ML20207K386
| ML20207K386 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/30/1986 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Berkow H Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20207K390 | List: |
| References | |
| P-86682, TAC-63576, NUDOCS 8701090408 | |
| Download: ML20207K386 (15) | |
Text
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Cornpany of Colorado December 30, 1986 Fort St. Vrain Unit No. 1 P-86682 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Mr. H. N. Berkow, Director Standardization and Special Projects Directorate Docket No. 50-267
SUBJECT:
Analysis of Firewater Cooldown for 39% Power Cperation
REFERENCES:
- 1) LER 86-020, dated August 11, 1986 (P-86513)
- 2) LER 86-026, dated August 17, 1986 (P-86587)
- 3) PSC Letter, Warembourg to Berkow, dated October 10, 1986 (P-86580)
- 4) PSC Letter, Williams to Berkow, dated October 22, 1986 (P-86585)
Dear Mr. Berkow:
The purpose of this letter is to provide the NRC with the analysis that justifies power operation up to 39% power based on Safe Shutdown Cooling following a 90 minute Interruption of Forced Circulation (IOFC). As reported in References 1 and 2,
inadequacies were identified in the FSV FSAP analysis for Safe Shutdown Cooling from 105% power utilizing the steam generator reheater and economizer-evaporator-superheater (EES) section.
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P-86682 Dacemb:;r 30, 1986 As a result of these reports and as indicated in Reference 3, PSC has developed a plan and schedule for the re-analysis associated with the use of the steam generators for Safe Shutdown Cooling. As indicated in Reference 4, PSC has also developed a plan and schedule for re-evaluating the Appendix R model in light of the impact of the LER contained in Reference 2.
These plans and schedules are contained in of this letter.
Please recognize that this submittal is but one part in the overall rise-to-power program for FSV.
PSC is preparing the overall rise-to-power plan which will be submitted to you in the imediate future.
l This letter contains the analysis to justify power operation up to 39% power based on utilizing either the reheater or EES section for Safe Shutdown Cooling following a 90 minute IOFC. This power level l
is limited by the reheater section of the steam generator. As shown in Attachments 5 and 7, higher power levels (72%) can be supported by an EES cooldown. The various analyses contained in this submittal are summarized as follows:
l - Corrective Action Responses for LER 86-020 (Reheaters)
This attachment is provided to inform the NRC of what actions PSC is pursuing to resolve the issues identified in the LER. - Corrective Action Response for LER 86-026 (EES)
This attachment is provided to inform the NRC of what actions PSC is pursuing to resolve the issues identified in the LER.
Firewater Cooldown Using One Reheater Module (1 1/2 hour delay) - GA Report 909113
Purpose:
This report justifies that cooldown utilizing the reheater section of the steam generator is adequate for power levels up to 39% based on the assumption that only one reheater module may be flooded by firewater.
Major Assumptions:
Firewater cooling following a 90 minute 10FC Only one of six reheater modules is conservatively assumed to be flooded 39% reactor thermal power prior to 10FC m
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P-86682 Decemb:r 30, 1986 Resul ts:
Peak fuel temperature of 1800 degrees Fahrenheit is below the FSAR fuel temperature limit of 2900 degrees Fahrenheit The steam generator maintains its integrity since the maximum economizer outlet tube temperature of 1312 degrees Fahrenheit is below the allowable temperature of 1350 degrees Fahrenheit and the tube temperature does not exceed 1300 degrees Fahrenheit for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (See Figure 4-4)
The circulators remain operable since the maximum circulator inlet temperature of 1251 degrees Fahrenheit is below the allcwable temperature for circulator operation The PCRV relief valve may lift. However, this can be avoided by partially depressurizing as analyzed in Attachment 10. - EES Cooldown From 39% and 78% Power Using Condensate or Firewater (1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> delay) - GA Report 909268
Purpose:
This report justifies that cooldown utilizing the EES section of the steam generator is adequate from a power level of 72% for the Appendix R condensate train model and a power level of 78% for the Appendix R firewater train model and for Safe Shutdown Cooling using firewater.
