.(Leak rate)y = -9.825 %/ min
@ 64.6 g/% for 70 water
= 634.7 gpn Corrected for water temperature:
Initial cond: 2235 psig vf = .0269 70 F vg = .01605 Corrected leak rate = 634.7 x .01605 = 378.7 gpn (Leak rate)2 = (-0.168749 %/sec) (60 sec) (64.6 g) (.01605) min % .0269
= 390.3 gpn Minimum Pressurizer Water Inventory Pressurizer level is calibrated so that 0% should correspond to the lower level tap on the pressurizer for hot pressurizer conditions. For a different set of conditions the lower level tap may correspond to a slightly positive or a slightly negative reading. Figure 1 shows that the lower level tap corresponded to a readiry of about -5%, the point at which pressurizer level indication became flat.
By extrapolating the data frm the least squares fit through -5%, one can obtain a conservative estimate of the time at which the bottm pressurizer level tap was uncovered: Time 0 = 14h 22m 25s Frm least squares fit: Pzr level = 9.54% - 0.16875 % (time in seconds) sec For Pzr level = -5%, Time = (-5% - 9.54%)
.16875 %/sec
= +14.54%
.16875 %/sec
= 86 sec Time Pzr level reached bottm level tap = 14h 23m 51s Water volume in pressurizer at bottom level tap = 527g x .01605 = 314 g
.0269 Frm Figure 2, at 14h 24m 33s pressurizer pressure was increasing steadily indicating that RCS water inventory was increasing.
'Ihis calculation assumes leak rate continued at maximum (390 gm) frm 14h 22m and 25s, when pressurizer level drop was steepest, until 14h 24m and 33s, when pressurizer pressure began increasing. 'Ihis assumption is conservative in two respects: (1) Charging pumps were manually started at 1423 and 1424 and SI pumps autmatically started at 1424 and 14 sec. (2) The leak driving force Ap was dropping steadily frm 1422 and 25s to 1424 and 33s. Drop in Pzr level below lower pressure tap = 390 gpm x 42 sec x min 60 sec
= 273 gallons Pzr inventory at low level tap = 314 gallons Net minimum pzr inventory = 314g - 273g = 41 gallons TABLE 1 Time Pressurizer Iefel 14h 20m 57s 19.7 14h 2Lm 05s 18.5 14h 21m 13s 17.1 14h 21m 21s 15.8 14h 2Lm 29s 14.5 14h 21m 37s 13.2 14h 2Lm 45s 12.1 14h 21m 53s 11.8 14h 22m Ols 12.1 14h 22m 09s 11.9 14h 22m 17s 11.1 14h 22m 25s 9.7 14h 22m 33s 8.0
. 14h 22m 41s 6.8 14h 23m 49s 5.5 4.2 14h 23m 57s 14h 21n 05s 3.2 14h 23m 13s 14h 23m 21s 1.2 14h 23m 29s 14h 23m 37s -0.7 14h 23m 45s -1.6 14h 23m 53s -2.6 14h 24m Ols -3.3 Trip 14h 2tn 17s -4.6 14h 24m 25s -4.6 14h 24m 33s -4.6 14h 24m 41s -4.7 14h 24m 49s -4.7 14h 24m 57s -4.8 14h 25m 05s -4.8 14h 25m 13s -4.8 14h 25m 21s -5.0 H. RADIOACTIVE REE ASES
%e plant health physics staff monitored the plant envirorrnent for offsite release and determined activity levels of raiioactive liquids and gases contained by plant systens or released to the envirorinent.
Airborne Activity Gas releases were made through the following paths: (1) Air ejector discharge (2) Turbina driven auxiliary feed pump exhaust (3) Atnos@eric steam dump (4) Gland steam exhaust he followirg is a summary of the releases made via those paths Periodic grab samples were taken on the air ejector exhaust in order to accurately assess the activity released via this path. We air ejector was used during the period 1414, 2 October 1979 through 0730, 3 October, 1979. Even though the 11 steam generator was isolated at 1441, the health physicists desned it prudent to nonitor the exhaust until the unit was secured. We followiry table summarizes the activity released durity the period. TABE 2 Gaseous Releases via the Air Ejector Nuclide Activity (Cil Xe l33 21.9 Xe l35 4.63 Ar 41 2.00 E-05 0 Kr 0.034 Kr 0.446 H 1.94 E-05
'Ibtal Activity 27.01 C i n e turbine driven auxiliary feed pump was operated for approximately 24 minutes (until after SI was reset). Based on the 1515 and 1715 steam generator activity data and assumirg full flow with the auxiliary feed pump (a conservative assumption), the following is a sunnary of the radioactive releases via this path:
TAB E 3 Gaseous and Iodine Releases via the TD Aux Feed Pump Nuclide Activity (uc) 131 I 3.01E+1 I 1.14E+1 133 Xe 3.07 E+5 l35 Xe 8.73 E+4 Kr 00* 2.46 E+4 Kr 0 2.20 E+4 Ar 41 3.92 E+4 One of the 2 atnospheric steam dump valves off loop A opened for 3 seconds after the reactor trip. This time is fairly typical based on previous operating experience. These valves normally take about 3 seconds to stroke full open. In order to determine the possible release via this path, it was assumed that full flow passed for 3 seconds ard activities were backcalculated frm the 1515 and 1715 samples (as for the TD auxiliary feed pump evaluation). The releases were as follows: TAB 2 4 Gaseous and Iodine Releases via the Steam Dump Nuclide Activity (pc) l31 I 3.4E0 l33 I 1.3E0 133 Xe 3.42 E+4 l35 Xe 9.72 E+3 0 Kr 2.74 E+3 0 Kr 2.45 E+3 41 Ar 4.37 E+3 For these release paths, iodine releases are based on 1515 data which are nore realistic. Gaseous releases are conservatively calculated using the 1715 data since more emplete data was available. Additional information is provided in Appendix I. No estimate of the release via the gland steam exhaust is available at this time. Release by this path is expected to be less than that via the air ejector path. Liquid Activity he air ejector condensor drains to the turbine building sump. Liquid samples were taken during the period that releases were made frm the turbine buildirg sump. We table telow sumarizes the releases made during the period 2 October 1979 through 4 October 1979. TABLE 5 Date Nuclide Activity Released (11c) 2 Oct 133 Xe 1.23 E+3 135 Xe 7.56 E+3 3 Oct 133 Xe 1.08 E+4 l35 Xe 2.51 E+3 4 Oct l33 Xe 1.13 E+3 Total Activity Released = 2.32 E+4 pc Steam generator blowdown was being directed to the condensor prior to the tube failure, thus this reduced one possible path for liquid release to the environment. We circulatirg water system was operating in the recycle mode where only 150 cfs may be returned to the Mississippi River. Eis tended to spread out the releases frm the turbine buildirg sump to the river over a longer period. %ere was no detectable iodine in the liquid sanples during this period. I. OFFSITE TLD MEASUREMENIS TLD's had been placed in four locations near the Prairie Island facilit.y m 2 October 1979 (see table below for exact details) as the fourth quarterly TLD's. Rese were replaced on 3 October 1979 because of the incident. Rese TLD's were shi; ped to and developed by Hazelton Environmental Sciences Corporation, of Northbrook, Illinois. Rese devices had a higher uncertainty because of higher in-transit exposure compared to those experienced in the field. Table 6 summarizes the TLD location, exposure period, duration and total exposure as reported by Hazelton Services. TABLE 6
'ILD Gamma Radiation Exposure Iocation Exposure Period Duration of Exposure Total Exposure (Times) (Hours) (mrem)
