ML20207K512
| ML20207K512 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/13/1986 |
| From: | Connors G, Mcdonald C, Nichols M GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | |
| Shared Package | |
| ML20207K390 | List: |
| References | |
| 908861, TAC-63576, NUDOCS 8701090437 | |
| Download: ML20207K512 (25) | |
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SUMMARY
TITLE FSV CALCULATIONS FOR CIRCULATOR O R&D 2 APPROVAL LEVEL TEMPERATURE-RELATED OPERATING LIMITS &8 S GN DISCIPLINE SYSTEM 00C. TYPE PROJECT l 00CUMENT NO. ISSUE N0lLTR. I M 21 CFL 1900 903861 l N/C i QUALITY ASSURANCE LEVEL SAFETY CLASSIFICATION SEISMIC CATEGO RY ELECTRICAL CLASSIFICATION QAL I FSV-I FSV-I N/A APPROVAL d B ENGINEERING QA
^' ^
JfCT PR ECT C'. O.
N/C
$1/,h r NOV 131986 M. Nichols
, ,i j.P.Connors M JM
.J. Kennedy Initial Release C.h Donald CUBS 2970106 l
CONTINUE ON GA FORM 14851 NEXT INDENTU RED DOCUM ENTS Text 1-8 = 8 l
Appendix A Al-All = H N6757 Total 19 8701090437 861230 PDR F
ADOCK 05000267 pga l
REV l l l l l l l SH l l l l l l l l l l REV l l l l l l l l l
l l l l l l l SH 29 l 30 l 31 l 32 33l 34 35 l 36 l 37 38 l 39 l 40 41 42 43 44 45 l 46 47 l 48 49 50 l 51 l 52 53 l 54 l 55 l 56 :
REV l l l l l l l l l l l l l l l l l l l l l l l l SH 1 l 2 l 3 l 4 l 5 l 6 l 7 l 8 l 9 l 10 l 11 l 1213 14 l 15 l 16 l 17 l 18 l 19 20 21 l 22 l 23 l 24 l 25 l 26 l 27 l 23 TROYER, M. AAWOR Doc. No. 2, SR-7342
- r. i:.\i... w CONTENTS
- 1. INTRODUCTION . . . . . . . . . . . . .. . . .. . . . . .. 3
- 2. METHOD AND ASSUMPTIONS . . . . . . . . . . . . . . . . . .. 3 3 RESULTS AND DISCUSSION . . . . . . . . . . . . . . . . . .. 5 4 REFERENCES . . . . . . . . . . . . . . . . . . . . . . .. . 5
- 5. CALCULATION REVIEW REPORT . . . . . . . . . . . . . . . . . 8
- 6. APPENDIX A . . . . . . . . . . . . . . . . . . . . . . . . . A-1 FIGURES 1 FSV circulator estimated operational limits .. . . . . . . 4
- 2. Cross section of dovetail . . . . . . . . . .. . .. . .. 7' l
l Page 2 l
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- 1. INTRODUCTION A study was made and a calculation was performed to establish operating limits for the Fort St. Vrain (FSV) circulator related to the helium temperature passing through the compressor bladingr This information can be used to evaluate options for operation of the plant for certain postulated licensing events.
, 2. METHOD AND ASSUMPTIONS A review of the materials properties included in Refs. 1 through 4 was conducted for the compressor blades (SST 422) and the compressor disk (Ladish D6AC). Precise data at elevated temperatures for the particular heat treatments specified on the circulator drawings (Refs. 5 and 6) were not found, but sufficient information was available to establish the approximate curves shown on Fig. 1.
i The following assumptions were factored into the calculation:
- 1. Normal bearing water pressure and temperature conditions exist within the circulator bearing cartridge.
- 2. The helium temperature represents the maximum temperature I
reached by the compressor blades and the outer rim of the compressor disk.
3 Normal buffer helium conditions are present.
4 Blade and disk stresses are proportional to speed squared.
- 5. Unlimited operation of the circulator is permissible if the calculated maximum stress is I 67% of the material yield strength.
Page 3 1
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- l 3 RESULTS AND DISCUSSION A summary of the calculations performed for this study are included in Appendix A. The blade footings are designed to fail first (Fig. 8 of Ref. 7, also included in Appendix A) to provide a failure mode (blade
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shedding) that would preclude disk fracture for virtually all postulated conditions. Comparison of the available materials data, for various temperatures, to the expected circulator rotor operating stresses for various speeds confirmed that the blade material should be considered as the limiting factor. Thus, Fig. I represents the estimated FSV circu-lator operating limits based on the strength of the compressor blade footing (dovetail root) where the blade is installed in'the compressor disk (see Fig. 2).
