ML20046B393
| ML20046B393 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 07/23/1993 |
| From: | Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LD-93-111, NUDOCS 9308040154 | |
| Download: ML20046B393 (72) | |
Text
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A ED ED 7%E9E9 ASEA BROWN BOVERI ,
July .23, 1993 LD-93-111 Docket No.52-002 ,
U. S. Nucear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555
Subject:
Response to NRC Requests for Additional Information
References:
- 1. FAX May 17, 1993, M. Franovich (NRC) to J. Longo (ABB-CE)
- 2. FAX June 8, 1993, T. Wambach (NRC) to F.
Carpentino
Dear Sirs:
The References requested additional information for the NRC staff review of the Combustion Engineering Standard Safety Analysis Report -- Design Certification (CESSAR-DC). Enclosure 1 to this letter provides our responses to all of these questions with the exception of RAIs 440.209 and 440.224. RAI 440.209 will be provided shortly under separate cover. RAI 440.224 has previously '
been submitted to the NRC.
Should you have any questions on the enclosed material, please contact me or Mr. Stan Ritterbusch of my staff at (203) 285-5206.
Very truly yours, COMBUSTION ENGINEERING, INC.
C. B. Brinkman Af Acting Director Nuclear Systems Licensing
Enclosures:
As Stated cc: J. Trotter (EPRI) .
T. Wambach (NRC)
) ,
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Enclosure I to A LD-93-111 Ouestion 440.197 Provide the revised TSs that reflect the assumptions and results of Chapters 6 and 15 reanalysis.
l Response 440.197 The Systo 80+" Technical Specifications have been updated to reflect j esulting from the revised safety analysis. These updates '
adjustmero primarily ,c alved adjustments to LCOs rather than changes which affected a fundamental basis for having the Technical Specification. As a general rule, numerical values stated for LCOs in the Technical Specification are representative of the values that would appear in a site specific set of Technical Specifications. This is indicated in the CESSAR DC Chapter 16 ;
Technical Specifications by placing a bracket around such values or !
conditions. The actual values that would be included in a Licensee's l Technical Specifications would reflect the specific equipment procured and installed, uncertainties for instrumentation, calibration, equipment uncertainties, among others. The attached Table 440.197-1 provides a list of the Technical Specifications that were revised to reflect Chapters 6 :
and 15 reanalyses. These changes will be implemented in future revisions to CESSAR-DC.
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TABLE 440.197-1 Technical Specification CESSAR-DC LCO Paragraph No. Parameter Old New Source LCO 3.1.4 Moderator Temperature 0.0x10-4 -0.1x10-4 Table 15.3.1-2 Coefficfent-HFP
( A Q / F)
LCO 3.1.11 Number of Charging Pumps 2 1 Section 15.4.6 Operating LC0 3.4.1 a.) Pressurizer .
1905- 2175- Table 15.0-3 Pressure Range (psia) 2375 2325 j T
cold Range ( F): .
LC0 3.4.1 b.) < 90% Power 543- 543- Table 15.0-3 I 4
565 561 LCO 3.4.1 b.) 90%-100% Power 553- 550- Table 15.0-3 )
563 561 1 Rx Trio Setpoints:
LCO 3.3.1 High Pressurizer Pressure (psia) 2445 2434 Table 15.0-3 LOCO 3.3.5 CPC Coincident ----
2015 Table 15.0-3 Low Pressure /DNBR (psia)
Table 15.0-3 Limits
- NOTE: The RCS Pressure and Temperature Limit Curves were revised to reflect the 3% power increase j and the initial condition space.
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i Ouestion 440.198:
Figures 6.3.3.2-11 and 6.3.3.3-12 Provide the results of sensitivity study for the break size spectrum to show that the ljmiting cases are the DEG/PD for large break LOCAs and the break of 0.1 ft for small break LOCAs.
