ML20215J735
| ML20215J735 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 05/01/1987 |
| From: | Brey H PUBLIC SERVICE CO. OF COLORADO |
| To: | Calvo J NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| P-87158, TAC-66574, NUDOCS 8705080225 | |
| Download: ML20215J735 (11) | |
Text
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2420 W. 26th Avenue, Suite 100D, Denver, Colorado 80211 May 1, 1987 Fort St. Vrain Unit No. 1 P-87158 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555 Attention: Mr. Jose A. Calvo, Director Project Directorate IV Docket No. 50-267
SUBJECT:
Additional Information for 82%
Power Regarding Fuel Temperature Sensitivity Analysis
REFERENCES:
- 1) PSC letter Warembourg to Berkow, dated December 30, 1986 (P-86683)
- 2) NRC letter Heitner to Williams, dated February 3, 1986 (G-87031)
- 3) PSC letter Brey to Berkow, dated February 17, 1987 (P-87055)
- 4) GA Document No. 909393(N/C),
dated May 1, 1987
- 5) PSC letter, Williams to Berkow, dated January 30, 1987 (P-87038)
Dear Mr.
alvo:
l The purpose of this letter is to submit results of the Public Service Company of Colorado (PSC)
Fuel Temperature Sensitivity Analysis.
This analysis, performed for PSC by GA Technologies (GA), evaluated the effects of controlling steam generator outlet teaperature at a value below that previously evaluated and reported to the NRC for shutdown cooling using Safe Shutdown Cooling and Appendix R Cooldowr, l
models (Attachment 1 to Reference 1). This analysis was performed in l
response to NRC Request No. 3 (contained in Reference 2), and the PSC l
commitment (contained in Reference 3) to forward the sensitivity analysis results.
8705080225 870501
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-P-87158 Page'2 May 1, 1987 Th'e' GA analysis (Reference 4) evaluated the effects of reduced steam generator outlet temperatures during the Safe Shutdown Cooling and Appendix R-Train A Shutdown Cooling, to determine the resultant effects on core fuel temperature due to the reduced heat transfer.
The need to perform a temperature sensitivity analysis was identified-i to account for the potential for reduced heat removal rates during shutdown cooling when steam generator outlet temperatures are below the values utilized in the base case analyses.. The values of steam generator outlet temperature utilized in the base case analyses of to Reference 3 (GA Report 909269 IssueA)are. identical to the values which the operators will be required to maintain in the procedure for Safe Shutdown Cooling and identical _ to the values anticipated.for the Appendix R Train A Shutdown. Cooling Procedure, although this procedure has not been finalized.
The lower steam generator outlet temperatures utilized in the Fuel Temperature Sensitivity Analysis accommodate (1) conservative allowances for temperature deviation from the specified control point and (2) temperature instrument inaccuracy. 'provides a summary of the results contained in the GA sensitivity analysis- (Reference 4).
The major finding of this analysis is-as follows:
(1) For the EQ case from 87.5% reactor power, only 1.35 percent of the fuel exceeds the 2900 degrees F FSAR limit.for approximately three hours, reaching a maximum temperature of 2959 degrees F.
-(2) For the Appendix R case from 83.2% reactor power, less than 0.5 percent of the fuel exceeds the FSAR limit for approximately three hours, reaching a maximum temperature of 2939 degrees F.
As further noted in Attachment 1, it is PSC's conclusion that these conditions, although slightly in excess of the FSAR limit, are acceptable due to the -fact that these conditions will not lead to significant fuel damage. Maximum fuel temperatures would be less if the fuel temperature sensitivity analyses had been performed from an initial reactor power level of 82%, which is the power level restriction proposed by PSC and undergoing review by the NRC~ staff.
The following conclusions remain unchanged from the original analysis of Reference 1:
(1) The hottest module of the steam generator was maintained at-subcooled conditions at all times for both cases (EQ and Appendix R) evaluated.
(2) The pressure required to supply water to the steam generator for cooldown is within the capability of the condensate and the firewater pumps used within the shutdown cooling models for each Case.
(3) Maximum steam generator tube temperatures remain within allowable limits.
r.
6 P-87158
..Page 3
.May.1, 1987 Results of the Reference 4 analysis, as summarized.in Attachment 1, support safe operation of FSV at reactor power _ levels up to 82%,
which is a power restriction proposed by PSC to the NRC 'in Reference
- 5...