Operation of the FSV circulator beyond line 1 of Fig. 1 is not reco= mended except for short durations under emergency requirements.
Line 2 of Fig. 1 is the approximate limit for 1000 h of operation. The circulator should not be operated to the right of line 2 except for less than 100 h as indicated by the shaded area in the lower right portion of the graph. Because of the approximate nature of the materials data, it is recommended that the information on Fig. 1 be used only for analytical studies.
4 REFERENCES
- 1. Department of Defense Aerospace Structural Materials Handbook, -
i 1974, 1976.
- 2. Metals Handbook. Eighth Edition, Vol. 1, published by the American Society for Metals.
I 3 Briggs, J. Z., and T. D. Parker, "The Super 12% Cr Steels,"
Climax Molybdenum Co. of Michigan.
Page 5
- 908861nJC 4 Smith, G. V. , " Evaluations of the Elevated Temperature Tensile and Creep Rupture Properties of 12'. to 20", Chromi a Steels,"
ASTM DS 59.
- 5. Ccmpressor roter disk drawing 90-C2101-362, [C.
- 6. Cc= presser rotor disk drawing 90-C2101-363, [J.
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1 0^5?b' a .: G CROSS SECTION OF DOVETAIL
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t G A 1543t R EV.11/30) 1 CALCULATION REVIE1N REPORT l
TITLE: ffv C ficc a,m:ss f04 t ' A'C ~J n ,2, APPROVAL LEVEL '-
T L f 4c m ;,c d - f 5c.J 7 n c M. tw4,- A m ,ry QAL LEVEL l DISCIPLINE SYSTEM 00 C. TYPE PROJECT DOCUMENT NO. a ISSUE NO., LTR.
M d( C FL t400 9 088lo / O[d-INDEPENDENT REVIEWER:
NAME VY$5I#
O RG ANIZATION V m 'T v 56? v C Si , ,
REVIEWER SELECTION APPROVAL: BR MGR
\
cu 0) DATE M\6fi R EVIEW METHOD: YES NO ERROR DETECTED ARITHMETIC CHECK LOGIC CHECK d0 ALTERNATE METHOD USED #
SPOT CHECK PERFORMED COMPUTER PROGRAM USED I V REMARKS: (ATTACH LIST OF DOCUMENTS USED IN REVIEW)
CALCULATIONS FOUND TO BE VAll0 AND CONCLUSIONS TO BE CORRECT:
INDEPENDENT REVIEWER ATE ll !'d ![A (Sjf/ NATURE l
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APPENDIX A FSV CIRCULATOR STRESS CALCULATIONS TO DETERMINE TEMPERATURE-RELATED OPERATING LIMITS The purpose of this appendix is to document the logic and calculations that were used to generate Fig. 1, included in this report 908861. Figure 1 provides the approximate operating limits for the FSV circulator based on the compressor blading (helium inlet) temperatures.
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r- -W e 9 4 ATTACHMENT 12 l l i l 1 i i i 4 t l l l I k.
