BVY 04-026, Technical Specification Proposed Change No. 257, Supplement 7, Implementation of Arts/Mella at Vermont Yankee, Revised Anticipated Transient Without Scram Evaluation
| ML040830676 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 03/11/2004 |
| From: | Thayer J Entergy Nuclear Vermont Yankee |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BVY 04-026 | |
| Download: ML040830676 (15) | |
Text
_lk 7Entergy Nuclear Vermont Yankee, LLC Entergy Nuclear Operations, Inc.
Brattleboro, VT 05302-0500 March 1, 2004 BVY 04-026 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Subject:
Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)
Technical Specification Proposed Change No. 257 - Supplement No. 7 Implementation of ARTS/MELLLA at Vermont Yankee Revised Anticipated Transient Without Scram Evaluation By letter dated March 20, 2003, as supplemented by letters dated March 31, 2003, April 17, 2003, June 11, 2003, July 21, 2003, December 11, 2003, and February 10, 2004, Vermont Yankee' (VY) proposed to amend Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station (VYNPS) by implementing an expanded power-to-flow operating domain based on the Average Power Range Monitor, Rod Block Monitor Technical Specifications/Maximum Extended Load Line Limit Analysis (ARTS/MELLLA) methodology.
During a recent review of the Anticipated Transient Without Scram (ATWS) event analysis that was performed by General Electric (GE) in support of the MELLLA, VY identified certain design input discrepancies. The subject ATWS analysis was included as Section 9.0 of Attachment 52 to the March 20, 2003 license amendment request (LAR). Due to the discrepancies, VY re-verified the entire set of inputs provided to GE and had GE re-perform the ATWS analysis. VY verified that no other ARTS/MELLLA analyses were affected by these changes. Because VY has now completed the design for a modification to install an additional spring safety valve (SSV) and increase the capacity of the existing SSVs during the upcoming refueling outage (scheduled for the spring 2004), the revised ATWS analysis also incorporates this design change.
The changes in key input parameters result in more accurate or realistic analytical inputs to better represent actual plant configuration or planned modifications that will be implemented during the upcoming refueling outage. The following are the key input parameter changes:
- Standby Liquid Control System (SLCS) Liquid Transport Time (changed from 33.3 sec to 44.9 sec) - Discrepancies noted in the length of SLCS piping led to a recalculation of the transport time. Additionally, conservatism was added by assuming the length of pipe between the tank and the pumps is not filled with sodium pentaborate solution.
'Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. are the licensees of the Vermont Yankee Nuclear Power Station.
2GE Nuclear Energy, "Vermont Yankee Nuclear Power Station, APRM / RBM / Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA),"
NEDC-33089P, March 2003.
BVY 04-026 / Page 2
- ATWS Recirculation Pump Trip Sensor Time Constant (changed from 0.0 sec to 0.2 sec) -
The original analysis assumed a generic value of 0.0 sec. Review of VYNPS plant-specific instrumentation resulted in the change to a more appropriate 0.2 sec value.
- Residual Heat Removal System Heat Exchanger Effectiveness - (changed from 176 BTU/sec-
°F to 179.375 BTU/sec-0 F) - This value was revised based on heat exchanger performance for the assumed suppression pool temperature range.
- SSV Capacity per Valve (changed from 932,500 Ibm/hr to 1,126,200 lbm/hr) - When the original analysis was performed the specifics of the planned SSV modification had not yet been finalized. The revised analysis now assumes the new increased capacity for the three SSVs (i.e., the modifications to the two existing SSVs and the addition of another SSV. These modifications will be implemented during the upcoming refueling outage).
- Safety Relief Valve (SRV)/SSV Configuration (changed from 4/2 to 3/3) - The original ARTSJMELLLA ATWS analysis assumed the then-current configuration. The revised analysis updates the configuration to the post-refueling outage configuration with one SRV conservatively-assumed to be out-of-service.
The revised analysis is provided as Attachment I to this letter. It supplements, but does not totally replace the evaluation provided as Section 9.0 of Attachment 5 to the March 20, 2003, LAR. In particular, statements including proprietary information regarding fuel integrity and the acceptance criteria of IOCFR50.46 are unchanged.
