ML20151N769
| ML20151N769 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 12/19/1985 |
| From: | Bond G, Broughton T, Nicholas Trikouros GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20151N762 | List: |
| References | |
| TR-022, TR-022-R00, TR-22, TR-22-R, NUDOCS 8601030113 | |
| Download: ML20151N769 (38) | |
Text
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,1 i COMPARIS0N OF STEAM GENERATOR TUBE PLUGGING WITH THE TMI-1 DESIGN BASIS TR-022 (REV. 0)
Project No.: 5300-51724 L. C. Po Approvals:
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Tianager, Safety Analysis & Plant Control
/) {f W Date
' N. G. Trikouros
/7-s919r huclear Analysis & Fuels Director Date G. R. Bond a ,. ,,. ds f 2 . , i. ,T Director, Systems E.ngineering Date T. G. Broughton d h601030113 B51230 PDR ADOCK 05000289 P PDR s
TR 022 Rev. O Page 1 ABSTRACT This Topical Report provides a discussion on justification for plugging of up to 2000 tubes for the TMI-l steam generators.
An earlier Safety Evaluation had concluded that the margins to safety as defined in the Tecnnical Specifications will not be reduced oy operating the TMI-l steam generators with up tc 1500 tubes removec from se vice.
In December 1984. S&W was reauested to review the e=isting analyses to determine if they would support cceration with additional steam generatcr 1
tubes removed from service. Their response has concluded that a total of 3000 '
tubes with a plugging ratio of up to appro<imately 3:1; that is, as many as i 2250 in one, and as few as 750 in the other steam generator, can be removed from service with no impact on plant safety.
4 In order to ensure that the current Technical Specification limit on total reactor coolant flow remains bounding, the total number of plugged tubes is set at 2000 in this safety evaluation.
Both operational performance and effects on accident analyses are reviewed.
The results show no change in the unit's design basis.
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TR 022 Rev. O Page 2 TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
, ........................... S 2.0 OPERATIONAL PERFORMANCE ........... . . . . . . . . . . 7 2.1 RC Flow Rate and Margin to Minimum DNSR . . . . . . . . . . . 7 2.2 Asymmetric RC Locp Flow ....... . . . . . . . . . . . 9 2.3 RC Flow Coastdown Rate . .... . . . . . . . . . . 10 2.4 Steam Generator Water Inventory . . . . . . . . . . . . . 11
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2.5 Capability for Natural Circulation . . . . . . . . . . . 12 4
3.0 ACCIDENT AND TRANSIENT PERFORMANCE . . . . . . . . . . . . 13 3.1 LOCA. Analyses . . . ... . . . . . . . . . . . . . . 13
- a. SBLOCA Concerns . . . . . ................ 13
- b. LBLOCA ... ...................... 18 3.2 FSAR Analyses of Other Transients . . . . . . . . . . . . 20
- a. Uncompensated Operating Reactivity Changes . . . . . . . . 20
- b. Startup Accident /CRA Hithdrawal at Power . . . . . . . . 21
- c. Moderator Dilution Accident ................ 21
- d. Cold Hater Accident (Pump Start) . . . . . . . . . . . 21
- e. Loss of Coolant Flow ................ . . 22
- f. Stuck / Dropped Rod Event . ................ 22
- g. Loss of Electric Power .... . . . . . . . . . . . . 22
- h. Steam Line Failure .................... 22
- 1. Steam Generator Tube Failure ~ . . . . . . . . . . . . 23
- j. Fuel Handling Accident . ...... . . . . . . . . 23
- k. ' Rod Ejection Accident . .................. 24
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i TR 022 Rev. O Page 3 TABLE OF CONTENTS (continued)
PAGE
- 1. Manimum Hypothetical Accident . . . . . . . . . . . . . 24
- m. Haste Gas Tank Ruoture . . . . . . . . . . . . . . . 24
- n. Loss of Main Feedwater/ Fee: water Line Break . . . . . 24
- c. Steam Generatcr Overfill . . . . . 25
- p. Locked Rcter Incident . . . . 26
4.0 CONCLUSION
S .. . . . . . . . . . . . . 25
5.0 REFERENCES
. . . . . 29
TR 022 Rev. O Page 4 LIST OF FIGURES PAGE FIGURE 1 Reduction in RC Flow VS. Number of Tubes Plugged per
- Steam Generator . .................... 32 FIGURE 2 Comparison of Four Pump Coastdown Curves - FSAR VS. 1500 and 3000 Total Plugged Tubes . . . . . . . . . . 33 L FIGURE 3 One and Four Pump Coastdown Curves from TMI-l with Plugged SG Tubes . .. . . . . . . . . . . . . 34 i
FIGURE 4 TMI-1 Effect of Tube Plugging on Natural Circulation.
Total Core Flow VS. Time . . . . . . . . . . . . . . . . . 35 4
FIGURE 5 TMI-1 Effect of Tube Plugging on Natural Circulation.