PI-1 10-2; 1100 to 28 0.18 + 0.42 Prescott 10-3; 1500 TLD Gamma Radiation Exposure Iocation Exposure Period Duration of Exposure 'Ibtal Exposure (Times) (Hours) (mrem) PI-2 10-2; 1030 to 29 0.33 1 0.40 Ellsworth 10-3; 1535 . PI-3 10-2; 1315 to 27 0.27 1 0.42 NN Sector 10-3; 1615 PI-4 10-2; 1330 to 27 0.28 1 0.39 Iock& Dam #3 10-3; 1620
%ese measurements show the low dose that could have been received and these should have been low due to the low level release that occurred.
J. RCS IODINE BEHAVIOR Over 90 RG I 1I smples were measured in the 51/2 days after the event because of the interest in the spiking phenmenon and the transfer of coolant frm the RCS to the secondary with subsequent possig release of activity to the environment. Figure 18 shows the I activity behavior during the period 1515, 2 October g 9 through 2200, 7 October 1979. Also indicated on the plot is the I level measured at 0730, 2 October, and shown as the pre-trip gvel. We Iodine -131 activity leveled out at approximately 8.5 x 10 pc/ml because the purification system ion exchangers had been valved out after the trip to prevent reduction in the RCS boron con ntration. K. STEAM GENERA'IOR INSPECTION / CORRECTIVE ACTIOG An inspection program of the steam generator tubes was ccriducted with the following objectives: (1) to identify the ruptured tube (2) to obtain information on the possible cause of the tube rupture (3) to detect any further tube degradation in the steam generator, and (4) to determine the general condition of the ste m generator tubing. he evaluation of the break and identification of the exact failed tube was acomplished by draining water frm the secondary side of the steam generator into the RCS through the opening of the leaking tube. Once the water on the secondary side stopped draining, the break elevation j was determined. %en by slowly adding water to the secondary side and visually inspecting the tube sheet frm the primary side through the manways, the specific tube was identified. Since the leaking tube (Row 4, Column 1, Inlet side) was located in the outer periphery of the tube bundle and the break was just above the tube sheet within the flow lane, foreign ob.)ect damage was suspected. %erefore, the eddy current inspection (shown in Figures 19 and 20 for 11 steam generator and Figure 21 for 12 steam generator) was concentrated in the outer periphery tubes. We inspection program was conducted in accordance with Technical Specifications, Section 4.12. %e results of the eddy current examinations revealed that the Row 3, Colum 1 tube (adjacent to the failed tube) had indication of a 65% reduction in wall thickness. We Fow 2, Column 1 tube (next tube) had a possible indication of <20% reduction in wall thickness. All of these indications were at approximately the same elevation. A complete visual examinatica of the outer peripheral area of the tube bundles and the flow lanes for both steam generators revealed no other signs of foreign objects. In addition, the eddy current examinations conducted for both ste m generators revealed no other tubes with wear or degradation. With the aid of mirrors and fiber optics, a visual examination of the three degraded tubes verified the eddy current results and revealed that the break res s bled a classical overpressure burst (running approximately 1 1/2" in the longitudinal direction of the tube with an opening width of approximately 1/2"). W e other two tubes (Row 3, Column 1 and Row 2, Column 1) showed signs of wear. All wear marks were on the outer peripheral side of the tubes. %ese patterns were documented by photographs and are depicted in Figure 22. During this time, a coil spring, later measured to be 8.5" long, 1.25" diameter, and 3/32 gauge was found to be lying on the tube sheet adjacent to the defective tubes. One end of the spring was lodged under the tube lane blocking device and the other end was free to move. Figure 23 is an overhead view showing the location of the spring, tube lane blocking device, and the peripheral tubes. Figure 24 is a side view, and Figure 25 is a front view. A definite wear pattern on the tube sheet indicated that the spring had noved during operation. Later, a second spring, identical to the first, was found on the mld leg side. It was located just opposite the first spring in a similar condition with one end lodged under the tube land blocking device and the other end extending out onto the tube sheet. In addition, part of an aviation hose clamp was found nc.xt to this spring on the cold leg side. A close visual examination ot the spring on the cold leg side, the tubes and the tube sheet surface revealed no signs of spring movement, tube damage or wear. %is was confirmed by eddy current examination of all eight tubes in Column 1 in close proximity to the second sprirq. Once these objects were removed frm the steam generator it was apparent that the spring and piece of hose clamp frm_ the cold leg side were encrusted with oxides indicating no active novement while the mils of the spring frm the hot leg side had definite signs of wear. %ese springs have been shipped to Westinghouse for further examination. %e springs and clamp appear to have been part of sludge lancing equipment used in one of the previous outages. Table 7 summarizes the work history on the Unit 1 steam generators. It appears that this spring was dropped into the steam generator prior to installation of the tube lane blocking device. Investigations in this area are continuing. Three courses of corrective action were identified based on the following facts: (1) the pattern of tube wear (2) the location of tube wear relative to the sprirg location (3) the wear pattern of the spring on the tube sheet (4) the signs of wear on the sprits wire - (5) no other tubes that were inspected other than the three previously discussed had any signs of wear or tube degradation (6) No other foreign objects were found. It has been concluded that the tube break was the result of reduction of wall thickness by the wearing action of the spring against the tube. The three courses of corrective actions identified were - (a) plug the two defective tubes and several adjacent tubes, (b) remove ; , % ion of the defective Row 4 tube and plug the two defective tubes, or (c) stabilize the defective Row 4 tube fran the primary side and plug the two defective tubes. Discussions anong NSP, Westinghouse, and NRC staff re-vealed that option (a) should be selected for several reasons - (1) in accordance with ALARA, the occupational exposure would be least, (2) unmrtainties in the design concept of tube stabilization, and (3) minimize effects of possible instability of the defective Bow 4 tube. In addition to the failed tube, and the tube with 65% wall reduction, the remaining four adjacent tubes to the failed tube were plugged. This action was taken to eliminate the remote possibility of the failed tube breaking further and damaging the adjacent tubes. A secondary side pressure test is being conducted to assure leak tightness of the plugs.