Response 440.198:
TheresultsofsensitivitystudiesthatdemonstratethatthelimitiggLB LOCA is the 1.0 DEG/PD break and the limiting SB LOCA is the 0.1 ft DVI line break are shown in Figures 6.3.3.2-11 and 6.3.3.3-12.
As shown in Figure 6.3.3.2-11, the break size spectrum analysis performed at 3876 MWt identified the 1.0 DEG/PD break as the limiting break for the System 80+ LB LOCA analysis. The limiting LB LOCA break size is determined by the thermal-hydraulic response of the core during the blowdown portion of the LB LOCA transient. The differences in the System 80+ design in going from 3876 MWt to 3992 MWt do not change the thermal-hydraulic response of the core sufficiently to cause a change in the limiting break size. This is shown by a comparison of CESSAR-DC Figures 6.3.3.2-5A through SE to Figures 6.3.3.2-10A through 10E. Therefore, the 1.0 DEG/PD break is also the limiting LB LOCA at 3992 MWt.
As shown in Figure 6.3.3.3-12, the break size spectrum analysis performed at 3876 MWt identified the DVI line as the limiting break location for the System 80+ SB LOCA analysis. The major reason the DVI line is the i
limiting break location is that, in conjunction with a diesel generator failure, it results in the minimum amount of SI pump flow reaching the core (100% of the flow from one SI pump). For a break in the RCS hot or cold legs,100% of the flow from two SI pumps reaches the core. (See CESSAR-DC Section 6.3.3.3.2).
Three DVI line breaks (0.05 ft2 , 0.08 ft z, and 0.1 ft )z-were analyzed at a core power of 3992 MWt in order to determine the limi The limiting breag size was determined to be the break.0.( Breaks ft}ing break size.
larger than 0.1 ft are less limiting than the 0.1 ft break because they are large enough such that the RCS will depressurize to the point that SITS come on and reflood the core befope the peak cladding temperature can exceed that calculated for the 0.1 ft break.
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Question 440.199:
Table 6.3.3.2-2 Why the core and the RCS flow rates are assumed to be the same?
Response 440.199:
The core and RCS flow rates are not the sage in Table 6.3.3.2-2. For example, the initial RCS flow rate is 165.8x10 lbm/hr and the initial core flow rate is 160.8x106lbm/hr for the analysis at 3992 MWt.
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Tables 6.3.3.2-2 and 6.3.3.3-2 Why the initial core outlet temperatures are different for LB and SB.
LOCAs? .
Response 440.200:
The LB LOCA blowdown analysis explicitly models core bypass flow, whereas, the SB LOCA blowdown analysis does not. Consequently, the initial core outlet temperature listed in Table 6.3.3.2-2 for LB LOCA is based on the _.
actual core flow rate of 160.8x10' lbm/hr while the initial core outlet temperature listed in Table 6.3.3.3-2 for SB LOCA is based on the RCS flow 6
rate of 165.8x10 lbm/hr. Since the SB LOCA blowdown analysis uses a higher initial core flow rate (and has the same initial core power and inlet temperature), it has a lower initial core outlet temperature. i i
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Question 440.201: l 1
Table 6.3.3.3-2 ;
What is the limiting burnup resulting in a highest PCT for SB LOCAs? <
Response 440.201:
As prescribed on page 18 of CENPD-137P, " Calculative Methods for the C-E Small Break LOCA Evaluation Model," August 1974 (Reference 3 of Section 6.3.3 of CESSAR-DC), the SB LOCA analysis is performed for the burnup "at which the initial stored energy in the fuel is highest." For the CESSAR-DC SB LOCA analysis, this corresponds to a hot rod burnup of 1000 MWD /MTU.
Furthermore, the peak cladding temperature for the limiting SB LOCA is ,
driven by decay heat and is not sensitive to the initial stored energy in <
the fuel. -
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Question 440.202:
Assumption I, page 6.3-35 Provide the basis for the boric acid precipitation of 27.6% at the containment pressure of 14.7 psia. ;
Response 440.202: !