, THOT Loop 1 VS. Time ............ . . . . . 36 FIGURE 6 TMI-l Locked Rotor Analysis Result . . . . . . . . . . . . 37 f
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TR 022 Rev. O Page 5
1.0 INTRODUCTION
Section 8 of TR-008, Rev. 3, " Assessment of TMI-l Plant Safety for Return to Service After Steam Generator Repair" (Reference 1), described the results of analyses performed to determine if the steam generators could be safely operated with up to 1500 tubes plugged. Those analyses had concluded that the margins to safety as defined in the Technical Specifications will not be reduced by operating the TMI-l steam generators with up to 1500 tubes removed from service.
In December 1984. B&W was requested to review the e<! sting analyses to determine if they would support operation with additional stean generator tubes removed from service. Their response, as documented in letter GPUN-84-252 dated December 17, 1984 (Reference 2), has concluded that a total of 3000 tubes with a plugging ratio of up to approximately 3:1; that is, as many as 2250 in the A steam generator, and the rest in the B steam generator can be removed fron service with no impact on plant operation.
Section 2 of this TR reviews the operational considerations associated with_ operation with tubes removed from service. This includes the reduction in total RC flow, the margin to departure from nucleate boiling, asymmetric flow distribution effects, the flow coastdown rate, steam generator mass inventory and the capability for natural circulation. In order to ensure that the present-Technical Specificaticn I limit on RC flows remains bounding, the evaluation is set at 2000 total
TR 022 Rev. O Page 6 plugged tubes. Section 3 reviews the effects of removing tubes from service on small and large break loss of coolant accidents as well as all
. other accidents and transients analyzed in the FSAR including a locked reactor coolant. pump rotor event. Section 4 is the conclusion of this report.
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TR 022 Rev. O Page 7 2.0 OPERATIONAL PERFORMANCE 2.1 RC Flow Rate and Margin to Minimum DNBR THI-l has a procedure for calculating the RC flow on a periodic basis. Reference 3 documented the flow calculations prior to the unit's restart and is summarized in Table 1. It was noted that
! there had been no systematic flow reduction over the previous four cycles and the minimum appeared at Test No. 7 where 109.57. of design was calculated. The maximum error on the calculaticn is 1.5%.
Therefore, the minimum total core flow prier to Cycle 5 is conservatively determined to be 1087..
The calculated RC flow rate for all four RC pumps operating as a function of an equal numcer of tubes plugged in each steam generator is shown on Figure 1. Generally, a reduction in tubes available for RC flow will cause the tube bundle pressure drop to increase. Since the remaining system pressure ' asses are about four times greater than the tube bundle pressure losses, only a slight reduction in total RC flow rate . vill result. The total RC core flow for 3000 plugged tubes will be tne same as the symmetric case, i.e. 1500 in
, each steam generator. From the figure, the reduction in total RC flow will be from 109.5'/. of design to 107.57., a change of 2%
(References 4 & 5). Further subtracting the calculational error, ,
the minimum RC flow after plugging 3000 tubes will be 106%.
In order to determine tne impact en the e<isting steady state Departure from Nucleate Boiling Ratio (DNSR) resulting from tne RCS 4
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TR 022 Rev. O Page 8 flow reduction at steady state, a study was performed to determine the minimum RCS flow rate required to maintain the existing DNB ratio for the TMI-l licensed power level of 2535 MWt. The existing DNB steady state ratio of 2.0123 was determined at a conservative power level of 2568 MWt and an RCS flow rate of 106.5% of design flow.
The methodolcgy for tne analysis was to calculate tne hot bundle flow by using a CHATA (Reference 6) core model which took heat balance input from the CIPP code (Reference 7). By using the hot bundle flow in t9e computer code TEMD (Reference 3), the Minimum DNBR (MDNBR) was caicJlated for the hot sub channel from the BAW-2 correlation.
T e results indicate that a DNBR value of 2.0123 can be maintained with an RCS flow rate of 104% of design and a power of 2535 MWt as compared with the original 106.5% flow and 2568 MWt.
According to Reference 9. the core ONB margin will be adequate with up to a total of 3000 steam generator tubes removed from service since 2% reduction of RC flow will result in a minimum of 106% of design, which is substantially above the 104% requirement to maintain the same DNBR.
Even tnough plant safety will nct be affected by plugging 3000 tubes, the current plant Technical Specifications are based upon a minimum flow of 106.5% of design flow at a power of 2568 MWt.
4 TR 022 Rev. O Page 9 Figure 1, adjusted for error, shows that for 2000 plugged tubes, the minimum RCS flow will be at or above 107% flow. Thus, this safety evaluation has been limited to 2000 plugged tubes to ensure the present Tech Spec limit of 106.5% flow is still bounding.