Major Assumptions:
Firewater or condensate cooling utilizing the EES section of one steam generator following a
90 minute 10FC 78% reactor thermal power prior to 10FC for the Safe Shutd6wn Cooling firewater flow path and the Appendix R firewater train 72% reactor thermal power prior to 10FC for Appendix R condensate train Existing Safe Shutdown Cooling firewater flow path t
P-86682 December 30, 1986 Existing Appendix R flow paths for Train A and B Results:
All three cases were analyzed and based on fuel temperature, the Appendix R condensate train model was determined to be the worst case The Safe Shutdown Cooling case using firewater and
.the Appendix R firewater train model were determined to provide an adequate cooldown from 78% reactor power The Appendix R
condensate train model was determined to provide an adequate cooldown from 72% reactor power based on interpolation of fuel temperatures for the cases that were analyzed - Effect of Firewater Cooldown Using Reheater on Steam Generator Structural Integrity - GA Report 909190
Purpose:
To evaluate the most critical regions (maximum allo,iable tube temperature) of the steam generator during a sirgle cycle cooldown using firewater in one reheater module Major Assumptions:
Firewater cooling following a 90 minute 10FC from 105% power Only one reheater module is flooded and the EES is dry Maximum tube differential pressure of 832 psi Resul ts :
The reheater tubes will withstand the stress due to thermal shock from a firewater cooldown following operation at any power level including 105% power 1
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P-86682 Dscember 30, 1986 Steam generator EES tubes maintain integrity if core outlet helium temperatures are limited to 1350 degrees Fahrenheit maximum temperature and do not remain above 1300 degrees Fahrenheit for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> - Effect of Firewater Cooldown Using EES Bundle on Steam Generator Structural Integrity - GA Report 909204
Purpose:
To evaluate the most critical regions (maximum allowable tube temperature) of the steam generator during a single cycle cooldown from 105% power using firewater in the EES bundle Major Assumptions:
Firewater cooldown following a 90 minute 10FC from 105% power Maximum tube differential pressure of 832 psi Resul ts:
The reheater tubes without cooling water, which are the most critical items during an EES cooldown, will not suffer a creep collapse during the single cycle firewater cooldown event if the maximum-helium temperature is less than 1660 degrees Fahrenheit for less'than 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> and the corresponding helium pressure is not more than 350 psig. - Summary of Structural and Metallurgical Effects of a Firewater Cooldown Transient Upon the EES and Reheater-Sections of the Fort St. Vrain Steam Generators The purpose of this report is to summarize the structural and metallurgical effects of a firewater cooldown on steam generator tube integrity. - Analysis of the Capability of the Fort St. Vrain Steam-Generator to Withstand the Firewater Cooldown Transient Following an Appendix R Fire
Purpose:
To analyze the structural integrity of the steam generators following an Appendix R firewater cooldown with respect to thermal shock, vapor lock, water hammer and overpressurization.
P-86682 December 30, 1986 Major Assumptions:
Firewater cooling utilizing the EES section of one steam generator following a 90 minute 10FC 100% reactor thermal power prior to IOFC A conservative 1000 GPM of firewater flow enters the steam generator at 80 degrees Fahrenheit A conservative 1600 degrees Fahrenheit helium enters the steam generator Results:
The thermal shock on the tubes results in stress levels that are not significant from a fatigue standpoint Vapor lock and flow stoppage will not result due to a firewater cooldown but boiling must be prevented by limiting helium flow rate to avoid reduced cooling water flow rates Damage to the steam generators due to condensation induced water hammer will not occur An overpressure condition will not occur during flooding of the steam generators with firewater This analysis determined that the EES section can be flooded with firewater within approximately 14 minutes.
This is in conflict with assumptions used in other reports contained in this submittal.
However, all of these reports are conservative since a minimum of 90 minutes was assumed for the period of 10FC.
PSC's position is that the 10FC will not exceed 90 minutes and water flow initiation will begin in approximately 75 minutes. 0 - Effect of Intentional Depressurization on Cooldown from 39% Power Using 1 Reheater Module (1 1/2 hour delay)
GA Report 909217
Purpose:
To analyze the effect of intentionally depressurizing the PCRV to avoid lifting of a PCRV relief valve
P-86682 December 30, 1986 Major Assumptions:
Firewater cooling following a 90 minute 10FC from 39% power Only one reheater module is flooded and the EES is assumed to be empty Depressurization times of 700, 1200 and 1800 sec.
were analyzed Resul ts:
Peak fuel temperature of 1800 degrees Fahrenheit is below the FSAR fuel temperature limit of 2900 degrees Fahrenheit The 1800 sec depressurization case is the worst case Maximum economizer tube temperature of 1323 degrees Fahrenheit due to this worst case is still below the 1350 degrees Fahrenheit maximum temperature and it does not exceed 1300 degrees.