If you have any questions, please contact Mr. M. H. Holmes at (303) 480-6960.
Very truly yours, OQhD D
H. L. Brey, Manager Nuclear Licensing and Fuels Division HLB /CRB:jmt Attachment cc: Regional Administrator, Region IV ATTN: Mr. J. E. Gagliardo, Chief Reactor Projects Branch Mr. R. E. Farrell Senior Resident Inspector Fort St. Vrain
ATTACHMENT 1 TO P-87158 FUEL TEMPERATURE SENSITIVITY ANALYSIS:
EFFECTS DUE TO REDUCED STEAM GENERATOR OUTLET TEMPERATURE DURING C00LDOWN-I.
PURPOSE This -attachment-summarizes the analysis performed by GA as documented in Reference 1,.GA Document No.
909393 (N/C),
"FSV EES Cooldowns (1.5 :Hr Delay) Steam Generator Outlet Water.
Temperature Sensitivity." The objective of this analysis was to determine the effects on maximum fuel temperatures if the actual temperature maintained at the steam generator outlet was lower than those utilized in the' base case analysis through a combination of conservative allowances for temperature deviation from the specified control point and temperature instrument inaccuracy._ This analysis evaluated shutdown cooling modes for the following cases (descriptions of these shutdown models are contained in Reference 2):
EQ Case Safe Shutdown Cooling on equipment satisfying environmental qualification (EQ) testing requirements, consisting of an open loop arrangement supplying firewater to one-steam generator economizer-evaporator-superheater (EES) section and a Pelton wheel drive to one helium circulator.
Appendix R Train A Case cooldown using equipment satisfying 10CFR50 Appendix R requirements involving water supply from the condensate system operating in.an open loop for. five hours, followed by closed loop operation.
Parameters for this cooldown model are more restrictive than those evaluated for the Appendix R _ Train B model and-therefore only the Train A case was evaluated.
Evaluation'of each of these scenarios was based upon restoration of forced circulation cooling following a 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Interruption of Forced Circulation (10FC). The total temperature reduction evaluated by the analysis was 13 degrees F below that evaluated in the base case analysis of Attachment 2 to Reference 4, which was determined by combining the maximum instrument temperature t
inaccuracy
(+8 degrees F) and a conservative, constant temperature control allowance 5 degrees F below the specified
-temperature control point.
The analysis demonstrates that in this worst case scenario (i.e., actual temperature maintained 13 degrees F below the established control temperature for each case), that the heat removal rates would not be decreased to a point-which would result in significant fuel failure.
For the Appendix R Train A case, however, the closed loop portion does not use a reduced temperature because the operator is no longer controlling the outlet temperature.
The pressure across the 3
Pelton wheel nozzle driving the circulator has reached its upper limit so the helium flow can no longer be increased to maintain steam generator outlet temperature. Therefore, the 13 degree F
r to P-87158 Page 2 temperature reduction is not applicable during this phase of the cooldown.
II. BACKGROUND A.
Basis for the Instrument Inaccuracy The maximum instrument inaccuracy was determined by Proto-Power Corp. in Reference 3 for instrumentation in the main steam vent lines used to locally monitor steam pressure and temperature.
Results of the Proto-Power.
calculations indicate that the accuracy of the temperature instrument is 18 degrees F.
B.
Basis for the Operator Control Tolerance By procedure, the operators will be required to maintain the following temperatures and pressures during each of the following cooldown cases:
EQ Case - maintain steam generator outlet temperature at 255 degrees F with a main steam backpressure of 76 psia.
Appendix R Case maintain steam generator outlet temperature at 311 degrees F (steam backpressure of 198 psia) during the initial open loop phase of the cooldown, and maintain the outlet temperature at 368 degrees F (steam backpressure of 268 psia) during the closed loop phase of the cooldown.
In each case, no allowable control band has been identified and operators will be required by procedure to maintain indicated temperatures and pressures at the above specified values. Due to the fine control that can be achieved in controlling circulator speed (+20 RPM), actual control of outlet temperature is expected to be within 11 degree F of the specified values.
This assures that subcooled conditions will be maintained within the steam generator hot module at all times and that no significant fuel damage will occur due to inadequate decay heat removal during shutdown cooling.
The 5 degree F temperature below the specified control point was evaluated to conservatively envelope unanticipated deviations from the specified control point.