TMI-1 is currently (as of October 1985) critical and in the process of power escalation to full power. Total core ficw will be calculated according to the established procedures at various power levels. This will verify the core flow witn resDect to the existing plugging cattern and against the Technical Specification limit of 4
106.5% cf design. In addition, should any further plugging beyond the current total (1542) and/cr beyond the limit established in this evaluation (2000) become necessary, the established total core flow will then determine the eventual plugging limit and unit power level.
4 2.2 Asymmetric RC Loop Flow Distribution The analysis supports a plugging ratio of 3:1 of the tubes in the A l Steam generator to that in the B steam generator. In order to investigate the asymmetric effect of the RC loop flow rates, an evaluation with the plugging of 1500 more tubes in one of the 1MI 1 steam generators has been performed. The Loop A flow rate will be 4
approximately 2-1/2% smaller than Loop B. This mismatch between flows will.have no significant effect on the performance of the i
plant.
TR 022 Rev. O Page 10 During the last several cycles, Table I has shown that the A loop has typically about 3% more flow than the B loop. The result of more plugging in the "A" steam generator will thus be a somewhat more balanced loop flow distribution. The new flow difference is expected to be approximately 0.5%.
2.3 RC Flow Coastdown Rate With a significant number of tubes being plugged, the resistance factor for the RC flow passing through the OTSG will be increased.
This increased resistance may change the ficw districution if one RC pump is tripped while the other pump in the coolant loop is 9aintained in operation. The combined core flow during the coastdcwn may also be different.
Further, the minimum margin to DNSR during a loss of flow event is known to be dependent on pump coastdown rates. To address these issues, the following analyses were done.
A computer analysis has been conducted using the B&W code " PUMP" (Reference 10) for the TMI-l type reactor coolant system's flow coastdown curves with zero and 1500 tubes plugged in the A steam generator. The FSAR analyses served as the base case for the four pump coastdown transient. Results of the analyses with 1500 tubes plugged in one steam generator show that the FSAR coastdown rate is still bounding.
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TR 022 Rev. O Page 11 Figure 2 summrizes the data obtained from the four pump coastdown transient performed with the TMI-l version of the PUMP code and compares this data with the FSAR analyzed flow coastdown. This comparison shows that the flow with 1500 plugged tubes starts at a higher level than the flow assumed in the FSAR analysis and coasts down at approxistely the same rate; the minimum mrgin to DNB will not be changed.
A comparison of 3000 plugged tubes relative to 1500 was mde. The reactor coolant system flow as a function of time during 1 and 4 pump coastdowns is shown in Figure 3 (Reference 5), and the 4 pump coastdown curve is also superimposed into Figure 2. It was found that the 3000 plugged tube coastdown curves are generally 1% lower than the 1500 cases. There is still a substantial argin above the FSAR assumed coastdown curve. The cycle's flux-to-flow reactor trip setpoints remain the same (Reference 9).
2.4 Steam Generator Water Inventory The water inventory in the steam generator will increase by a sell amount due to the decrease in average quality in the plugged section. Total secondary side flow will increase only slightly with decreased steam outlet temperature. This would tend to cause a slight increase in pressure drop. This increase will be offset by the reduction in average quality (increase in density). The net effect (on OTSG level) should be a smil increase in the startup level.
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TR 022 l Rev. O !
Page 12 2.:i Cr: ,111ty for Natural Circulation The impact of steam generator tube plugging on the capability for a stable transition to natural circulation was examined by using the B&W computer program AUX (Reference 11). Symmetric plugging of 1500 tubes in each side, for a total of 3000 tubes, was assumed to be the bounding sse. Figures 4 and 5 compare the analysis results of 1500 tubes per steam generator to no tubes plugged. At about 500 seconds after the reactor coolant pumps trip, the stable natural circulation flow with 1500 tubes plugged is about 87. less than that with no tubes plugged. At no time is natural circulation lost and the
, reactor coolant system remains suocooled. The subcooling margin for 1500 tubes plugged per steam generator is only about 2*F less than the case without plugging (about 90*F). Therefore, these analyses have shown that natural circulation is still an effective method for decay heat removal.
Although the analysis was conducted for a symmetric case of 1500 plugged tubes in each side, a 3:1 distribution will not change the results significantly. Section 2.2 discussed the effects of asymmetry.
During the current restart tests of TMI-1, a low power natural circulation test has been conducted prior to power escalation.
Successful natural circulation patterns observed have verified the unit's capability for natural circulation.
TR 022 Rev. O Page 13 3.0 ACCIDEST AND TRANSIENT PERFORMANCE 3.1 LOCA Analyses The patentiel effccts of SG tube plugging on generic Large Break LOC?. (LBLOCA> and Small Break LOCA (SBLOCA) analyses (Reference 14) with 1500 tubes plugged has been examined. With about 67. Of the tubes in the A steam generator and about 27. tubes in tBe B steam generator being pluggec, the generic (2772 MWt) LOCA analyses for B&W 177FA Lowerec icop Plarts remain valic for TMI-1, with continued operation at core ccwer levels up to the licensed 2535 MWt at the existing LOCA limits. An overview of this examination is provided Delow. It is ther evaluated for 3000 tubes removec ^ rom service.