3 Fahrenheit for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 1 - FSV Calculation for Circulator Temperature - Related Operating Li. nits - GA Report 908861 3
Purpose:
To establish operating limits for the helium circulators related to the helium temperature passing through the compressor blading Major Assumptions:
Helium temperature represents the maximum temperature of circulator components Results:
Operating limits are plotted as a function of speed and temperature in -this report 2 A comparison of FSAR EES and Reheater Transients Currently Being Analyzed with Previously Completed 39% Power Level Cooling Analyses i
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P-86682 December 30, 1986
Purpose:
To reaffirm the acceptability of the scenarios in the FSAR which rely on the EES Section or Reheater Section of a steam generator for decay heat removal Major Assumptions:
The seven scenarios looked at rely on feedwater condensate or firewater as analyzed in the FSAR Resul ts:
Adequate decay heat removal can be achieved from operation at 39% power for each of the 7 FSAR scenarios analyzed The above documentation regarding cooling using the reheaters is based upon flooding only one of six reheater modules of a steam generator when an accident requires initiation of Safe Shutdown Cooling to protect the health and safety of the public. However, it is noted that tests are in progress to determine whether the phenomena that prevents the flooding of more than one reheater module can be overcome by a change in procedures.
If the cause can be determined, then procedures can be developed to assure that firewater flow through all six reheater modules of a steam generator can be established and maintained during any postulated condition.
Therefore, PSC's position is that the postulation of only one reheater module of a steam generator being available for Safe Shutdown Cooling could change in the future.
If ou have any questions, please contact Mr. M. H. Holmes at (303 480-6960.
Very truly yours, 0 VrYw d y D. W. Warembourg, M(nager Nuclear Engineering Division DWW/KD:pa Attachments
P-86682 D:cember 30, 1986 P-86682 Attachment 1 Steam Generator Technical Evaluation Plan and Schedule As discussed in the cover letter, PSC has committed to submitting various analyses to support Safe Shutdown Cooling utilizing the two steam generator sections.
The plan and schedule for the technical 4
analysis to support safe shutdown is listed below. The logic _ behind each step is explained and the power level that can be supported is listed.
Step 1 - P1 Power Level The goal of Step 1 is to identify a power level at which both the steam generator reheater section and economizer-evaporator-superheater(EES) section can support Safe Shutdown Cooling. This results in an acceptable power level which does not involve a change to the Technical Specifications or the existing Appendix R models and does not require further plant modifications.
Analysis submitted with this letter supports power operation up to 39% power.
The analysis includes reports to justify cooldown utilizing the reheater section up to 39% power.
Separate reports also evaluate cooldown utilizing the EES section at 72% and 78%
j power.
This power level envelopes 39% power and utilizes existing flow patns for Safe Shutdown Cooling scenarios and shutdown cooling following a fire per 10 CFR 50 Appendix R.
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Step 2 - P2 Power Level Step 2 involves a change to the FSV Technical Specifications. A submittal will be sent to the NRC by January 12, 1987 which will i
request removal of the reliance on the reheater section of the steam i
generators for Safe Shutdown Cooling.
A Safety Evaluation Report will accompany the Technical Specification change request.
As a
result of the Technical Specification change, a plant modification is required to meet single failure criteria.
This modification involves installation of a vent line on each main steam loop header and will be installed prior to power operation at P2 power level.
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A separate submittal will be sent to the NRC by January 12, 1986 which will specify.the P2 power level and will provide the technical justification for power operation at this power level. Attachment 5 to this letter already provides justification for power operation up to 72% reactor power in which the Appendix R flow paths represented the bounding case. Since a power level of greater than 72% is desired, a change to the Appendix R model will be required. A separate submittal will be forwarded to 'the NRC by March 17, 1987 documanting a revision to the Appendix R model.
Step 3-100% Power Operation Based on preliminary analysis, it is felt that the P2 power level will be below 100% power even after the plant modification and Appendix R model revision discussed above have been incorporated.
Therefore, further plant modification may be required to justify power operation up to 100% power. The alternatives for supporting a power level above P2 have not been evaluated at this time.