'IABLE 7 UNIT #1 STEAM GENERATOR HISTORY OUrAGE DATE NORK PERNRMED EDDY CURRENT SLUDGE LANCING 12/17/73 PERNRATED PIATES NWE NO 09/0d/74 ENE BOUI S.G.s BOTH S.G.s 04/25/75 NNE BOTH S.G.s BOTH S.G.s 03/04/76 EDIFIED S.G.ORI- BOTH S.G.s BOIH S.G.s FICE, DECKS, F.W.
RING & BIDWDOWN IANES 04/17/77 KNE BOUI S.G.s BOIH S.G.s 03/26/78 MNE S.G. #12 BOIH S.G.s Os/06/79 NWE NONE BOTH S.G.s L. RBDOVERY
.%e followirg schedule is planned for return to power dependity upon activity cleanup in the secondary system:
Date Event . Oct 17 Fill and vent RCS Oct 19 Heatup the RCS Oct 21 Reactor critical Oct 23 Unit at 50% power M. PIANNED ACTIONS Durirg the next refuelirg outage, the plant staff will inspect the area of the failed tube for ancrnalous behavior of the plugged tubes. N. COMPARISON OF FSAR RESULTS VS ACIUAL BPIAK RESULTS We Prairie Island FSAR Section 14.2.4 addresses a steam generator tube rupture. W e table below summarizes differences between the FSAR calculations and the October 2 break results: TABLE 8 Cagarison of FSAR and Actual Break Results Prairie Islam Item FSAR Tube Break teak rate, gpn ~ 616 ~390 Percent defective fuel, % 1.0 ~ 0.01 Previous leak rate prior 5.0 0.0 to break, gpn Ibs., stean transfer 120,000 ~5,000* 30 minutes R
$ 31 g activity released,iC equivalent 21,700 ~
30 (actual Xe133) I 209 ~ 37.0E-6 (dose equivalent)
- NOTE - Steam released fran operation of the turbine driven auxiliary feed-pump aM steam dump to atmosphere actuation. Actual water transferred was significantly nore.
'Ibe FSAR assumed specific radionuclide con ntrations in the RCS coolant as shown in Table 9 which cxxnpares selected radio nuclides to the actual concentrations in the RCS coolant as measured at 0730 on 2 October 1979. It shculd be noted that all actual concentrations were well below the postulated FSAR concentrations, thus assuring that the radioactivity levels released during the event wem bounded by the FSAR analysis.
TABLE 9 Canparison of FSAR and Actual RCS Activities (pc/ml) 1 Noble Gases FSAR Measured Activit'1 Kr85 y,yy ___ Kr 0 1.46 1.19 x 10-2 Kr 0 0.87 1.61 x 10-2 00 Kr 2.58 - l33 1,74 x 10 2 Xe 3.35 x 10 ~1 l3h Xe 1.97 5.63 x 10-3 l35 Xe 4.95 8.32 x 10-2 Xe l3h 0.14 138 Xe 0.36 Ar41 1.18 x 10-3 Corrosion Products / Activation Products 4.2 x 10-3 54 Mn 1.19 x 10-3 8.1 x 10-3 S8 Co 3.18 x 10~4 4 Na 3.83 x 10-3 Non Volatile Fission Products 131 I 1.55 2.15 x 10-2 133 I 2.55 1.08 x 10-2
-15
Comparison of FSAR and Actual RCS Activities (pc) ml 1 FSAR Measured Activity 7.0 x 10 -3 l I 1.4 l36 Ca 0.33 6.7 x 10 -4 l37 Cs 0.43 1.73 x 10-3 Cs l39 - 2.45 x 10-4 Tc99* - 1.40 x 10-3 99 Mo 2.11 8.47 x 10-4 88 Y - 1.78 x 10-3 W ___ I'ES
- 1. ESAR Appendix D4
- 2. Sauple taken 2 October 1979, 0730
- 0. RECOMMENDATIONS A review of the event has resulted in the following recmmendations and
- rxvents --
(1) A note of caution should be added to the steam generator tube rupture procedure to have the rperator stop the turbine-driven auxiliary feed pump as soon as ,ossible and to shut the steam supply valve frm the affected steam generator. This was done even though this caution had not been in the procedure. (2) Add a note of emphasis to the operator to isolate the leaking steam generator as soon as possible and to keep in mind that the MSIV bypass can be used to protect the steam generator frcm overpressure. Also, reducing the RCS pressure quickly, while maintaining adequate RCS subcooling, (whid was done during this event) will help prevent overpressure of the secondary side. (3) Consider operation of the reactor coolant pumps during a steam generator tube break. This would allow use of the 1 pray valves in the depressurization process, which could minimize the chance of blowing the pressurizer relief tank rupture disc. It should be pointed out that even though the disc failed, there was little discharge to the containment. (4) The industry should cansider the problems associated with a tube break recovery, in particular - (a) increasing pressure indicates level has probably reached its lowest point and that level is recovering (even if off-scale low), and (b) bringing pressure up in the pressurizer to greater than 2000 psig leads to increasing flow OUP of the break (thus slower recovery) and decreasing makeup flow fran the SI pumps (due to pump-head curve diaracteristics) (5) Evaluate the feasibility of not isolating instrument air to the containment, since the PORV's are used to reduce RCS pressure in this event. Under current logic and procedures, contairrnent isolation must be reset to repressurize the PORV accumulators. (6) Upgrade procedures for control of material into and out of the steam generator and other enclosed spaces. P.