- a. SB LOCA Concerns The evaluation models used in the existing SBLOCA analyses (Reference 14) assume equilibrium conditions within the control volumes used to model the SG secondary 'n.
i For this reason, the localized cooling effects of EFW spray on particular tubes, and the effects of this cccling if particular tubes are de-activated, cannot be accurately predicted with tnese models.
In the application of the revised SBLOCA evaluation model (Reference 15), it was assumed that:
- 1. The percentage reducticn in the number of peripheral tubes removed from service will degrade the EFW spray cooling heat removal capability in a 1:1 relationship.
TR 022 Rev. O Page 14
- 2. The degradation in heat removal capability from EFW spray cooling translates directly to a reduction in depressurization rate by a 1:1 relationship.
In reality, these relationships are expected to be conservative because if EFW spray impacts a de-activated tube, it will not be t
heated and/cr flashed immediately but will either:
- 1. Be redirected onto adjacent active tubes, or:
- 2. F1cw inward into the tube bundle, providing cooling to active interior tuces, cr:
- 3. Fall into the saturated stean or saturated water region, resulting in increased cooling witnin these regions and/or an increase in the fill rate to tne accropriate level setpoint.
Therefore. the water is available in the steam generator and will result in greater EFW spray cooling and a higher depressurization rate than predicted'by the analyses. A more detailed discussion of EFW spray effectiveness and the TMI-1 response to SBLOCA with plugged tubes is given in Reference 12.
In this evaluation (Reference 4), two break cases were considered.
The first is the worst case with respect to peak clad temperature fo'r a small break LOCA, identified as accroximately a 0.07 ft cold leg break. The second belongs to tne category of breaks in which SG heat removal is needed to nelp depressurize the RCS. The 0.01 ft break was analyzed because this was the largest break size wnich would result in RCS repressurization.
TR 022 Rev. O Page 15 Flugging 1500 tubes was used as an upper bound which represents a de-activation of approximately 5% of TMI-l's total tubes. Also, because substantial tube plugging will be done in the peripheral SG tube's regions, it is estimated that about 18% of TMI-l's total peripheral tubes will be de-activated.
Worst Case 0.07 ft 2 Cold Leg Break with 0.ne HPI Train For this break size, the primary system cressure decreases below 1000 psi (approximately the secondary side pressure) at about 300 seconds. After tnis time. SG near removal is no longer possible, and the secondary side becomes a heat source for the primary system. Core uncovery begins at about 1350 seconds and ends at about 1750 seconds. The maximum time that SG (and EFW spray) cooling can be of benefit during the accident is tne first 300 seconds. This is very short when compared with the time to begin core uncovery.
The plugging of 1500 SG tubes will result in a reduction of the initial RCS liquid inventory by about 200 ft'. This results in the core being uncovered about 3 seconds earlier and in approximately a 10F increase in peak cladding temperature (to about 1100*F). This will have minimal impact on the outcome of this accident. It should also be noted that the generic analyses show 4
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that, for a 0.07 ft: break with 2 HPI trains, the core does nct uncover and temperature remains below 700*F. l i-1 i
TR 022 Rev. 0 0.01 f t 2Cold Leg Break Page 16 The 0.01 f t2 cold leg break case was evaluated using the revised SBLOCA model. All of the original analysis assumptions were preserved, including a 20 minute operator delay in initiating emergency f.eedwater. Two cases were analyzed considering the effects of tube plugging on the decrease in RCS depressurization rate which could increase the time to initiate ESFAS and the rate of heat transfer in the boiler condenser mode. It was found that:
With a 1600 psig low RCS pressure ESFAS setpoint, the 0.01 ft2 case will result in ESFAS actuation regardless of the reduced peripheral and interr.al SG heat removal caused by the plugging of 1500 SG tubes. Before ESFAS actuation, the RCS was subcooled and either forced or natural circulation existed. Consequently, SG heat removal was found to take place throughout the entire SG tube region, not predominantly in the peripheral regions. Therefore, the SG heat removal rate is not reduced by more than 5% during the period prior to ESFAS, and this will delay only slightly the activation of ESFAS.
2 The 0.01 f t break case will cause the RCS to enter the boiler-condenser (B-C) mode. In this mode, EFW spray cooling of the peripheral tubes is an important factor in the RCS depressur-iza tion. Thus, peripheral SG tube plugging could have a more significant effect on this cooling mode. The evaluation showed that, with an average of 18% of the peripheral tubes in the two steam generators plugged (peripheral is defined as 5 external rows),
sufficient steam generator EFW spray heat removal capability remains
TR 022 .
