ML051290321
| ML051290321 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 05/08/2005 |
| From: | Barry Miller Entergy Operations |
| To: | Alexion T Office of Nuclear Reactor Regulation |
| References | |
| Download: ML051290321 (6) | |
Text
Thomas Alexion - Waterford 3 Draft Additiona Information Regarding Instrument Uncertainty Page 1 From: "MILLER, D BRYAN <dmil14@entergy.com>
To: "'Thomas Alexion'" <TWA nrc.gov>
Date: 5/8/05 3:16PM
Subject:
Waterford 3 Draft Additional Information Regarding Instrument Uncertainty
<<Draft for NRC 5-8-05.pdf>>
Tom, The information we added as a result of our conference call last week is identified by revision bars in the attached document. Let me know if this addresses Kent's remaining questions.
Thanks, Bryan CC: "KALYANAM, N. KALY" <nxk~nrc.gov>
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Subject:
Waterford 3 Draft Additional Information Regarding Instrument Uncertainty Creation Date: 5/8/05 3:16PM From: "MILLER, DlBRYAN" <dmilll4@entergy.comn>
Created By: dmill14@entergy.com Recipients nrc.gov owf4_po.0WFNDO NXK CC (N. Kaly Kalyanam)
TWA (Thomas Alexion)
Post Office Route owf4_po.0WFNDO nrc.gov Files Size Date & Time NMSSAGE 237 05/08/05 03:16PM TEXT.htm 877 Draft for NRC 5-8-05 .pdf 28570 Mime.822 42223 Options Expiration Date: None Priority: Standard Reply Requested: No Return Notification: None Conceal~ed
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Technical Specification 3.2.6 Footnote
- to Technical Specification 3.2.6 allows the upper limit on Tcold to increase to 5590F for up to 30 minutes following a reactor power cutback in which (1) regulating groups 5 and/or 6 are dropped or (2) regulating groups 5 and/or 6 are dropped and the remaining regulating groups are sequentially inserted. This value is considered an arbitrary value to which uncertainty need not be applied. Footnote
- to Technical Specification 3.2.6 was in the Waterford 3 Technical Specifications at the time of initial licensing. There has been no change to that footnote between initial licensing and EPU.
A reactor power cutback is a non-safety system which is initiate the e of a load rejection, such as a turbine trip or a loss of one of two main feedwat p exceeds the capacity of the turbine bypass valves. The reactor power cutbac lys em he dropping of one or more preselected CEA groups. This rapid reduction o t pow te which is greater than that provided by the normal high speed CEA ins n, he rgoEhe plant to within prescribed operating ranges. Reactor power cutback i e d in a iRCh 1ft 15 safety analyses.
The reactor power cutback system will also throttle the turbin a ission valve (for a loss of a feedwater pump) to rebalance turbine and reactor ower. If th minor mismatch and core power is greater than turbine demand, cold le eure will s crease. With a negative MTC, the increasing temperature s rower to r se to match the turbine demand, resulting in a stable pow he rebel set byhe turbine. Since this power is substantially below full power, t er is n ien ermal margins.
Reactor power cutback will occur eponse t aoss of pe event, such as those documented in FSAR Section 15 cluding th ss of nal Load and Turbine Trip events.
These events, categorized a ases in He y the Secondary System, are events in whic actor po C ack is utilized reduce RCS power to avoid reactor trips which w o erwise re high pressurizer pressure. The response of Waterford 3 to the limiting FSA nticipat Mirational Occurrence, the Loss of Condenser Vacuum (without credit-for ower c is documented in the EPU License Amendment Request of lett vember 13, 2003. As can be seen per Figure 2.13.2.1.3-1 LOC vent whi oes not challenge DNBR margins. As discussed in the PUR an he Waterfo SAR, the response to a Loss of External Load or a Turbine Trip is boun bythe resp o Loss of Condenser Vacuum. The responses to those events with e ctor powerc present would also be bounded by the Loss of Condenser Vacuum anal s The loss of one femp is an event which is bounded by the total Loss of Feedwater event (without credit for reactor power cutback) which is documented in FSAR Section 15.2.2.5 and in Section 2.13.2.2.5 of the 3716 MWt Extended Power Uprate License Amendment Request, W3F1-2003-0074. As shown in Figure 2.13.2.2.5-12 of the EPU LAR, this event also does not challenge DNBR, with a relatively constant DNBR prior to reactor trip.