The change in some of the input parameters to the ATWS analysis caused certain subtle and readily understandable changes in the key performance parameters noted in this submittal. The peak suppression pool temperature and peak containment pressure for the limiting ATWS events increased slightly, due mainly to the increase in the transport time from the SLCS tank to the point of vessel injection. The calculated peak vessel pressure decreased slightly for these events due to the increased capacity of the SSVs. As can be expected, the timing of some of these events also changed slightly.
The net results of these changes are increases in peak suppression pool temperature of approximately 2IF and peak containment pressure of 0.4 psi, and a decrease in peak vessel pressure of approximately 9 psi. These results remain within applicable limits.
Additionally, the input parameter changes caused some less subtle and more apparent changes in certain output parameters. For instance, the peak neutron flux for the pressure regulator failure-open (PRFO) event at end-of-cycle changed from 696 % rated in the original analysis to 334 % rated in the revised analysis. This change is attributed to the fact that the changes that were made to the input parameters caused a slight decrease in the void reactivity at the time of the peak neutron flux. The void reactivity behavior is driven by pressure waves developed during main steam isolation valve closure. The difference in the neutron flux, integrated over time, is small enough that the difference in the peak heat flux is insignificant.
Another change to the output is noted in the times of peak suppression pool temperature. While the times in most events increased slightly, the time of peak suppression pool temperature for the PRFO event at beginning-of-cycle (BOC) decreased slightly. Since the suppression pool temperature response is essentially flat from about 1,000 to 2,000 seconds, the timing of SRV cycles can have an impact on the location of the peak, while not affecting the magnitude of the peak. Because the original ATWS analysis was performed with four SRVs, and the revised analysis assumed three SRVs, a change in timing would be expected. A similar situation occurs in the time when hot shutdown is
BVY 04-026 / Page 3 achieved. The time for most events increased slightly, while the PRFO event at BOC time decreased slightly.
VY evaluated the impact of the above changes on the response 3 to Question 8 of NRC's request for additional information. There are no consequential impacts on the key ATWS mitigation success criteria for the events evaluated.
Although analytical inputs have been changed in regard to some operating conditions and equipment performance for the ATWS events, the results of the revised ATWS analysis illustrate that the key performance parameters of reactor vessel pressure, suppression pool temperature, and containment pressure remain within applicable limits. Therefore, operation of VYNPS under the conditions proposed by the LAR has no adverse effect on the capability of plant systems to mitigate postulated ATWS events.
This supplement to the LAR does not change the scope or conclusions in the original application, nor does it change VY's determination of no significant hazards consideration. VY regrets any inconvenience this change may have caused. If you have any questions in this regard, please contact Mr. James DeVincentis at (802) 2584236.
Sincerely, Jay. Chafer. I Sitkyce President STATE OF VERMONT )
)ss WINDHAM COUNTY )
Then personally appeared before me, Jay K. Thayer, who, being duly sworn, did state that he is Site Vice President of the Vermont Yankee Nuclear Power Station, that he is duly authorized to execute and file the foregoing document, and that the statements therein are true to the best of his knowledge and belief.
SaliA.ndstrum, Notary Public My Commission Expires February 10, 2007 p...
Attachment i SA4X cc: USNRC Region I Administrator (w/o attachment) N RN USNRC Resident Inspector - VYNPS (w/o attachment) N USNRC Project Manager - VYNPSi Vermont Department of Public Service Ho 3 Vermont Yankee letter to U.S. Nuclear Regulatory Commission, "Technical Specification ' Clt r1.
Proposed Change No. 257, Implementation of ARTS/MELLLA at Vermont Yankee, Response to Request for Additional Information," BVY 03-56, June 11, 2003.