At the present time, PSC considers the 100% operation goal to be a longer term goal (beyond the next refueling). Therefore, if step 3 is going to be pursued, PSC will provide the NRC with the necessary analysis and justification to support 100% power.
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1 to P-86682 Corrective Action Responses to LER 86-020 (Ref. 1)
CORRECTIVE TION:
1.
An analysis is being done on changes to the input parameters used in the analysis to determine if the reheaters can be made to function adequately for a firewater cooldown following a 90 minute delay.
For example, the residual water in the EES tube bundles during the reheater cooldown, the back pressure control setting was assumed in past analyses to be 850 psia rather than the operating value of 2412 psia.
Back pressure of 850 psia would be appropriate for certain transients such as the wrong loop dump situations, since this is the automatic action taken by the Plant Protective System (PPS).
However, for other accidents including the seismic event, this automatic PPS action would not occur and the EES back pressure control would be maintained at 2414 psia unless manually run back by the operators.
The higher back pressure would retain more inventory in the EES and enhance cooling.
RESPONSE
An increase in back pressure was investigated. However, the effect on increased cooling was not appreciable. Therefore, to be conservative, the analysis contained in this submittal takes no credit for heat transfer to the EES section of the steam generator during a cooldown utilizing flow to the reheat section.
CORRECTIVE ACTION:
2.
Plant inodifications are being investigated to determine if they would enable the reheater to meet the design objective of removing the decay heat when supplied with firewater.
to P-86682
RESPONSE
In order to maintain low enough tube temperatures during cooldown on the reheaters, flow to all modules would be required.
It is still our estimation that extensive modifications will be required to ensure flow to all reheater modules. We are however, pursuing actual tests to accurately determine what would be required to obtain flow to all modules.
CORRECTIVE ACTION:
3.
A change to the licensing basis is being investigated that would delete the reheaters as an optional method for safe shutdown using firewater.
In conjunction with this option, PSC is re-analyzing the use of the EES sections for all accidents involving a 90 minute 10FC followed by a cooldown with firewater. This option would necessitate a License amendment.
RESPONSE
As discussed in the cover letter, this is the option that PSC is pursuing.
Future submittals will provide the analysis, proposed Technical Specification revision and Safety Evaluation to support this approach at a power level greater than 39% power.
This is discussed in further detail in Attachment 1.
CORRECTIVE ACTION:
In addition to the above corrective action, PSC committed to analyzing the maximum power level at which the reheaters remain viable as a safe shutdown option. This letter contains that analysis which supports a power level of 39%.
to P-86682 Corrective Action Responses for LER 86-26 (Ref. 2) 1 CORRECTIVE ACTION:
1.
A review, and re-analysis as necessary, will be performed on the various accidents described in the FSAR that rely on shutdown cooling using the steam generator EES section with water supplied by a condensate pump or the various cases utilizing firewater pump capacity. This will determine if the recent discoveries regarding steam binding in the EES with water supplied by a firewater pump after imposing a 1 1/2 hour delay would have any impact on other analyzed FSAR accidents.
kESPONSE:
A summary of the accidents reviewed can be found in 2.
CORRECTIVE ACTION:
2.
Re-analyses will be completed as necessary to support a graduated rise-to-power program based on an acceptable cooldown of the plant for the relevant accident scenarios and proposed reactor power limits. The re-analyses will be submitted for NRC review and approval.
RESPONSE
As described in the cover letter, Attachments 5, 7 and 8 provide the analysis to show that the steam generator EES section is adequate for Safe Shutdown Cooling up to at least 39% reactor thermal power. The analysis may be based on a higher power level but the present analysis bounds the conditions at 39% power.
As discussed in Attachment 1,
further analysis will be submitted to the NRC to support the P2 power level.
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to P-86682 CORRECTIVE ACTION:
3.
The analyses required to support the above corrective actions will be thoroughly evaluated to verify that adequate capacity exists to provide the required secondary flow rate.
All results will receive a thorough independent review,
RESPONSE
The analysis provided in Attachment 5 evaluates the system pressure drops and the firewater or condensate pump capacity.
All the analyses have received an independent review by GA who is a qualified vendor for safety related services.
PSC and Proto-Power Corporation from Groton, Connecticut have also reviewed the analysis to ensure its accuracy.
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