Thus, the resultant core conditions after a reactor power cutback would result in no challenge to thermal margins.
Control system analyses conducted in support of Waterford 3 EPU have modeled the plant response to transients involving reactor power cutback. For example, reactor power cutback would result in a core power of about 50% for an End of Cycle (EOC) Turbine Trip. With no 5/8/2005, 2:09 PM Page 1 of 4
operator action to drive in additional CEA's, there would be about a 70F rise in Tcold from a nominal 5430 F to about 550'F. This is consistent with the 100 F rise allowed per the footnote to TS 3.2.6.
As stated in Attachment 1 to W3F1-2003-0074, the 3716 MWt Extended Power Uprate License Amendment Request, this value is being revised from 568OF to 5590 F for EPU, in conjunction with the change to the Tcold LCO; the LCO is being revised from a range of 541'F to 5580 F to a new range of 5360 F to 5490F. The revision of this value to 5590 F maintains the existing 100 F difference to the maximum Tc0 ld.
Waterford 3 was licensed with a Tcold range of 541OF to 558 0F, based on a nominal temperature ramp from 545OF at Hot Zero Power (HZP) to 5530 F t Hot F ower (HFP).
Under 10CFR50.59, Waterford 3 revised this nominal tem rare o o a constant 5450 F value in the early 1990's. For power uprate, a 20F ram ne A, with nominal Tcold 0
ranging from 541OF at HZP to 543 F at HFP. Thus, wi yeimpl iin of EPU, there will be a more restrictive differentialange of 16'F (559 0 F eus 5 normal Tcold andto the footnote value compared to the pre-EPU dife ti e of 2 8F 5450 F).
The original 5680 F value in Technical Specifications wa arb chosen to be 10OF above the upper limit of the LCO, on the basis that it is reasonable to alloo deviation for a short period of time (30 minutes) to allow recovery bs uent pla ization after the reactor power cutback. This also prevents unnece ae s into T idaI Specification ACTION statements. The 10F offset is c nged r Wt Po rUprate.
Operators select the appropriate CE gro p(s) t op duri actor power cutback. The selection ensures that the reactor r followi utback I less than the capacity of the turbine bypass valves of about 6 ecause eactor p r cutback is a plant transient of short duration, no additional c t or transie d to occur simultaneously during the 30 minute ift period o e 3.2.6 footnot cold may be above the explicitly analyzed rang A , due to .duced power, there is significantly less energy and latent heat in the react after the k.
The Core Prot or (CP initiates automatic protective action to assure that the specified ptable e sign limi DNBR and LPD are not exceeded. The Low DNBR and I LPD trips d by CPC are discussed in FSAR Section 7.2.
The CPC W ange Tcold tends from 4950F to 580OF, as documented in Bases Section 2.0 o aford 3 T ical Specifications. CPC will produce conservative calculations of DNBR and L oues within the Wide Range band, and would be capable of fulfilling their funci ang a reactor trip when needed. As stated in Technical Specification Bases, the DNBR algorithm used in CPC is valid within these Wide Range limits and operation outside of these limits will result in CPC initiating an Auxiliary trip for the parameter being out of range. Thus the CPCs would continue to adequately protect the core during the temporary 30 minute Tcold excursion to 5590 F following the reactor power cutback which is well within the CPC Wide Range upper limit of 5800F (i.e., 21 'F below the upper limit).
Note that pre-EPU Technical Specifications allowed a temporary 30 minute Tcold excursion to 5680 F (i.e., 120F below the upper limit) following a reactor power cutback.