SUMMARY
This 14-day repart has attempted to amnpletely address all areas of interest to the NRC. Additional reports will be subnitted, as appropriate. LO480A 24 + I FIGURE 1 I I I I Pressurizer Level I I 21 + I % I I I* I I
- I 18 + I I
- I I I I
- I 15 + I
% I
- I I
- I I I 12 + * **
- I I
- I I I e I
- I
% 9 + I e I
- I I I I
- I 6 + I I
- I i
- I I I 3 +
- I .
I I I
- I I I O + I I
- I I
- I I
- I
-3 +
- I I I I 12* ** * **
- I I * * ** ** * ** ** * ** *
-6 + I I + --- - - + - - - - - - - + = - + - - - - + - - - - - - + -
+---- - - + - - - - - - +
-200 -100 0 100 200
-l 50 -50 Time (sec) 50 150 HORIZ
PO480A 2380 + I FIGURE 2 I I I I Pressurizer Pressure I I 2300 + I - I I ' I I I I 2220 + I I I - 1 I I I 2140 + I I ** I PSIG' I ** I . I ** **
- I 2060 + ** I I
- I
~
I ** I - I ** I 1980 + ** I I ** I O I ** I y I
- I **
1900 + I *** I I **
- I I * * *
- I I ***
1820 + I2 **
- I I * **
- I I I I 1740 + I I I I I I I 1660 + I I I I I I I 1580 +* I I+ --
+ - - - + - - = - + - - - - - - + - - - - - +----- - - + - - = + ---+
-200 -100 O I00 200 T m,e (sec)
-150 -50 50 150 HORIZ
N0052A FIGURE 3 100 + ** ** * ** I I
- I Power Range I
- I I
- I 90 + ** ** * ** ** * ** ** I I I -
1 I - I I 80 + I I I
% I I .
I I 70 + I I I I I I I 60 + I I I I I I I 50 + I i I I
$ I I e I I 40 + I I I I I I I 30 + I I I
~
I I I I 20 + I I I I I I I 10 + I I I I I I I2*
- O + I * * ** ** * ** ** * ** ** * **
- I + -- -
-+------s----- +---- --+ - - - - + - -
+ - + - --+
~200 -100. 0 100 200
-150 -50 Time (sec) 50 150 HORIZ
T0401A FIGURE 4 564.0 + I y y 11 RC Loop T I I I I 561.5 + I I ** I I
- I I
- I 559.0 + *
- I I * ** * * * *** I I * ** I -
I I 556.5 + I I I y I I I I 554.0 + 1 I I
- I 12 U' I I 551.5 + I I I I I I I 549.0 + I I I I I I I 546.5 + I*
I I - I , I I I 544.0 + I I I
- I I I I
- 541.5 + I * **
I I
- I I ***
I I * ** * ** ** * **
- 539.0 + I I+ - - - - + - - + - - - - - + =-- - - - - - + - - - + - - - - +- - - + - - - - +
-200 -100 0 100 200
-150 -50 Time (sec) 50 150 HORIZ
T0406A 540.0 + I I FIGURE 5 I I I 11 RC Loop Cold Leg Temperature I I 538.5 + I I I I I *
- I I * * * * * **
537.0 + I **
- I I * *
- I I **
I I 535.5- + 1 I I ** F I I I I
- 534.0 + I I I I U I I*
' I
- I2 532.5 + I I
- I
- I * *
- I I *
- I 531.0 + *
- I I
- I I
- I I I 529.5 + * **
- I I * * ** I I
- I I I 528.0 + 1 I I I I I I 526.5 .+ I I I I I I I 525.0 + I I+------+-------+- -
+ - - - - - + - - - - = +-------+------+---- =-+
-200 -100 0 100 200
-150 -50 50 150 HORIZ Time (sec)
T0403A 103 + I FIGURE 6 I I I * * * *
- I 11 RC Loop Delta T I
- I 93 +
- I I
- I I * * ** * * ** ** I I I 83 + I I I
- I I I I 73 + I I I I I I I2 63 + I I I
, I I w I I T 53 + I I I I I I I 43 + I I I I I I I 33 + I ,
I I I I I I 23 + I* I I I I I I 13 + I
'I I * * **
- I I * * **
I I * ** *** 3 + I ** * **
- I+--------+--------+----- + - - - - + - - - - - - - + - - - - - - - + - - - - - +---- --+
-200 -100- 0 100 200
-150 -50 50 150 HORIZ Time (sec)
PO401A FIGURE 7 940 + I I I 11 Steam Generator I I y y Outlet Pressura
~
915 + I I I ** * ** I I* **
- I I ** **
890 + I * ** I I ** I I * ** I - I2
- 865 + I I I PSIG I I I I 840 + I I I I I I I 815 + 1 i I I
% I I s I I 790 + I I I I I I I 765 + 1 I I -
I
- I I * **
- I 740 + * * * ** ** I I I I I I
- I 715 + I I I I * * ** I I **
- I '
690 + I I+ -- + - - - - - + - - - - - + - - - - + - - - - - - + - - - - - + - - - - + - - - --+
-200 -100 0 1.00 200
-150 -50 Time (sec) 50 150 HORIZ
E+E'.E E.E R M.n rtr1 i s 2-11_ l.1C_"T. .': M =-1100. -._ .152.=-- __ =5 :-- . .- =.\ '=:\r = q=m
- ==. .3_:q_ =__;-f.\
3_ - .