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. to P-87158 Page 3 III. METHOD The Reference 1 GA -temperature sensitivity analysis used the same shutdown cooling'models as those previously identified in to Reference-4 The only parameter which was changed was steam generator outlet temperature; all other i
parameters. remained. unchanged. The following assumptions were made in the GA analyses of References 1 and Attachment 2 to Reference 4, the base _ case analysis:
(1) The PCRV liner cooling system was unavailable.
(2) The thermal capacity of the steam generator reheater module was ignored.
. (3) The cooldown was assumed to occur.from 87.5% power in the EQ case and 83.2% power in the' Appendix R Train A case.
The following computer codes were used to perform the analysis.
The TAP code was used to evaluate steam generator performance.
The RECA code was used to provide detailed core performance.
The SUPERHEAT code provided detailed steam generator water side pressure drop and was used-to determine the required helium flow rates for input to the RECA code.-
Hot.and cold helium temperatures during the 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 10FC-were obtained.from a previous analysis that used the RATSAM code (Reference 5).
To determine if there would be _ boiling in the hot module, the HOT
Previous Proto-Power calculations were used to determine the overall water side pressure drop.
'IV.
CONCLUSIONS Operating conditions _-for each of the two cases evaluated are-
- identified in the attached Table-1.
Evaluation of the. water side pressure drop by Proto-Power.in Appendix-A of Reference 1
~ indicates that the variation in steam generator water outlet temperature introduces no significant change in cooling water flow rate. Overall results of the analysis are presented in Table 2.-
Analyses -results indicate that the cooldown can be performed within the head capability of the supply pump for each of these p'
cases.
The minimum subcooled margin is 31 degrees F for the EQ case and 42 degrees F. for the Appendix R Train A case.
The maximum hot module steam generator helium inlet temperature for the EQ case was 1495 degrees F and for the Appendix R Train A l
p case was 1471 degrees F; both maximum temperatures are within allowable values for the helium inlet temperatures. The primary
- system pressure remained below the PCRV relief valve setpoint I
throughout the cooldown for both cases.
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-Attachment.1 to P-87158 Page 4 The maximum. fuel temperatures slightly exceed the FSAR temperature limit of 2900 degrees F.
For the EQ case, the results of this analysis indicate that the temperature limit of 2900'de three hours grees F was exceeded for a period of approximately (by 1.35 percent of the fuel) and the hottest transient node temperature was 2959 degrees F.
The temperature limit of 2900 degrees F was also exceeded during the Appendix R Train A case (by less than 0.5 percent of the fuel) for a-period of approximately three-hours and maximum temperature experienced-was 2939 degrees F.
These fuel.-temperatures were based on continuous operation at 87.5% rated power during the EQ case and i
83.2% rated power during the Appendix R Train A case prior to l
the event.
As noted in a previous submittal to the NRC (e.g.,
Confirmatory Action 1 to Reference 6),
2900 degrees F is a conservative lower limit for the onset of fuel failure. Because of the magnitude of the peak fuel temperature and the short duration at these' temperatures, significant failure of the affected fuel would not be anticipated.
As stated:in Reference 6:
"It is important to note that the 2900 degrees F value _is a conservative lower limit for fuel failure.
The actual temperature at which fuel failure takes place is several hundred degrees higher.
During the fuel manufactur.ing
-process the coated fuel particles within the rod are exposed to temperatures in the range of 2900 degrees to 3340 degrees F -for a period of one to two hours.
In addition, tests have been perfonned which demonstrate that the 2900 degrees F limit is an unrealistically low value.
The results of these tests were documented in an attachment
.to a letter dated July 24, 1979 (P-79157), Swart to Speis (NRC). Also see FSAR Table A.2-9 for further fuel particle coating failure data as~ a function of higher temperatures."
The -maximum peak fuel temperature in the PCRV liner cooling analyses for 35% power discussed in Reference 6 was 3160 degrees F
(i.e.,
over 200 degrees F hotter than the 2959 degrees F of this analysis). This submittal was the basis for NRC approval of interim operation at 35% power (Reference 7).
Even if it is assumed that a maximum of 1.35 percent of the core will exceed 2900 degrees F during these' cooldown events and suffer failure-of the fission product barrier fuel particle coatings, the radiological consequences would not increase.
Primary coolant leakage and consequent radiological dose calculations in the FSAR assume a core average of five percent failed fuel, which is the basis for the "879 Mw(t) Design" (i.e., 105% power) fission product primary coolant inventory of
-FSAR Table 3.7-1.