Docket No. 50-271 BVY 04-026 Attachment 1 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 257 - Supplement No. 7 Implementation of ARTS/MELLLA at Vermont Yankee Revised Anticipated Transient Without Scram Evaluation
BVY 04-026 / Attachment I 1.0 ANTICIPATED TRANSIENT WITHOUT SCRAM 1.1 Introduction Section 9 of Reference I detailed the Vermont Yankee Nuclear Power Station (VYNPS) plant-specific evaluation of the Anticipated Transient Without Scram (ATWS) event in support of VYNPS operation in the Maximum Extended Load Line Limit Analysis (MELLLA) domain. Subsequent to the submittal of Reference 1 to the NRC, Entergy identified several discrepancies in the input parameters used in the ATWS evaluation. In addition, VYNPS has completed the design for the modification of the spring safety valve (SSV) capacity that -vill be installed in the refueling outage (scheduled for spring 2004) prior to APRM / RBM / Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA) implementation. Therefore, a revised ATWS analysis was performed to correct the input parameter discrepancies and reflect the final design of the SSV modification.
The key input parameters that have changed from the Reference 1 analysis are:
Parameter Rcfcrcncc I Revised Analysis Analysis Standby Liquid Control System (SLCS) Liquid 33.3 44.9 Transport Time (sec)
ATWS Recirculation Pump Trip Sensor Time 0.0 0.2 Constant (sec)
Reactor Heat Removal (RHR) heat exchanger 176 179.375 effectiveness (Btu/sec- 0 F)
SSV Capacity - per valve (Ibm/hr) / Reference 932,500 / 1240 1,126,200 / 1240 Pressure (psig)
Safety Relief Valve (SRV) / SSV Configuration 4 /2 3/3 used in the analysis 1.2 Approach/Mcthodology The approach/methodology for the revised ATWS analysis is unchanged from that of Reference 1.
The basis for the current ATWS requirements is 10 CFR 50.62. This regulation includes requirements for an ATWS Recirculation Pump Trip (RPT), an Alternate Rod Insertion (ARI) system, and an adequate SLCS injection rate. The purpose of the ATWS analysis is to demonstrate that these systems are adequate for plant changes associated with operation in the MELLLA region. This is accomplished by performing a plant-specific 1-1
BVY 04-026 / Attachment I analysis in accordance with the approved licensing methodology (Reference 2), to demonstrate that ATWS acceptance criteria are met for operation in the MELLLA region.
The ATWS analysis takes credit for ATWS-RPT and SLCS, but assumes that ARI fails.
If reactor vessel and fuel integrity are maintained, then the ATWS-RPT setpoint is adequate for the proposed plant modifications. If containment integrity is maintained, then the SLCS injection rate is adequate for the proposed plant modifications.
The two limiting ATWS events for VYNPS were re-evaluated at the most limiting MELLLA point (100% of Current Licensed Thermal Power (CLTP) and 75% of rated core flow (RCF)) with ARI assumed to fail, thus requiring the operator to initiate SLCS injection for shutdown. These limiting events are:
(1) Closure of all Main Steam Isolation Valves (MSIVC); and (2) Pressure Regulator Failure (Open) to Maximum Steam Demand Flow (PRFO).
The loss of offsite power (LOOP) and inadvertent opening of a relief valve (IORV) events wvere also considered, but found to be non-limiting. The following ATWS acceptance criteria were used to determine acceptability of VYNPS operation in the MELLLA region:
(1) Fuel integrity:
- Maximum clad temperature < 22000 F
- Maximum local clad oxidation < 17%
(2) Reactor Pressure Vessel integrity:
(3) Containment integrity:
- Peak suppression pool bulk temperature <281 'F
- Peak containment pressure < 62 psig The disposition of the compliance with the fuel clad integrity criteria is unchanged from that provided in Reference 1.
The adequacy of the margin to the SLCS relief valve lifting as described in NRC Information Notice 2001-13, "Inadequate Standby Liquid Control System Relief Valve Margin," was also assessed and determined to be adequate.
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BVY 04-026 / Attachment 1 1.3 Input Assumptions Along with the initial operating conditions given in Table 1-1, the following assumptions were used in the analysis. These assumptions are unchanged from the Reference 1 analysis:
Analytical Assumptions Bases/Justifications Both beginning-of-cycle (BOC) and end-of- Consistency with generic ATWS evaluation cycle (EOC) nuclear dynamic parameters bases.
were used in the calculations.