._. 9. -... . ==
_ .. p7og33 g
-1=w;:\~ .__=
=_ - -)-} Mn=-it -80 =-2) ==&
=J=1.m_ w --< .__.2 T; Z. ee;;= .J= =
;C ':=
Pressurizer level
. :. v- _ _ . . - .: --.=. -.~
=:.===
- == --
.1 - :=; -
. 70 & level setpoint
.-~ ..m - - . . . a s. .- =:=-= = = _: ... =
J ..' . .-- '- 2 'Un -d= =1= EEh &~ s= tc _-. O y.a ~Q2{E M_.f_ -_-3 E -h' = T== -- 5$ ~"
== -
- = = = -- pw. MVe ll . - a= =-=
- = :-4 ..: .t
=_ _.=-* g _ . _ =r u_:L:_- = - - - -.-
-- -+-- -
- = _ - - 5 . _:-- -
- 5 -s :-? r -54 : u- - - g = == ==. : -
g ;_ . '.-- e- .[- '- ,';..- .
- .1=
Mh : . ' . .
- 2_:. .:q-- -
v
.i) r .. ra.s. _ c=.
- . 1 =e .=, a
. g.
l
%30' y
- -. _ T_.
.; : .. . ;= ,=-..
.. -- -= - - .
+:Mk 2:3l = =
== I:1.= -
Q
--l6~}MIRf.Q.g:Eli)=5=-f5:+=l=E _. - #-.--
7 M 1 --- :.6 M I- ':r h Q_, (^
~'
-- t=_ .___=
5 y=-rg
.f.
-~s-=m --
9
-_ ~~l l?E~f_.
~ 0.k i- -
'M[m m m a a po m d r'
#d n>n n n a n .n n n p 2500. -
4._._ i .-
'=- %.. . v ai_3... ._- -
- . _. +
y 2400:; - 4 _. __-r__-t - .. FIGURE 9
. L-_: _
_-4i Pr ... .;.J ~
'2300
'~
Pressurizer
~~
Pressure T~
- 2200
=._
~ ~
2100 te .. = . . . ==. _ _,nnn-.
--.- .= =
c-- -- J -= .m-p-i f .~
-. = -
Q- - 1
= ~ --
1900; c ., _
.=-
- = .
p f .~--'- = ." ~ W 5 " ~ ~ " '
-180E.
b.! 2 . _ _
=: _. _.j$ , .
~
[-- 1 TIM :
]s - - -+-------- .$~- ~..
I - sn . qi r et a e e m a m m n4 N h ' h } N I O U_ _ U U U O I _
- ~.- - ; y--
\i-M 4 .
i i > > >
\
l 1
. ..\ .
\
1
\
\
i
.t i . . - i i i i __ 1 5i 1 I - - -o k_nn d. --
FIGURE 10
.1 m\tt!.$5p*.}?~dS . . \ ~W ~--~ ,
I 1 l I
> _..._ =
2y RCS' Loop Pressure
--y r Lujpt-
. , ,mn
.- s t _-
- u .._: . ,
=_ ,y +
3 !-# -
; e ,.
l ._Jr- g L.,
.+ r.: ..
1 l , ., , nn - . j a r ' -
- --I p: _ - .r. --
1 - _ .t. w mn
- g. - -y w a.
I T L i m 1 R i' i - l---r-l l l-
-'! . ' ,ll l
,'l ,.' . .F-~ nri~ . .
l
-l - ^- l .: l l T '" l l -- %
W - hw-- .- ,l Ir, l 2. n . l l6 -
- = - l = ^-l '~l~ :: "l C l l&
,: - l
.l - l ~^
,l - l ".
,l ,
9 g w.1.v..wgg g l# # n~n'n nin n n n'n
- g%
S
\ \ n, \ '
50-i
>_ g , i,,---\\ - -. ,-\ '4
-A.___ ..
i > i s wy_ . _ i i
- . \-; ~\ \-- .9 _ -\.2,. . ;: =
y . . - . .. n. .
- y. .
~ rhlt$,8= hg..
i i g p_ ._ n'l FIGURE 11
~; * - '-
s
..- -...]f x. . .?'.&._, ,. ,- +. . . Ai' Z~%-.
^ - -
~2 .. _:~ :$A
&n4
*:'emperature -
p r . , . -
, x ,, ~
p --. a w v-p*.*w 11 & 12 Cold Legs
, W ' ** J SmO ArMp-
! bkl.55y '.'$&$$ $~
& .}'{lgh),[ ~h v.
~
._~...
~._~
. . sh
,.'& q m T se V ' .
t' CJ9 *$n j Qf
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3_ .# V V
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- c. l I l .--l..., ..
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g .. .- .
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. l l l . .-swl .
+ r, w n r r- - -- r - :
2A *N 055BB5W 4, - '., _ - - - '650-y
,, g
--.+ . --. . . . . . , .
g
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mac & 1 i i iy y a 1 I R
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, - ...,4 - .. .,
... d d Y- - 2
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x, . , m;. . .
-- 1 I nin" '*~-
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-+%- - = , y-p.- .- . .d vv p , .
m .m :_m - - -- .--- L J, _l %~
, , . . ML;,T r._
FIGURE 12
$ R,.'$i
_. w W%f a- . . . m' 42 ... . .
-~~~-
rl w g - - _x - g . , - M , Temperature -
, r 4w - - - ~ ~
n r,, -_
, ,-y,_ =.