Operations to date indicate the fuel failure is actually at least an order of magnitude below this design basis. Therefore, an additional 1.35 percent fue! failure would not approach the five percent failure assumed for most FSAR dose
. calculations.
to P-87158 Page 5
-Further, this analysis has shown that the PCRV safety valves would not lift and the primary coolant pressure boundary would not be challenged.
As a consequence, any fission product release from the fuel would be expected to remain within the PCRV until removed by the helium purification system.
Results of the Reference 1 analysis indicate that the original conclusions of Reference 4 remain suitably justified, and support safe operation of FSV at reactor power levels up to 82%.
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to P-87158 Page 6 REFERENCES (1) GA Document No. 909393 (N/C), dated May 1, 1987 (2) PSC letter, Warembourg to Berkow, dated December 30, 1986 (P-86683)
(3) Proto-Power Calculation No. 82-16-(Rev. A), dated March 6, 1987.
(Proto-Power File No. 7511482)
(4)* PSC letter, Brey to Berkow, dated February 17,1987(P-87055)
(5) Cadwallader, G.-J., "PCRV Depressurization Analysis During LOFC
- FSV (105% Power, 2-Hour Delay Through As-Built Train with Rerouted 2-Inch Pipe)," SAM:113:GJC:77, May 3, 1977 (6) PSC letter, Walker to Berkow, dated December 10,1985(P-85460)
(7) NRC letter, Denton to Walker, dated February 7, 1986 (G-86062)
Reference 4 has attached GA Document No. 909393 Issue A, which.
is the base case analysis.
I to P-87158 Page 7 TABLE 1 i
SLMRRY T TERATING CCEDITICES JTpaMiv R Train A EQ Base
-13 F water
-13 F Open Closed Open Closed Parameters Base water Icop I. cop Icop Icop Water flow, gun 940 940 700 491 700 491 Water inlet temp., F 80 80 100 139 100 139 Water outlet temp., F 255 242 311 368 298 368 Water outlet press., psia 76 76 198 268 198 268 Hoagdutyem=hility 82.3 76.2 73.5 56.2 69.0 56.2 10 Btu /hr 39f 24.1 22.3 21.5 16.5 20.2 16.5 (Excerpted from GA Document No. 909393 (N/C))
i
1 1
I j
TAELE 2 7E#
j EES C00LDOWN
$ aE mb EQ Appendix R Train A Base
-13 F water Base
-13 F water SS i
Primary Side Results:
)
Min core inlet helium temp., *F 100 102 114 118 Circulator helium flow, 1 1.4 - 3.6 1.3 - 3.6 1.4 - 3.6 1.3 - 3.6 j
Max. fuel temp.,
F 2858 2959 2875 2939 l
Max. avg. module S.G. helium inlet temp., *F 1406 1400 1391 1378 Max hot module S.G. helium inlet temp.,
F 1501 1495 1484 1471 1
l Steam Generator Results:
Initial feedwater flow, I 85 85 80 80 Initial reactor power, I,F 100/139(,)
100/139 87.5 87.5 83.2 83.2 Inlet water temperature, 80 80 i
Water flow, gym 940 940 700/491 700/491 j
S.G. outlet water temp.,
F 255 242 311/368 298/368 l
Max. circulator inlet temp.,
F 120 120 150 148 j
Max. economiser outlet tube temp.,,F 90 90 155 157 Max. superheater outlet tube temp.,
F 307 307 333 346 Hot module boiling margin, *f 12 31 23 42 l
S.G. outlet pressure, psia (S.G. ring header) 95.6 96.1 217/284 217/284
]
S.G. outlet pressure, psia (main steam header) 76 76 198/268 198/268 l
S.G. pressure drop, paid 38.2 38.1 22/12 22/12 1
S.G. inlet pressure, psia 133.8 134.2 239/296 239/296 i
j Calculated Results:
Total pump flow, gym 1065 1065 825/666 825/666 l
S.G. water flow, gym 940 940 700/491 700/491 S.G. outlet pressure, psia (main steam header) 72 76 198/268 198/268 j
Required S.G. inlet pressure, psia 133.3 133.3 239/296 239/296 j
Available S.G. inlet pressure, psia 137 133.5 239/298 239/298 (a) Dual values denote open loop / closed loop values for Appendix R Train A Case i
(excerpted from GA Document No. 909393 (N/C))
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