Rev. O Page 17 so that the rate of RCS depressurization is reduced by only about 12%. Even with this reduction, calculations with all other original analysis assumptions unchanged show a minimum of five feet of coolant remains above the core throughout the event. Since the 2
0.01 ft break is approximately the largest braak which would result in RCS repressurization, the plugging of 1500 SG tubes is expected to have only minimal effect on SBLOCA transients.
In summary, these small break LOCA analyses show that for the previously limiting case of 0.07 ft* cold leg break with only one HPI train available, peak clad temperature increased by only 10*F to about 1100*F. For the 0.01 ft: cold leg break, the slight delay in ESFAS actuation and the reduced area for EFW cooling have an insignificant effect on the outcome of the transient since a minimum of five feet of coolant remains above the core for both the plugged tube and unplugged tube cases. Therefore, the generic LOCA analyses remain valid for Tril-1 even with a reduction in SG heat removal caused by tube plugging.
Evaluation for 3000 Total Plugged Tubes As the total number of plugged tubes is increased to 3000, there is a further reduction of effective peripheral tubes in the number assumed in the previous analysis. The new analysis (Reference 2) has concluded that EFW spray cooling is still adequate for SBLOCA.
The argument is that if heat transfer is less effective, there will s
TR 022 Rev. O Page 18 probably be a continued-decrease in the RCS liquid level. This will increase the rate of condensation heat transfer by increasing the effective heat transfer length of the SG tubes. This may result in halting the RCS liquid level decrease by lowering RCS pressure, decreasing break flow and increasing HPI flow. If not, one of the following will probably occur:
- 1. Commencement of B-C when the RCS liquid level falls below the secondary level (pool B-C) or
- 2. Uncovering of, and direct steam venting througn, the break.
In References 16 and 17, it was shcwn that Doiler-condenser mcde cooling by a secondary liquid pool will be adequate if as few as 24733 tubes are active or more than 6000 tubes removed from service. This is true provided that an adequate secondary level (70% operate range plus instrument error is achieved and maintained, along with the HPI flows assumed.
Both B-C and break uncovering will result in RCS depressurization and increased HPI flow. This should occur well before, and should prevent core uncovering. On this basis, the assumed pattern of active peripheral SG tubes should be adequate to provide steam condensation heat transfer during a SBLOCA and ensure core cooling.
- b. LBLOCA The important parameters for the LBLOCA which are effected by plugging of tubes are the initial flow and flow coastdown. The 4
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o TR 022 Rev. O Page 19 effects of the reduction in coolant volume associated with 1500 plugged tubes (200 ft') are negligible fo- this event. The i*
plugging of 1500 SG tubes at TMI-l will reduce total system flow.
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However, the reduced flow (10% of design) will still be greater than i
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the flow rate used in the generic LBLOCA analyses (106.5%). This, i coupled with TMI-l's Icwer core power (2535 MW: vs. the generic I 2772 MWt) provices margin in initial conditions for TMI-1, relative to the generic analysis. During the early portion of a LBLOCA transient when the reactor coolant pumps are coasting down, the analy ec system f!cw rate (See 2.3) w'th additional resistance due l
to plugging will be greater than assumed design flow rate with the case of no plugging.
The reduction of 200 ft' of primary coolant volume will have little impact to the consequence of LBLOCA, even though the OTSG's are unevenly plugged with more tubes plugged in A than B. If a cold leg break occurs in the A side, the reduction of RC fluid is part of that blown out of the break, and there will be no impact to the result at all. Hcwever, a break in the B side will result in I slightly less total fluid passing through the core during the blowdown period. For about 11,000 ft' total fluid loss within approximately 24 seconds, the reduction of 200 ft' will correspond to 0.4 second shorter blowdcwn time and thus a slightly earlier fuel 1
heatup between blowdown and refill. This difference is minimal and' therefore, the resulting peak cladding temperature that occurs during this developed reflood stage should not be changed. Also,
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TR 022 Rev. O Page 20 vent flow is conservatively neglected during the refill /reflooding phases of LBLOCA analyses for the 177FA Lowered Loop Plants. Tube plugging will therefore have no impact on core flooding rates.
Examination of the effe' cts of plugging a total of 3000 tubes concluded that effects of initial RCS liquid inventory are insignificant (Reference 18). Even with as many as 3000 SG tubes plugged, TMI-l flows are expected to be greater than analysis assumed flows. However, for the reason of not exceeding the current Technical Specification limit on total RC flow (106.5% of design),
the number is instead set at 2000 in this report.
3.2 FSAR Analyses of Other Transients An assessment of the impact of the plugged steam generator tubes on the ability of the NSSS to safely respond to FSAR transient conditions has been performed. The plant is expected to be operated with up to a total of 3000 SG tubes plugged at the licensed rated power level. Each event in the TMI-l FSAR will be addressed in light of the expected impact of steam generator tube plugging on assumptions used to produce the current FSAR analysis,
- a. Uncompensated Operating Reactivity Changes This event is core burnup related and is normally compensated for by Integrated Control System action over the life of the fuel cycle. <
Steam generator plugging will not affect the core kinetics and thus have no impact on the event.