The boric acid precipitation of 27.6 wt% at a pressure of 14.7 psia was !
obtained from Figure C-3, page 4. Amendment 1 of CENPD-254-P-A, " Post-LOCA ;
Long Term Cooling Evaluation Model," June 1980 (Reference 9 of Section t 6.3.3 of CESSAR-DC). j P
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l Ouestion 440.203: 1 Table 6.3.3.4-1 :
Provide an explanation for the meaning of the note for SCS entry conditions (temperature and pressure) with consideration of the associated instrumentation errors. ;
Resoonse 440.203: }
r The values of 360 F and 330 psia are the minimum values for the actual hot !
leg temperature and pressurizer pressure, respectively, when the control room instrumentation is indicating the maximum allowable post-accident values for entry into shutdown cooling, namely, 380 F and 400 psia. The ,
values of 360 F and 330 psia are based on post-accident instrument errors ,
of 20 F and 170 psi, respectively.
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Ouestion 440.204:
Figure 6.3.3.4-5 Provide analytical results to show that for small break LOCAs with break sizes smaller than 0.01 ft ,z the SCS entry conditions can be achieved in '
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to support the resolution of Open Items 6.3.3-1 (see C-E letter, LD-93-048, dated March 17,1993). ,
Response 440.204: 1 As part of the post-LOCA long term cooling analysis that is described in Section 6.3.3.4 of CESSAR-DC, the NATFLOW and CEPAC computer programs are !
used to calculate a natural circulation cooldown of the RCS. It is t calculated to determine the earliest time that the SCS entry temperature is reached following a LOCA. The analysis simulates a conservatively slow cooldown rate and, consequently, a maximum value for the earliest time that the SCS entry temperature is reached. For example, the analysis assumes a 0.0 ftj break size (i.e., no RCS energy goes out the break; it -
is all removed via the steam generators) and that only one atmospheric ;
dump valve per steam generator is available. Al so, steam generator ,
cooldown is assumed to begin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the start of the LOCA.
The natural circulation cooldown analysis determined that the SCS entry ,
temperature of 360 F is reached at approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the start !
of the LOCA. (As stated in the response to Question 440.203, 360*F !
corresponds to the minimum actual hot leg temperature when the-indicated hot leg temperature is 380 F, the maximum allowable indicated post-accident '
hot leg temperature for entry into shutdown cooling.) Therefore, the analysis demonstrates that the SCS entry temperature can be achieved well before 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after the start of a LOCA. In addition, reaching the SCS entry temperature at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> leaves ample time for the operator to j depressurize the RCS to the SCS entry pressure and initiate shutdown cooling. !
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i Ouestion 440.205:
RCS Flow Rates i The design RCS flow rate is 444,650 gpm stated in Table 4.4-1. The design .
RCS flow used for the transient analysis is 445,600 gpm as stated in Table 15.0-3. Clarify this discrepancy and revise the transient reanalysis ,
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reflecting the correct RCS flow rate. .
Response 440.205:
Table 15.0-3 identifies a range of reactor vessel flows to be used for !
initial conditions for the safety analyses. The flows range from 423,320 gpm to 516,896 gpm or approximately 95% to 116% of the design flowrate of ,
444,650 gpm. since the design RCS flow rate falls well within the range l considered in the safety analyses, there is no discrepancy. 1 t
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s Ouestion 440.206: ;
The initial DNBRs for the transients are significantly higher for the -
reanalyzed cases than that of existing cases. For example, the initial :
DNBR for the inadvertent opening of a SG ADV increased from 1.24 to 1.36, l while the rated power increased from 3800 to 3914 MWt and the maximum radial peaking factor increased from 1.4 to 1,44 (Table 15.1.4-3), which could decrease the initial DNBR. The applicant is required to provide the 3 technical basis for the significant increase in initial DNBRs for all the ;
events reanalyzed. j Response 440.206:
The initial DNBRs for the System 80+ power upgrade reanalysis are higher than those of the original analyses (Amendment H) due to a change in the !
overpower margin reserved for the analyses. This Required Overpower l Margin (ROPM) represents the distance, in units of power, from the ;
specified acceptable fuel design limit (SAFDL). A higher R0PM results in j an initial DNBR that is further from the SAFDL and is thus higher. !