Dynamic void and Doppler reactivity are ATWS analyses are performed based on VYNPS Cycle 23 data. conservatively compared to a nominal basis, which bounds cycle to cycle variation. Thus, utilization of VYNPS Cycle 23 fuel parameters is appropriate.
One SRV out-of-service (SRVOOS), Consistency with the VYNPS Technical specified as the valve with the lowest Specifications.
setpoint.
SRV setpoints are adjusted to be consistent Consistency with the VYNPS Technical with the 3% setpoint tolerance relaxation. Specifications.
An additional SSV will be installed and the In order to ensure compliance with the capacity of the existing SSVs will be reactor vessel overpressure criterion.
increased prior to the implementation of MELLLA.
Main Steam Isolation Valve (MSIV) closure Consistency with generic ATWS evaluation starts at event initiation (time zero) for the bases.
MSIVC event.
Maximum combined flow through the main Conservatively bounds current VYNPS steam line flow limiters is 125% of rated Updated Final Safety Analysis Report steam flow (PRFO event). (UFSAR) PRFO analysis basis.
Currently, VYNPS has four SRVs and two SSVs. As stated in Reference 1, prior to the implementation of MELLLA, an additional SSV will be installed with the capacity listed in Table 1-1.
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BVY 04-026 / Attachment I 1.4 Analyscs Results MELLLA conditions provide the greatest effect on peak vessel pressure and peak long-term containment response (suppression pool temperature and containment pressure).
The constraints of one SRVOOS and SRV setpoint tolerance relaxation provide the greatest effect on the peak short-term reactor vessel pressure.
Tables 1-2 and 1-3 summarize the key transient responses for the MSIVC and PRFO events analyzed for the Reference I analysis and the revised analysis. As shown, the peak vessel bottom pressure result for both events in the revised ATWS analysis is 1358 psig, which is below the ATWS vessel overpressure protection criterion of 1500 psig. Also as shown, the revised analysis result is a decrease of 9 psi from the Reference 1 analysis peak vessel bottom pressure of 1367 psig. The decrease in the peak pressure is due to the higher SSV capacity used in the revised analysis that more than offsets the effect of a slightly longer ATWS Recirculation Pump Trip Sensor Time Constant.
The highest calculated peak suppression pool temperature is 185 0 F for the revised analysis, which is well below the ATWS limit of 281'F. Also, the peak containment pressure of 11.5 psig is a small fraction of the 62 psig limit. Thus, the RPV and containment integrity criteria for ATWS are met. The highest calculated peak suppression pool temperature for the revised ATWS analysis is slightly higher (20 F) than that reported in Reference 1. This difference is due to the following:
. The longer Standby Liquid Control (SLC) transport time used in the revised analysis results in a slightly longer delay to start boron injection and initiate reactor shutdown with SLC.
- The use of a 3 SRV / 3 SSV configuration in the revised ATWS analysis results in a longer time before the boron injection initiation temperature is achieved and subsequently the start of boron injection and the action to reduce feedwater flow to the reactor pressure vessel. The net result of these longer delays is a slight increase in the integrated heat addition to the suppression pool and a subsequent slightly higher suppression pool temperature.
As stated in Reference 1, compliance with the applicable acceptance criteria necessitates the installation of the third SSV. The additional SSV, as well as the increased SSV capacity of the existing two SSVs, has been designed and will be installed during the upcoming refueling outage (scheduled for spring 2004) in accordance with applicable codes and standards. In addition, the installation was evaluated for jet impingement considerations and found acceptable. The heat load to the drywell by this additional SSV and the modification to increase the capacity of the existent SSVs was evaluated and found to be acceptable.
For the revised ATWS analysis, the maximum SLCS pump discharge pressure during the limiting ATWS event is 1318.5 psig. This value is based on a peak reactor vessel lower plenum pressure of 1289.2 psia that would occur during the PRFO BOC event at the time 1-4
BVY 04-026 / Attachment I of SLCS initiation. The results of the revised ATWS analysis show a decrease of less than 1 psi in peak reactor vessel lower plenum pressure from the Reference 1 analysis.
Even though the revised ATWS analysis indicates a slight reduction in the required SLC test pressure, VYNPS will continue to test the SLC system using the revised Technical Specification 4.4.A.l. that was submitted with Reference 1.