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TR 022 Rev. O Page 21
- b. Startup Accident /CRA Withdrawal at Power The CRA withdrawal from startup conditions and at power results in primary system over-pressurization. The FSAR prediction of RC pressure and peak theral power is based on the conservative assumption that all heat produced in the core remins in the primary system, i.e. no stean generator heat transfer. The plugging of 3000 3
tubes will result in a 400 f t volume reduction of the primry coolant (4%). At peak theral power, the reactor coolant pressure increase was 118 psi to 2318 psia in the FSAR. With the s'all volume reduction and consequently slightly higher heatup rate, the peak pressure my increase slightly but will remin well below the 2750 psig limit. Therefore, this event even with plugged tubes will remin well below the acceptance criteria on thermi power and system pressure.
- c. Moderator Dilution Accident The moderator dilution event is a relatively slow over-pressurization transient due to increased reactivity by boron dilution. Change in steam generator plugging will not affect the basic assumptions of this analysis and therefore, the FSAR remins bounding,
- d. Cold Water Accident (Pump Startup)
The pump startup event is a smil overcooling transient due to an increase in flow from an idle loop. The analysis performed for Section 2.3 demonstrated that the pump characteristic curves differences are insignificant for the unplugged and the 3000 plugged
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TR 022 Rev. O Page 22 tube cases. The reactivity change will cause a power and RCS pressure increase. The transient will be terminated by either the high reactor pressure trip or the power / flow trip. The FSAR remains bounding.
- e. Loss of Coolant Ficw See'Section 2.3.
- f. Stuck /Drcpped Red Esent The FSAR analysis is bounding since SG heat transfer and RCS flow do not effect this event
- g. Loss of Electric Pcwer The unit will trip on loss of electric power. With the loss of the reactor coolant pumps, r.atural circulation in the primary loop and heat removal by the emergency feedwater system are required. The impact of tube plugging on the ability for natural circulation cooling was demonstrated in Section 2.S.
- h. Steam Line Failure The licensing basis for TMI-l is the double-ended rupture. This FSAR analysis is based on a very conservative prediction of SG secondary inventory. Operation with plugged tubes results in a secondary inventory greater than operation without plugged tubes but it is not as great as that considered for the FSAR analysis.
Secondary inventory is one of the parameters that determine the
TR 022 Rev. O Page 23 safety considerations of return to criticality, and reactor building pressure. The calculated water increase with 1500 plugged tubes is less than 5% which will result in a maximum steam generator inventory of 42,000 lb. per steam generator for a clean steam generator. The FSAR analysis assumption used a steam generator inventory of 55,000 lb. per steam generator. For 3000 tubes plugged, the estimated water inventory is less than 44,000 lb which is still substantially less than the FSAR value. It is therefore concluded that the FSAR case remains bounding.
- 1. Steam Generator Tube Failure The steam generator tube rupture accident is analyzed assuming a
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435 gpm leak from a completely severed OTSG tube. The RCS is depressurized and isolated at 34 minutes, at which time leakage from the RCS is assumed to stop. The reduced RC flow as a result of the plugged tubes is greater than the RCS flow assumed for this cooldown rate. Similarly, more than enough OTSG heat transfer area is available to cool the RCS, Offsite dose from the tube rupture event will not be affected by plugging 3000 tubes because neither the time required to isolate the OTSG nor the leak rate from the broken tube is affected by the tube plugging.
- j. Fuel Handling Accident This accident is assumed to occur du.-ing a refueling outage while the reactor is shutdown. Change in steam generator plugging pattern has no impact to the assumptions. t
TR 022 Rev. O Page 24
- k. Rod Ejection Accident fast reactivity excursiens are not influenced by SG heat removal.
The event is an adiabatic heatup. The FSAR analysis remains bounding.
- 1. Maximum Hypothetical Accident The analysis assumed that a given amount of radioactivity nas been released folicwing core exposure and studies the effectiveness of the building spray system and engineering safeguard systems leakage on to the environment. The steam generators are not related to the scenario and thus have no impact on the conclusion.
- m. Haste Gas Tank Recture The waste gas tank is located in the auxiliary building and the '~
analysis of its rupture is not related to the steam generator's function. The event is thus unaffected.
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- n. Loss of Main Feecwater/Feedwater Line Break A loss of feedwater accident is an event resulting in primary system heatup, increased pressurizer level and pressure, and reactor trip either by anticipatory function (loss of main feedwater pumps) or high RCS pressure. The long term cooling relies on emergency feedwater heat removal through the steam generators. With the plugging of 3000 tubes in the steam generators, the initial heatup
, rate will be slightly faster. However, the anticipatory triD on -
high pressure will shut the reactor down and reduce the heat input
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Rev. 0 Page 25 4
toitsdecayheatlevelregardlesso(theminordifferenceinheatup s
ra.e. Emergency feedwater has th'e flow capability of removing de:ay heat un to about 77. power. lThis is greate'r tlian the decay heat at any time after shut'own. d IntheSBLOCAan$lysisusingtherevised w .. I
- LOCA model, it was demonstr3ted that heat tr"ansfer rate is not signif*1cantly changed wt'th this number of plugged tubes. Therefore, w -
! the FSAR analysis of tM loss of feedwater accident remains valid.