_ .,,s __ --._ u v. 11 & 12 Hot Le9s
~ - ^ -
+ --
~
~~~ = -;p :. -
+ -
- - . vvv. . . _ . . _
gg ,
, o ya. __ . .
r '. ' , , __'P
?. .~.,,.'~^
,, ..~.-~.
n 1 I qAg -
-*l c ,z
' ^]! '{ y' ^
.,,u
,l^ '
- l & * *- . ..
4 ,- I L 7,ATi _y r g* -*--.m=-- I f I y y yM
^'
n" a m .h;i"_' . l - =*==- 42 * . - - - . z &'" ,il a =lS-..I %d~ ~~~~ 8 m- - m,1_ - . - - ,1, w: - - EAf *I i m i1
. . .Km m,g g y Q,
- - - - . - , 0, m'em?$g......
, ,s 9%
k 9N _x- - - . _ _ , _ . . . . _ , _ . . . . . . . . . _ . . _ . . . , , .-:
-[
I ..If J, 5 lR E E E M. Tl il N Y'w\T-WPW.%i=000
=--
\ ==-Age L:. ; \- =: \-
Q--: " Y._g: K 90 .= .... ,
- k= = m1-=.
'. =- FIGURE 13
--. =: { == i( _ -~ :. - i .: _
- L --{._ - . . i_ .. 1,8.0.} . _
_b__ f -Y. -. _. {-1 -- Steam Generator
* ""9 ~
f__.4IE'- , _ _ . 5 11 & 12 S/G's
~ -~
tr l .
=
.: 1
.1 =T: E= 1 lWBut
-3 l- .- :
j,
.Gb 3._ -? ~ ~
,A(0
_ ' I _S/f. ! .i.. ..._ . . 1 J-30 2
- . . _ .I ~
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j' < = = '
)_ + -l.-
]l= O.)-:
O .
~
l' Ol' - f ME -4:1+f/ 4/m sl=1 :n;czy .5~l[#n T__.. = =l = =,4 W = f = g +: :_.- , nrn -----,n r n ,g
'n i i f i e : a r a n n'
>--M 00 =noqino w =g= q- ~mo - a _
_ =g=g~- .
- 4 _ =_ n:
i
.i?$Mi .EEEata asa : FIGURE 14
-+:_]_ - i - --
2 .__. g l8 0q . - ~ 4 __ _ g i__._._ .=== 3 ___ ' =. :. Steam Generator 70 - _ , . _ - . _ _ _.__- 2 - - - -
== - - - Er
'dEE -:3 Narrow Range Level
~'~
~60
~
11 & 12 S/G's plI=~ G - l 250 r M =- == _i 7 N
- _ M ;- _
= =
__. ] sfizz'-=E =::? =E
.rg- -- _h VEE 5 gy --- t, y =. -=r - =-1 == ,
.< == p'^-e .
Q:m -r~"2b
=
g ;=i.3. p. , ?_*
- ' ygi Ilb $~ :=
F GM M 4= ==J =5a o/! $
#j f M =- ? ~-' - .
. . . . . w -- - y
1 tri n n n n n n n n n n w i - 4.s9 9,,.m Q&\ L _. .-- ). _ . . _. -- t
,4.0
; FIGURE 15
>\ -
1 . . . . _ . _ __ _' .JWE iM 11 Steam Generator inmi
~~~~
- _ Z;=.5 ~.:. - ---.
?l 3.0 Steam Flow &
Feedwater Flow 2.5 2.0
,.s p_. \
1.0 g a ._ r . I er g % n a-
-e-n- ~
1-e u . ___,[ -o , 9 ri - u.' PY' n m m ' re mm m ra gj nnnnronnnnnR
\ - ---\ _$5 u.:ssm
{ \ ...\'.. . \ ' 4.0 "'.\T' FIGURE 16 _l l--
~~
. ?? 12 Steam Generator Be-d .'
.Z Z. Steam Flow &
p,g _%. .. _..d, y ,_. 1 E. W T .. _ Feedwater Flow y.. 2.5
.t...
_ .1 . . . . . _ . . _
, .u . __2_.0 . . . . .
w _- . ....e _. . . . . 7
~~
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- l~ l ga] ' _ _ __
. a.t
~
- n. n n -------,c pti ,,M. . .gi,.--
..s ,
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~~ {i yi = _ ..K :. -
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~ - . . _ -
~
.~.~..
- _ _:- _ T = . . =
h - 2 --- - _ . _ _ _ rCE 5/[j FIGURE 17
~
~=
i -
~= 3i Turbine Steam Header Pressure
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$ en
~
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- m. +=. . s-e* -
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, , . , , ca , , _ , e : .c , .;--
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;- t
Illi APPENDIX I Update Report on Release Calculation and Parameter Behavior
Introduction This supplemental report provides additional information related to parameter behavior and revised dose equivalent I-131 calculations for the October 2, 1979 steam generator tube rupture at the Prairie Island Unit 1 facility. Operational Parameter Behavior Figures 1 through 23 illustrate the behavior of major key parameters during the 3 minutes before and 3 minutes af ter the reactor trip that occurred in conjunction with the steam generator tube rupture of October 2,1979. These plots are based on post trip review data accumulated by the PRODAC 250. Additional plots, some not included in the original report, are provided to illustrate behavior of parameters that can be used to determine which is the leaking steam generator earlier in the event. Additional notes have been included on the plots to explain transient behavior. It should be noted that although the 1 R15 radiation monitor alarmed initially at 1414, that the alarm cleared twice before staying in. Based on extrapolation from the short term data plots, it appears that the rapid change in RCS and secondary parameters started at about 4-1/2 minutes before the reactor trip (indicated by I or. the plots). For a leak rate of the magnitude experienc:d at Prairie Island, additional findings based on operational parameter behavior may be made:
- 1. Determination of the leaking steam ger.erator may be made by:
(a) Comparison of steam flow rate with feedwater flow rate for the af fected and unaf fected steam gene: ators before the reactor trip. (see Figures 20 through 23). Feedflow will drop off compared to steam flow for the affected steam generator. (b) Comparison of the wide range steam generator level behavior (Figures 13,15) for the af fected and unaf fected steam generators af ter the reactor trip. Wide range steam generator level increases in the affected steam generator after the reactor trip. (c) Observation of how rapidly narrow range steam generator level recovers af ter the reactor trip. The affected steam generator's narrow range level should return on scale first.