Many Chapter 15 analyses are conducted assuming initial conditions corresponding to Power Operating Limits, that is, assuming that there is no initial excess margin preserved. For such analyses, there is little impact associated with any specific initial condition, and the initiation of 5/8/2005, 2:09 PM Page 2 of 4
an analysis in the slightly extended indicated temperature range of 5490F to 5590 F associated with the footnote to TS 3.2.6 would have no appreciable effect. While some analyses assume a high bias for the initial Tcold, this is generally done to maximize the core initial energy for the transient; however, the thermal margin gains associated with the lower core power after a reactor power cutback would dominate any small impact due to an initial core temperature which is a couple of degrees higher. Thus, engineering judgment leads to the conclusion, based on the operation of the CPCs, increased thermal margins, and conservatisms in the analysis, that a Tcold up to 5580F for 30 minutes following a reactor power cutback is acceptable.
Further, since the plant would have already experienced a Chapter 15 event initiator (e.g., a Loss of Load or a partial Loss of Feedwater Flow) prior to the re tor po putback, it is not necessary or credible to postulate another event happenin du ue ed period of time (30 minutes) that the TS 3.2.6 footnote would be applicable e re wer cutback.
Because the 5590F value for the TS 3.2.6 footnote ap ed v e t 199 is not based on a specific analysis but is intended as a reasonable or the H I Tcol temperature swing following a reactor power cutback, t e o need to a ent uncertainty with respect to this parameter. This param e sidered a C item.
5/8/2005, 2:09 PM Page 3 of 4
Power Level for OPERABILITY of ADV Automatic Actuation Technical Specification 3.7.1.7 New Technical Specification 3.7.1.7 is being added due to EPU to specify OPERABILITY requirements for the Atmospheric Dump Valves. This TS is being added since the EPU Small Break LOCA Emergency Core Cooling System (ECCS) analysis credits one Atmospheric Dump Valve for the purpose of secondary pressure control; the ADV's were previously credited only for cooldown to shutdown cooling entry conditions and for their containment isolation function.
The small break LOCA analyses assume a maximum ADV setpoint of 1040 psia. This value is specified in the footnote to TS 3.7.1.7 and explicitly accounts for the instrument uncertainty offset from the nominal setpoint of 1007 psia.
The footnote to the LCO also documents that the ADV oatic channels are not required to be operable when the reactor has been at ishan o 70% Rated Thermal Power for greater than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (following long-term oation a Thermal Power of 3716 MWt). The value of 70% is specified based on r as Inginee dgme a power level below which automatic actuation of the AD i t required.
acceptability of this arbitrary value, a calculation was p eg t demonstra t e decay heat load associated with operation for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at 70% ate r al Power is such that the ADV's need not be credited to demonstrate acceptable ECCS o ance. The ADV's are not credited in the Waterford 3 Cycle 13 pre-uprat eak LOC E S analyses, which leads to the conclusion that long-term opert e evels of 3 W t (92.6% of EPU Rated Thermal Power) is acceptable with tediti = n the SB CA analysis. The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> time frame supports the Bases for GiIONs db TS 3.7.1.7, which calls for exiting TS applicability within 6 hourafte edu power teshan or equal to 70% of Rated Thermal Power.
Margin exists in the decay h a lysis betwe h -p ate power where ADV's are not required (e.g., dh term op at 3441 MWt) ecay heat corresponding to operation at 70% of uprA rmal po 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or less. The decay heat for a reactor trip after operation for six t 70% o f d thermal power after long-term operation at 3716 MWt is around 10% bel from Ion a operation at 3441 MWt. A strict analytical approach would result in easing Thermal Power as a function of time, that is, the reactor pow d be s crease o approximately 92.6% in order for this decay heat logic to be a ained. Ien ration of this margin and the fact that the decay heat load associated i 70% power o r will decrease with longer times, it is not considered necessary t ly any explic f to account for power measurement uncertainty to the 70%
value specifie nca fications.
Based upon this rin ntergy considers this to be an arbitrary value to which uncertainty need not be applied and therefore a Category D parameter. If explicit analysis were performed this value could be raised to a value closer to 92.6%.
5/8/2005, 2:09 PM Page 4 of 4