Finally, there is adequate margin to prevent the SLCS relief valve from lifting (per NRC Information Notice 2001-13). With a nominal SLCS relief valve setpoint of 1400 psig, there is a margin of 81.5 psi between the peak SLCS pump discharge pressure and the relief valve nominal setpoint, using the result of the revised ATWS analysis. This is a slight improvement from that reported in Reference 1.
1.5 Conclusions The results of the ATWS analysis performed for VYNPS to support operation in the MELLLA region show that the maximum values of the key performance parameters (reactor vessel pressure, suppression pool temperature, and containment pressure) remain
,within the applicable limits. Therefore, VYNPS operation in the MELLLA region has no adverse effect on the capability of the plant systems to mitigate postulated ATWS events in the expanded operating region.
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BVY 04-026 / Attachment I Table 1-1 Operating Conditions and Equipment Performance Charactcristics for ATWVS Analyses Paramctcr Refercnce 1 Revised Analysis
__ Analysis Dome Pressure (psig) 1010 1010 Core Flow (Mlb/hr / % rated) 36.0 / 75 36.0/ 75 Core Thermal Power (MWt / %NBR) 1593 / 100 1593 / 100 Steam / Feed Flow (Mlb/hr / %NBR) 6.45 / 99.8 6.45 / 99.8 Feedwvater Temperature (OF) 376.0 376.0 Initial Void Reactivity Coefficient - BOC / -16.8 / -12.0 -16.8 / -12.0 EOC (c/%)
Core Average Void Fraction - BOC / EOC (%) 58.7 / 46.5 58.7 / 46.5 Sodium Pentaborate Solution Concentration in 10.42 10.42 the SLCS Storage Tank (% by weight)
Nominal Boron 10 Enrichment (atom %) 43.0 43.0 SLCS Injection Location Lower Plenum Lower Plenum Number of SLCS Pumps Operating One One SLCS Injection Rate (gpm) 40.5 40.5 SLCS Liquid Transport Time (sec) 33.3 44.9 ATWS Recirculation Pump Trip Sensor Time 0.0 0.2 Constant (sec)
Initial Suppression Pool Liquid Volume (f13 ) 68,000 68,000 Initial Suppression Pool Temperature (F) 90 90 Initial Suppression Pool Mass (Mlbm) 4.216 4.216 Number of RHR cooling loops 2 2 RHR heat exchanger effectiveness (Btu/sec-0 F) 176 179.375 Service Water Temperature (1F) 85 85 High Dome Pressure ATWS-RPT Setpoint 1150 1150 (psig)_
SRV Capacity - per valve (Ibm/hr) / Reference 800,000 / 1080 800,000 / 1080 Pressure (psig) 1-6
BVY 04-026 / Attachment I Parameter Rcfcrcncc 1 Revised Analysis Analysis SSV Capacity - per valve (Ibm/hr) / Reference 932,500 / 1240 1,126,200(a) / 1240 Pressure (psig) II SRV / SSV Configuration 4 / 2(b) 3/3 Notes:
(a) The SSV capacity is increased by the plant modification planned for the upcoming refueling outage (scheduled for spring 2004), which is the plant refueling outage for ARTS/MELLLA implementation. The increase in SSV capacity supports future changes to VYNPS' licensing basis, e.g., Maximum Extended Load Line Limit Analysis Plus (MELLLA+).
(b) Currently, VYNPS has four SRVs and two SSVs. Prior to the implementation of MELLLA, an additional SSV will be installed with the capacity listed above.
The Reference I ATWS analysis assumed a four SRV / two SSV configuration, which is shown through evaluation to be representative of a three SRV / three SSV alignment, i.e., one SRVOOS.