- 2. The blow down radiation monitor was not very ef fective in aiding determina-tion of whether a steam generator tube rupture had occurred. This monitor did not alarm until after containment isolation had occurred in conjunction with the safety injection. This delay may be attributed to the f act that the radioactive fluid must pass through the SGB flash tank before being detected (Figure 24).
- 3. Load reduct ions, a prudent action in the steam generator tube rupture event, can obscure trends in steam generator feedflow and level that could affect ability to determine the leaking steam generator before the reactor trip.
1-1
Dose Equivalent '.-131 Calculations 133 The October 16, 1979 report estimated I and 1 releases based on a 1715 steam generator sample. Releases had been estimated for the TD AFW pump and atmospheric steam dump flow paths. A steam generator sample, taken at 1515, is more representative of the steam generator activity during the time of the actual TD AFW pump and atmospheric steam dump releases. The liquid sample is expected to be representative since the samples are taken from a location close to the break. Calculation of dose equivalent I-131 was performed by correcting the sample to the release period and using more conservative iodine separa-tion factors, as used by the NRC. A WASH-1258 iodine separation factor of 100 was used in the recalculations (Reference 1). Realistically, a higher value could be used if the following factors are considered:
- 1. Moisture carryover decreases significantly as steam flow rates are reduced (e.g. carryover rates at 75% power can be as much as forty (40) times lower than carryover rates at full power). This effect would decrease the iodine released.
- 2. Safety injection actuation tends to reduce RCS temperature. This ef fect promotes better scrubbing of the iodine in liquid passing from primary to secondary by reducing the jetting effect proposed in some theoretical analyses.
- 3. The break was close to the tube sheet. This would promote better iodine scrubbing since the radioisotopes would have to pass through a larger water height.
- 4. Sodium and iodine carryover test data in Reference (2) and the similarity of sodium carryover test results for Prairie Island and the referenced 2 loop plant.
- 5. Volatile iodine transport. Though a small fraction of the I-131 is usually considered as volatile, this could be the largest cause of release at lower steam flow rates.
- 6. No iodine was detected in the condensor where a .15 f actor is recommended (Reference 1).
Table 1 summarizes the estbnabedreleases using the higher conservative iodine separation factor. 1-2
Table 1 Release Calculations Activity, " i TD AFW Pump 131 I 30.1 1 11.4 Atmospheric Stean Dump I31 I 33 3.4 I 1.3 Total dose equivalent I-131 36.9 G Activities ( i): I (1.5E-3) ; 1133 (5.7E-4) Releases (1bs): TD AW pump (4416) ASD (492) These calculations show that even on a very conservative basis the Technical Specification 3.9.3 design objective for atmospheric steam dump iodine releases (1.1E-3 Ci) was not exceeded. Dose Associargd With I-131 Using a X/Q value of 9.8E-4, the maximum dose would have been 1.9E-2 mrem attributable to I-131, using the methodology of Technical Specification, Section 3.4 Reference (1) NUREG-0017 Calculation of Releases of Radioactive Materials in Gaseous and Liquid Ef fluents for Pressurized Water Reactors (PWR-GALE code) April 1976 (2) WCAP - 8253 Source Term for Westinghouse Pressurized Water Reactors I-3
List of Figures Figure Title 1 Pressurizer Level 2 Pressurizer Pressure 3 11 Hot Leg Temperature 4 12 Hot Leg Temperature 5 11 Cold Leg Temperature 6 12 Cold Leg Temperature 7 11 Tavg 8 12 Tavg 9 11 Delta T 10 12 Delta T 11 Power Range 12 11 NR Steam Generator Level 13 11 WR Steam Generator Level 14 12 NR Steam Generator Level 15 12 WR Steam Generator Level 16 11 Steam Generator Outlet Pressure 17 12 Steam Generator Outlet Pressure 18 11 Steam Pressure to Turbine 19 12 Steam Pressure to Turbine 20 11 Feedwater Flow 21 11 Steam Flow 22 12 Feedwater Flow 23 12 Steam Flow 24 Steam Generator Blowdown I-4
ll l TIME OF TdIP 14:24 14
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-150 'io Time (sec) 50 150 HORIZ
TIME OF TRIP 14:24:14 10TA g 595.0 + I te* MaAltb I [ but h ofE1UtrieA N h b *** I 11 Hot Leg Temperature I 589.5 + I * * * *
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-150 -50 50 150 HORIZ Time (sec)
TIME OF TRIP 14:24: 14 THOTB 595.0 + M% @N W 'MO I *
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- I I ** I I * ** I I * ** I 584.0 + I -
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-200 -I00 0 100 200
-ISO -50 ,
50 150 HORIZ Time (sec)
TIME Ji: fR I r* $4s21sl4 f0426A 540.0 + I I I I I 12 Cold Leg Temperature I I ' 52.5 + I I I I I I I 537.0 + I I I I I PoSSitsy be6T I II M 535.5 I I I
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-150 -50 Time (sec) 50 150 HORIZ
TIME OF 191P 14: 24814 TO400A 564.0 + I 11 RC Loop T I I I pg g epgm ae TiAltb 56I.5 + w I REboc:pmea I I I