The DNBRs stated in the question for the Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (IS0GADV) event are not the initial DNBRs ,
but are the DNBRs at 30 minutes into the event. The reason these two '
ONBRs are different is that a change in the approach used to select the initial conditions was applied for the reanalysis. In the original analysis a DNBR of 1.24 was forced to occur at 30 minutes as this was the minimum DNBR which could occur prior to a CPC reactor trip. This approach ;
is overly conservative since the initial thermal margin which would need :
i to be assumed for the DNBR to be 1.24 at 30 minutes would be less than that reserved by the R0PM. That is, the plant would have to have been operating in violation of a Power Operating Limit (POL) prior to event i initiation (see the response to RAI 440.215). The reanalysis assumed the plant to be initially (time =0) operating at a POL. This resulted in a DNBR of 1.36 at 30 minutes into the event. ]
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4 Ouestions 440.207:
Steam Line Break The control rod worth of 10% was used in the SLB analysis. This value is significant increased (from 8.86%). provide the basis for thE increase.
Clarify that whether the change is due to the control rod design change or the calculational method change for the control rod worth.
Response 440.207:
The change in CEA worth used in the steam line break (SLB) analyses froin 8.86%4/ to 10%4/ results from an analysis of the apportionment of conservative margins to the values of the reactivity components of the SLB analyses for potential post-trip return to power. Figure 15.1.5-0 of CESSAR-DC is useful in understanding this. As a reference point, a CEA worth of 10%df used together with the moderator cooldown function represented by the lower curve of Figure 15.1.5-0 results in post-trip reactivity values which are typical of those appropriate to SLB analyses (assuming the usual conservatisms such as end of fuel cycle). Either decreasing the CEA worth in conjunction with this curve or using a CEA worth of 10%4f together with a more adverse (more negative slope) curve will increase the conservatism of the analysis results. The CEA worth of 10%Af used together with the moderator cooldown function employed in the SLB analyses for CESSAR-DC (upper curve of Figure 15.1.5-0) yields results which are appreciably more conservative than the most adverse expected for System 80+. Employing both a value of 8.86%Af CEA worth and the upper curve would produce an extreme, excessively conservative bound for the expected post-trip reactivity change.
The comparison of available CEA worths and allowances presented in the attached Table 4.3-7 of CESSAR-DC has been updated based on ROCS /DIT calculations for an equilibrium 18 month cycle core design for System 80+.
The results of these calculations show that the worth available from the CEAs with all CEAs inserted except the most reactive CEA is 10.7%4/. The net CEA worth of 10.7%4fis the difference between the total worth of CEAs inserted (15.3%A/) and the stuck rod worth (4.6%A/). The calculated net CEA worth exceeds the 10% 4/ CEA worth assumed in the SLB analyses.
Therefore, the CEA worth used in the SLB analysis is conservative with respect to the actual reactivity avaliable from the CEAs.
CESSAR nainemo, TABLE 4.3-7 COMPARISON OF AVAILABLE CEA WORTilS AND ALLOWANCES Reactivity Condition (%Ap) ,
All CEAs inserted, hot, 587'F /5 f :f*M Total reactivity allowance, full power 7.7 (from Table 4.3-6)
Y. 6 -.Mr Stuck rod worth Excess, assuming most adverse combination of uncertainties J, e - -h=t.
A2nendment N April 1, 1993
- t Question 440.208:
?