Attachment 12 to P-86682 Page 1 of 5 A COMPARIS0N OF FSAR EES TRANSIENTS CURRENTLY BEING REANALYZED WITH PREVIOUSLY COMPLETED 39 PERCENT POWER LEVEL COOLING ANALYSES Six accident scenarios presented in the FSAR which rely on the EES section of the steam generators for safe shutdown cooling are being reanalyzed to reaffirm their acceptability for safe shutdown cooling. The accident scenarios being reanalyzed are identified in Table 1. Although the thermal-hydraulic analysis of these transients has not been completed, work accomplished to date on both these transients and on other transients indicates that safe shutdown from operation at 39 percent power level is assured. These six transient scenarios rely on feedwater, condensate and fire water cooldown flow paths. Of these, the fire water and condensate flow paths are similar to other conditions previously analyzed for 39 percent power level. Fire Water (Case 3) The fire water flow path is essentially the same as that for safe shutdown cooling following a high energy line break, which is the cooldown path previously analyzed by Proto-Power. (This previously analyzed flow path is shown in Figure A-1 of GAR Report 909268.) This analysis concluded that the flow path would support a 832 GPM flow through the EES section, which, per analysis performed by GAT and documented in GA Report No. 909268, is adequate to ensure safe shutdown cooling from a 77.9% power level. This previously analysis further assumes a 90 minute delay in forced circulation cooling. The abnormal shutdown cooling transient reported in the FSAR would be expected to provide cooling from an even higher power level since no extended delay in forced cooling is postulated. Condensate (Cases 2, 4, 5, and 6) These four transients rely on condensate cooldown via the EES section and use a flow path similar to the Appendix R Train A flow path previously analyzed by Proto-Power. (This previously analyzed flow path is shown in Figure A-2 of GAT No. 909268.) That Appendix R flow path and the condensate cooldowns being evaluated each rely on a 12-1/2% condensate pump to provide cooling to one steam generator EES section via the emergency condensate header. In both flow paths, steam generator outlet is directed to the bypass flash tank drain via the main steam desuperheaters. The condensate is directed to the Decay Heat Exchanger (for heat removal) in Appendix R cooldown, whereas, the main condenser is used for cooling with the FSAR transients. Since the path from the EES discharge to the condenser is less restrictive than the path from the EES discharge to the decay heat exchanger, increased flow rates can be expected. Both the Appendix R transients and the FSAR transients rely on helium circulation via the pelton wheel drives with emergency condensate. The major difference
l l l Attachment 12 l to P-86682 ' Page 2 of 5 l between these cooldown scenarios is the 90 minute Interruption of Forced Circulation (10FC) delay postulated for the Appendix R cooldown, which is not included in the FSAR transients. This is a major consideration in view of the difficulty in restarting flows after a 10FC delay. Since previous analysis has demonstrated that the Appendix R Train A cooldown can provide a 518 CP,*1 flow rate to the steam generators, and thus provide adequate cooldown from a 72% power level, the FSAR transients are again expected to provide cooldown from a higher power level, and would certainly be adequate for the 39% power level. - Feedwater (Case 1) The DBA-2 transient differs from other transients analyzed for the 39% power level in that feedwater is used for helium circulation and steam generator cooldown. Since the much smaller capacity condensate (580 gmp 9 300 psig) and fire water (1500 gpm, 150 psig) pumps have been demonstrated capable for safe shutdown cooling, the much higher capacity feedwater pump, rated for 1700 gpm at 3300 psig is expected to have sufficient capacity to provide steam generator cooling and pelton wheel operation of two circulators. High pressure feedwater is required for this accident since the circulators must be operated at 8000 RPM in order to provide adequate helium flow with the depressurized PCRV; the circulator pelton wheels will require approximately 415 gpm at 1500 psia each,during- the accident. GAT Report No. Depressurization/ Blowdown, Delayed Forced Coolir.g" analyzes the effects of an 10FC. during DBA-2. Based on helium temperatures reported by GAT in the report, a maximum heat removal capacity of 77 x E-6 BTU /hr will be required for cooldown following the accident from 39% power. Preliminary results of the. ongoing analysis indicates that the feedwater cooldown will provide more than adequate cooling for this heat load. Based on these results, there will be no steady state boilinc4 in the steam generator, and throttling will be required to m41ntain pressures in the cooldown path within allowable ranges. With this capacity, the feedwater . system will be more than adequate for decay heat removal cooling following a DBA-2 accident at the 39% power level. i ? , - _ - - -. ,, .-. , , - - - .