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556.5 + I I I F I I I I 554.0 + I I 12 5 o I I I I Ogg D REMC3tWL"TbF l s 551.5 + I 7 U I I I I 549.0 + I I I I I I I 546.5 + I* I I . I I I I 544.0 + I I I
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TIME OF TRIP I4:24:14 T0420A 564.0 + I I I I I 12 RC Loop T,yg 561.5 + ** DW D I I
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-200 -100 0 100 200
-150 -50 Time (see) 50 150 HORIZ
TIME OF TRIP 14: 24: 14 0403A 103 + I
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- I I * * ** * * ** ** I I I 83 + I I I I I I I 73 + I I I I I I I2 63 + I I I y I I Ow w mL MF 5 53 h I I ,
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- I 12 RC Loop Delta T I M NM 93 +
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- I I I 83 + I t I I I I I I 73 + I I I I I y I I
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-200 -100 0 100 200
-150 -50 Time (sec) 50 150 HORIZ
TIME OF TRIP 14:24: 14 10052A 100 + ** ** * ** I * [opgNunst. e*a mAM (w Rated I I Power Range I
- I I .* I 90 + ** ** * ** ** * ** ** I I I I I I I 80 + I I I t I I I I 70 + I I I I I m I I 5 60 + I &
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-200 -100 0 100 200
-150 -50 Tir.- (see' 50 150 HORIZ
TIME OF TRIP 14: 24: 14 LO400A 55 + I I TWM @@ I I I LQ3% M TO StaMr(.** I I 11 Steam Generator 50 +* ** * ** I Narrow Range Level I ** ** * ** I I I I ** a - gt:Jutmst D OC TO 45 + * * ** I 447 7I * * . I M M* I
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-200 -100 O I00 200
-150 -50 T m. (sec) 50 150 HORIZ
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-200 -100 0 100 200
-150 -50 Time (sec) 50 150 HORIZ
TIME OF TRIP 14:24*i4 LO423A 55 + I I I I M7, = Net.JW%. WOL- I 12 Steam Generator I I 50 + 'n D #
- I Narrow Range Level I MuN ** I I
- I I ** I 45 + ** ** I I ** ** I I * ** I I ** I g 40 + ** I I **
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-200 -100 0 100 200
-150 -50 Time (sec) 50 150 HORIZ
T I t.E Ur' Thle 148 4834 LO423A 60 + I I I 12 Steam Generator l { 59 + I Wide Range Level I I I I I I 58 + I I I I I I
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-ISO -50 Time (sec) 50 150 HORIZ
20400A , 950 + I I I 11 Steam Generator Outlet Pressure f , f 925 + I I I ** * ** I I ** * . I I* ** ** 900 + I ** I I * ** I I * ** I 12
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-200 -100 0 100 200
-150 -50 Time (sec) 50 ISO HORIZ
TIMci OF TRIP 14: 24 14 PO422A 950 + I I I 12 Steam Generator I I
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-200 -100 0 100 200
-150 -50 Time (sec) 50 ISO HORIZ
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- 390 + I* **
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+2
+II I +I II + III + II I+ III +II III+ III + III +III +I - 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 2 8 4 0 6 2 8 4 0 6 2 4 3 3 3 2 2 1 1 1 C A [e T
F0424A 4100 + I I I No%Sh @ (08ST fl IKcW 6) I [**** is maue w sceral= . 3700 I I *x I I * ** I 3300 + * ** ** g y * ** *
- y y ** I 12 SG Feedwater Flow I
- I (103 lb/ Im) 2900 + I I I I I I I 2500 + I I I ,,5 I I E I I $
M 2l00 + I y I I M y I I I I 1700 + I I I I I I I 1300 + I I I - I I I I 900 + I I I I I I I 500 + I I I I I2* ** I I
- 100 + I
- I+------ -+-- ---+ -== + - - - - - - - + - - -+-- +------ + - - - - - - - +
-230 -100 0 100 200
-150 -50 50 150 SECONDS Time
, F0425A 4100 + I , I I I I I I , 3700 + I I I I I , I ** **
- gg g I I
- OON# I y y 12 SG Steam Flow I 3 lb/ ng)
I I (10 2900 + * * ** ** * * * **
- I g I
- I I I y I I -
g 2500 + I y I I m I I g I I " i 2l00 + I I I I I 7 I I y 1700 + I I I I I I I 1300 + I I I I I I I 900 + I I I I I I I 500 + I I I I 12* I I ** * ** ** * ** * *
- 100 + I * * * * **
- I + --- = = + ---- --+ --- + - - - - - + - - - +-- + - - - - - - + - - - - - - - +
-200 -100 0 100 200 p -150 -50 50 i50 SECONDS Time ,
F0426A 3650 + I I I I I I I sS70 + I I I I I I I 3490 + I I I 12 SG Steam Flow I ** **
- I 3 lb/
I (10 Ita) I 3410 + I I I I I I I m 3333 + I E I I @ I I "' I " I 3250 + I 4 I I I *- On " g I 3170 I I I I I I I 3090 + I I I I I - 3010 I [g I I I I O
,, *' #l >I+
2930 +
- I y pr*g p 4 go pe*T I
- 2850 +
I
+----
-+--
- - - - + - - - -
--+-
I I
= +-------+
$N# $
E[
-------+ '
----+
I+ =
-200 -100 0 100 200
-150 -50 50 150 SECONDS Time
FIGURE 24-1 STEAM GENERATOR ELOWDOWN (Unit 1) (Simplified sketch) 11 Steam :;enerator
,- This sketch represents general flow paths and does not include all valves and instrumentation o
12 Steam Genera to r SGB Flash Tank To SGB Holdup Tanks (see igure 24-2) Y) Keylock [ SGB Heat Exchanger { { Valve
=
M C e To 1-R19 Discharge ( Radiation Canal { 1 ash Tank Pump Monitor Containment ; From Unit 2 l Steam Generator Blowdown Shield u lr' Building ; Auxiliary Building i ! t
From 1-R19 M FIGURE 24-2 (Figure STEAM GEhERATOR BLOWDOWN (continued) 24-1) X SGB Holdup Tanks O ac SGB Ion Exchanger s Aurf 11arv_Bu11d ng h Turbine Building l j~ ~ ~O 6 Unit 1 Condensor ) 44 _ SGB Monitor Tank AIyr Buildine Discharge Canal This sketch does not include all valves and instrumentation, but represents the general flow path}}