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BVY 04-026 / Attachment I Table 1-2 Summary of Key Parameters for Short-term ATWS Calculation Reference Ana ysis . R Reference I Analysis MSIVC PRFO BOC EG BOG EOC-Peak Vessel Bottom 1360 1357 1367 1362 Pressure (psig)
Time of Peak Vessel 10.4 9.9 38.9 42.2 Pressure (sec)
Peak Neutron Flux (% 210 224 224 696 rated)
Time of Peak Neutron 4.1 4.1 29.9 33.0 Flux (sec)
Peak Vessel Heat Flux 127 129 135 139
(% rated)
Time of Peak Heat 5.8 5.2 34.0 36.8 Flux (see)
Reised Analysis Revised Analysis
.MSIVC PRFO-.
BOC EGG BOC EOC Peak Vessel Bottom 1353 1352 1358 1353 Pressure (psig)
Time of Peak Vessel 8.7 8.2 45.4 42.6 Pressure (sec)
Peak Neutron Flux (% 210 238 223 334 rated)
Time of Peak Neutron 4.1 4.1 41.1 34.9 Flux (sec)
Peak Vessel Heat Flux 131 133 142 143
(% rated)
Time of Peak Heat 5.8 6.1 42.6 40.1 Flux (sec) 1-8
BVY 04-026 / Attachment I Table 1-3 Summary of Key Parameters for Long-term ATWNS Calculation
-;-Rcfercnce 1 Analysis - :Refcrnc 1 Analysis MSIVC - : PRFO
.BO h. EOC BOC EOG Peak Suppression Pool 170 183 172 182 Temperature (0F)
Time of Peak 1259 1485 1672 1574 Temperature (sec)
Peak Containment 8.5 11.1 8.8 10.8 Pressure (psig)
Time of Peak Pressure 1259 1485 1672 1574 (sec)
Rcvised Analysis Revised Analysis MSIVC . PRFO BOC : EOC BOC EG Peak Suppression Pool 175 185 175 184 Temperature ( 0F) l Time of Peak 1349 1657 1362 1731 Temperature (sec)
Peak Containment 9.4 11.5 9.4 11.3 Pressure (psig) I Time of Peak Pressure 1568 1657 1587 1731 (sec) II 1-9
BVY 04-026 / Attachment I Table 1-4 ATWVS Calculation Sequence of Events (Time in seconds)
Ent -. Reference 1 Analysis Reference i Analysis
- MSIVC PRFO BOG ' EGG. BOC EOC Turbine control and bypass valves - - 0.1 0.1 start open MSIV isolation initiation 0.0 0.0 24.7 28.9 MSIVs closed 4.0 4.0 28.7 32.9 High pressure ATWS setpoint 4.5 4.5 29.9 33.0 Opening of the first relief valve 4.5 4.5 32.8 36.5 Recirculation pumps tripped 5.0 5.0 33.2 37.0 Boron injection initiation 80 80 108 112 temperature achieved SLCS pumps start 124.5 124.5 152.7 156.4 Boron solution reaches lower 157.8 157.8 186.0 189.7 plenum . -
RHR cooling initiated 660 660 660 660 Hot shutdown achieved 1431 1650 1569 1740 Event : Revised-Analysis - Revised Analysis-
--:.- MSIVC^ - PRFO
_:. _ - _- _ - _ - - _: _ BOC E. B E GOCG.
Turbine control and bypass valves - 0.1 0.1 start open MSIV isolation initiation 0.0 0.0 33.8 31.0 MSIVs closed 4.0 4.0 37.8 35.0 High pressure ATWS setpoint 4.7 4.7 41.5 38.7 Opening of the first relief valve 4.6 4.6 41.4 38.7 Recirculation pumps tripped 5.2 5.2 42.0 39.3 Boron injection initiation 105 105 142 139 temperature achieved SLCS pumps start 124.7 124.7 161.5 158.7 Boron solution reaches lower 169.6 169.6 206.4 203.6 plenum RHR cooling initiated 660 660 660 660 Hot shutdown achieved 1435 1760 1425 1779 1-10
BVY 04-026 / Attachment I
2.0 REFERENCES
Technical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA)," NEDC-33089P, March 2003, (Attachment to Letter, BVY 03-23, VY to USNRC, "Technical Specification Proposed Change No. 257, Implementation of ARTS/MELLLA at Vermont Yankee," dated March 20, 2003).
- 2. NEDC-24154P-A, "Qualification of the One Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4)", February 2000.
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