TABLE 1 Attachment N3. 12 P-86XXX' . FSAR TRANSIENTS FOR REANALYSIS OF EES C00LDOWN Page 3 of 5 ,
-FSAR HEAT COOLING DELAY IN
.ASE SECTION FIGURE DESCRIPTION REMOVAL WATER COOLING FLOWPATH FOR COOLING WATER 1 14.11.2.2 14.11-13 DBA-2 RAPID DEPRESSURIZATION-TWO EES FEEDWATER 60 MIN. FEEDWATER PUMPS TO EES VIA EMER-HELIUM CIRCULATORS ON FEEDWATER GENCY FEEDWATER HEADER, THRU EES TO BYPASS FLASH TANK VIA DESUPER-HEATERS, AND THEN TO MAIN CONDEN-SER VIA FLASH TANK DRAINS. 2 14.4.2.1 14.4-2 ABNORMAL SHUTDOWN C00 LING-1 EES CONDENSATE 0 MIN. 12-1/2% CONDENSATE PUMPS TO EES HELIUM CIRCULATOR ON CONDENSATE VIA EMERGENCY CONDENSATE HEADER, THRU EES TO BYPASS FLASH TANK VIA DESUPERHEATERS AND THEN TO MAIN CONDENSER VIA FLASH TANK DRAINS. 3 14.4.2.1 14.4-3 ABNORMAL SHUTDOWN C00 LING-1 EES FIREWATER 0 MIN. FIREWATER PUMPS THRU EMERGENCY - HELIUM CIRCULATOR ON B00STED CONDENSATE HEADER, THRU EES TO FIREiATE2 BYPASS FLASH TANK VIA DESUPER-HEATERS AND THEN T0 MAIN CONDENSER VIA FLASH TANK DRAINS. 4 10.3.1 10.3-1 COINCivu.. .nt LOSS OF OUTSIDE EES CONDENSATE 0 MIN. SAME AS 2 AB0VE POWER AND MAIN TURBINE TRIP - SWITCH FROM STEAM TO CONDENSATE TO TWO HELIUM CIRCULATORS AFTER 25 MINUTES 5 10.3.2 10.3-2 SAME AS 4 AB0VE, BUT ONLY 1 EES CONDENSATE 0 MIN. SAME AS 2 AB0VE STANDBY GENERATCR - SWITCH FROM STEAM TO CONDENSATE.T0 ONE HELIUM CIRCULATCR AFTER 35 MIN. 6 10.3.7 SIMULTANEOUS LOSS OF 3 FEEDWATER EES CONDENSATE 0 MIN. SAME AS 2 AB0VE (HELIUM PUMPS - ONE HELIUM CIRCULATOR ON CIRCULATOR NOT STARTED UNTIL CONDENSATE CONDENSATE FLOW THRU EES HAS BEEN ESTABLISHED.)
r i Attachment 12 to P-86682 Page 4 of 5 A COMPARISON OF FSAR REHEATER TRANSIENTS BEING REANALYZED WITH PREVIOUSLY COMPLETED 39 PERCENT POWER LEVEL C0OLING ANALYSES As a matter prudence, accidents described in the FSAR that rely on the reheater section of a steam generator were reviewed to assure the adequacy of cooling down from 39% power operation. The steam generator tube rupture with a wrong loop dump accidents Case 2 (FSAR Section 14.5.3.2) and Case 5 (FSAR Section 14.5.3.4) have been reevaluated relative to the potential problems of cooling water flowrate, steam formation and the heat capacity of the isolated EES sections without exceeding a fuel temperature of 2900 degrees F. Case 5 is considered the worst case, enveloping the results of Case
- 2. In Case 5, following a tube rupture in an EES section of a steam generator, reactor decay and stored heat continues to be removed from the reactor by the leaking EES of the operating steam generator until recovery of cooling is made with the intact EES in the dumped loop.
If for some reason it is not possible to recover the intact and dumped EES, condensate may be admitted to a reheater as a backup means of cooling. Then the leaking EES loop is isolated and dumped. The reheater to be used is not placed in service until the core has been cooled (in about 30 minutes) to approximately 450-500 degrees F by the leaking EES. This delay in starting cooldown with a reheater minimizes thermal shock effects and enables the EES to remove a substantial amount of core decay .and stored heat before cooldown commences with the reheater. Reanalysis of the Case 5 accident from 105% power operation concludes that fuel temperatures remain below 2900 degrees F, but a PCRV relief valve will lift twice if only one of the six reheater modules in a steam generator is flooded with condensate supplied by one of the two small condensate pumps as described in the FSAR accident scenario. However, it was determined that if two reheater modules can be flooded with condensate, the PCRV relief valves will not lift during the reheater cooldown following the worst case wrong loop dump scenario (Case 5) from 105% power. Use of the large condensate pumps are being evaluated for the purpose of flooding two or more reheater modules of a steam generator for recovery from this accident. Since PSC does not at this time have an evaluation which supports flooding more than one reheater module with condensate, then only one ' reheater module should be considered to be available for removal of core and stored heat during the Case 5 accident following operation at 39% power. This scenario (using only one small condensate pump for cooling after operation at 39% power) should not result in lifting a PCRV relief valve as is the case for 105% power operation. The reason for this is that while the heat removal capacity of one reheater module is only one-half of that required to prevent lifting of a PCRV relief valve following 105% power operation, the reactor l decay heat generated from operation at 39% power is less than one-l half of that generated from operation at 105% power. Therefore, one
r- r Attachment 12 to P-86682 Page 5 of 5 reheater module flooded with water from one small condensate pump should have sufficient heat removal capacity to preclude lifting of a PCRV relief valve following the worst case wrong loop dump accident occurring from 39% reactor power. _. 5}}