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- o. Steam Cenerator Overfill ,
^
Steam gererator overfiii was analyzed as a part of the TMI-l Restart Report (Reference 19). Thisanaly0Qidentifiedthatittakesat S least 10 to 17 minutes fy.r ausiliary feedwater to overfill to the -
v
^ top of the steam generator's shroud. Operators are instructed to isolate the feedwater flow Dath as soon as the OTSG water level '
'\
reaches the high level alarrN'on the operating range and to trip.or
' throttle feedwater pumps ,if the level reaches high-high. :
\ _
s m s %, -
The impact of up to 2250 plugged. tubes in the "A" steam generatoE q % );
~
> will be about a 7.57. inventory Increase as represented by thie ' level indication. 7 dis implies a reduction of the overfilling tim by
\
about 80 to 120.seccnds. Thetibefortheoperatortorespondjto ,
u 3 the high leveltalarm will be shortened. Moreover since there is kg still sufficient time and unamei"gucus symptoms.available for the
~
ooerators, their prcmpt response is expected and thus the overfill )
w'ou'l'd be corrected. In addition, a stress analysis has been A, {-
3- %
i s ,
, 3 h'
m TR 022 h Rev. 0-Page 26 q performedontheconsequencesoffloodingtheTMIUnit1MainSteam s i l i,n e . The results of deadweight, internal pressure, and thermal
, expansion analyses show that the main steam piping can withstand these effects. Therefore, operating the steam generators with 3000
, total plugged tubes in a 3:1 ratio will not present a safety concern y f.
with respect to steam generator overfill, n .s 4 i
- i. 'p. Locked Rotor Incident cx A Cycle 5 analysis (Re#erence 20) was performed to demonstrate the Yi1 effectofpicgding3000tubesinthesteamgeneratorsforalocked
)
'r rotor incident. The TMI-l licensing analysis for a locked rotor is
\ s presented in the Cycle 2 Reload Report. Table 2 documents the key
'
- parameter values used for these two analyser..
m , ,, .
As"dnbeseenfromTable2,thecoreflowforboththeCycle2and Cycle 5 locked rotor analyses are almost identical. The Cycle 2 Nv[,
. i-
' S C anal sis assumes an open internal vent valve and orifice rods in the core., The Cycle 5 analysis takes credit for the vent valve surveillance program allowing the analysis to close the vent valve; however, the orifice rods have been removed increasing the core
.. bypass flow. The Cycle 2' analysis uses the safety analysis power
,m y ', - level and RCS flow while the Cycle 5 analysis uses the realistic
- s. L power level and RCS flow explained in Section 2.1. Finally, the A, Cycle 5 aeclysis uses the Cycle 5 design radial peak which was o redu:ed from the Cycle 2 design racial peak.
.i s t
?
e
)
- . -' 'a
- TR 022 Rev. O Page 27 Figure 6 presents the results of.the two analyses. From these results, it was determined that the Cycle 2 analysis bounds the Cycle 5 analysis. This demonstrates that the plugging of the 3000 steam generator tubes would not invalidate the conclusions drawn from the Cycle 2 locked rotor analysis.
L U -
p q.>y'r,
- "* TR 022 Rev. O Page 28
4.0 CONCLUSION
S Evaluation has shown that up to 3000 total plugged tubes with 75% in one steam generator would have no adverse effects on plant safety. The reductions in flow and heat transfer are not large enough to affect the licensing basis analyses for transients or accidents. Even though plant safety will not be affected by plugging 3000 tubes, the safety evaluation limits the total number of plugged tubes to 2000. This conservative approach will ensure that all of the e<isting TMI-1 Technical Specifications are still bounding.
It is therefore determined that the result cf tnis modification shall not: 1) increase tne possibility cf occurrence of an accident or malfunction of equipment Important to Safety previously evaluated in the Safety Analysis Report; 2) increase the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; or, 3) reduce the margin of safety as defined for-Technical Specification; and thus does not involve an "Unreviewed Safety Questicn" as defined in 10CFR50.59.
s.
i i '
TR 022 Rev. O Page 29
5.0 REFERENCES
- 1. GPU TR 008 Rev. 3, " Assessment of TMI-1 Plant Safety for Return to Service After Steam Generator Repair", September 2, 1983.
- 2. B&W Letter GPUN-84-252, December 17, 1984, L.J. Stanek to D.G. Slear.
- 3. " Flow Verification for TMI-1", Med Ed Memo from J. Wilkerson to J.J. Colitz. March 24, 1980.
- 4. BAW-177 " Preliminary Calculations of the Effect of Plugged Steam Generator Tubes on Plant Performance", March 1982.
- 6. CHATA-Core Hydraulics and Thermal Analysis-Revision 4, BAW-230, Rev. 4, Babcock & Wilcox, June 1979.
- 7. CIPP-CHATA Input Processing System.
- 8. TEMP-Thermal Enthalpy Miving Program, BAW-321, Rev. 2, Babcock & Wilcox, June 1979.
. 9. B&W Document 51-1155450-01, " Increased OTSG Plugging".
- 10. B&W Document No. 32-1135 309-00, " Pump Code Certification for Oconee II" by D.J. Halteman, July 1982.
- 11. " AUX. A Fortran Program for Dynamic Simulation of Reactor Coolant System and Emergency Feedwater System." B&W Document NPGD581, August 1981.
- 12. " Evaluation of SBLOCA Operating Procedures and Effectiveness of Emergency Feedwater Spray for B&W Designed Operating NSSS". B&W Report.
- 13. Requests for Information on Steam Generator Feedwater Addition Events Letter from Thomas Cox of USNRC to J.H. Taylor, (B&W) dated June ~ 20, 1982.
14 NUREG 0565, " Generic Evaluation of Small Break Loss-of-Cooldfd Accident Behavior of B&W Designed 177FA Operating Plants".. January 19,80.
- 15. Topical Report BAW 10092P Rev. 3, October 1982.
- 17. B&W Docueent 51-1155610-0, " Maximum Allowable SG Tube Plugging at TMI"
- 18. Preliminary Assessment, " Effects cf SG Tube Plugging on Post-LOCA Core Safety", 1/15/82, prepared for GPU by B&W.
- 19. Report in Response to NRC Staff-Recommended Requirements for Restart of l Three Mile Island Nuclear Station Unit One - Amendment 25.
' Flow Coastdown Curve".
TABLE 1
'IN_I-1 CYCLE 1 TO CYCLE 4 RC FLOW CALCULATIONS LOOP A LOOP B TOTAL CALCULATED CALCULATED CALCULATED TEST TEST RC FLOW RC FLOW RC FLOW % OF CYCLE 1 FLOW
- NUMBER DATE (MLB/HR) (MLB/HR)_ (MLB/HR) DESIGN FLOW IMBALANCE 1 06/15/76 73.559 71.428 144.986 110.41 2.94 2 11/05/76 73.538 71.765 145.302 110.65 2.44 3 02/09/77 74.223 71.690 145.914 111.11 3.47 4 05/21/77 72.986 71.106 144.093 109.73 2.61 5 07/29/77 73.901 70.891 144.792 110.26 4.16 6 10/25/77 73.748 70.563 144.311 109.89 4.41 7 01/29/78 73.298 70.509 143.809 109.51 3.88 8 05/08/78 73.973 71.206 145.179 110.55 2.67 9 07/26/78 73.639 71.701 145.340 110.68 2.67 10 11/05/78 73.721 71.414 145.136 110.52 3.18 11 02/05/79 73.376 70.895 144.272 109.86 3.44 Flow Imbalance = (Calc Flow A - Calc Flow B) *2/ Total Calc Flow X 100%
1
% ." o god
l l
l
. .. 1 _
TR 022 Rev. O Page 31 TABLE 2 TMI-I LOCKED ROTOR ANALYSIS REFERENCE CONDITIONS CYCLE 2 CYCLE 5 Power. level, MWt' 2568 2535'
-System flow (7. of FSAR value) 106.5 104 2
Vent valves 1 open All closed Effect of open vent' valse on systen flow (flow facter) 1.01 ---
Leakage ficw factor for cpen vent valve .943 ---
Orifice' rod assemblies Yes No Core bypass other than vent valves (ficw factor) .917 .896 Core flow (flow factors a system flow) .930' .932' Radial peak 1.78 1.71 Licensed power level.
Conservative-reduction in-system flow to account for the 3000 steam generator tubes plugged.
Fraction of FSAR. design system flow.
alii U22 FIGung i Rev. O Page 32 REDUCTION IN CC FLCW RATE VSINU PLUGGED PER STEAM GENERATOROF T 110 p-7 k
3 10sL I
h h \
3 10s L 9 1 Y
s.
I 104 ~
2 a i m
.a
$ 102 -
100 L- f g 500 1000 1500 2000 2500 3060 3500 4000 Number of Tubes Pluggedin OTSG
O COMPARISON OF FOUR PUMP F: LOW C0ASTDOWN CURUES - FSAR US.
}_
120E1 1S00 AND 3000 TOTAL PLUGGED TUBES un tra cuzzua n nu '
~
g, 90 -
a e
O
=
1 pS l
( # 1500 Phsgeed Tubes l
60 - 3 l I
! ~
E 'N 3999 PLUGCED TUBES
! o FSAR o o C
a g
a 30 - s 1
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