1 Steam Line Break--AFW Actuation (Page 15.1-15) ;
The staff agrees that an early actuation of the AFW will maximize the ,
3 cooldown effect and is a conservative assumption for the post-trip SLB ;
analysis. For an SLB with a loss of offsite power assuming failure of an ;
MSIV in the intact SG to close, a delay of AFW actuation could result in a complete depletion of the water inventory from both SGs. Under this -l condition, the injection of cold AFW could cause significant thermal '
stress on SGs and result in a further damage to SGs. The applicant is requested to provide an analysis showing that a delay of AFW actuation l (such as AFW on an automatic mode) will not result in a complete depletion of inventory from both SGs and a complete loss of SG heat removal 1 l 3
capability due to the thermal stress. The long term cooling capability :
with sufficient AFW resource should be demonstrated for the SLB with !
blowdown from both SGs.
Response 440.208: l.
3 for a steam line break (SLB) with an assumed failure to close of a main
! steam isolation valve (MSIV) in the intact steam generator, a delay of i
- emergency feedwater actuation may result in a complete depletion of the ;
liquid inventory from both steam generators. A SLB with conditions chosen ;
- to minimize the liquid remaining in the steam generators at the time emergency feedwater reaches the steam generators was analyzed. The
, initial steam generator water inventory was minimized and the maximum time delay for emergency feedwater delivery (60 seconds) was used. For this i
case there was still more than 25,000 lbm of liquid in each steam generator when the emergency feedwater reached the steam generator. Steam generator dryout occurred about 50 seconds later. Cooldown then continued with emergency feedwater flow to the dry steam generators. All incoming feedwater proceeds to be boiled off until the primary side temperatures are reduced to a value at which the primary to secondary heat transfer rate is less than that extracted by the boiling of all feedwater at approximately 20 minutes into the transient. After this time, the steam j generators begin to refill.
The greatest potential for large temperature gradients and consequently large thermal stresses is at the tubesheet at the time of steam generator dryout. The primary coolant temperature is approximately 400 F at this time. A bounding l
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Response 440.208: (continued) calculation' shows that the temperature of the emergency feedwater impinging on the tubesheet is greater than 330*F immediately following steam generator dryout. The primary to secondary temperature difference ,
at this time is, therefore, not greater than 70*F. This is less than the normal operating temperature difference on the cold leg or economizer side .
(106*F). Therefore, the stresses under this dryout condition are less than !
for normal operating conditions and are acceptable. Hence the heat i removal capability of the steam generators during postulated SLB transients would be preserved.
The analysis presented in Appendix 5D of CESSAR-DC demonstrates that ,
240,000 gallons of emergency feedwater is sufficient to achieve shutdown cooling entry conditions following reactor / turbine trip from full power. L conditions. Shutdown cooling entry conditions are shown to be achieved 9.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> after event initiation in Appendix 50. The analysis of )
Appendix SD uses a set of assumptions which maximizes the time needed to -l achieve shutdown cooling and the amount of emergency feedwater required. .
A cooldown starting from one half hour after the initiation of a postulated SLB event would involve removing less decay heat and less sensible heat than a cooldown starting following reactor / turbine trip from full power conditions and would take less than 9.75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br /> calculated i for the analysis of Appendix SD. A cooldown started 30 minutes after !
initiation of a SLB would, therefore, require less than 240,000 gallons of l emergency feedwater. The minimum available emergency feedwater capacity l for the System 80+ design is 700,000 gallons. If it were to be assumed ;
that the total maximum emergency feedwater flow of 800 gpm per steam ;
generator made no contribution to the cooldown during the first half hour !
of the SLB event, there would still be more than 650,000 gallons left to l affect plant cooldown at that point. Since less than 240,000 gallons of i emergency feedwater are required for cooldown after 30 minutes, the j emergency feedwater supply is ample for long term cooling capability for i a SLB with blowdown from both steam generators. ;
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' This calculation included only the effect of condensation of steam on an I undispersed stream of feedwater (conservatively assumed to enter the steam i generator at 40*F) falling through the downcomer. Other effects which
- would result in greater heating of the feedwater (such as heating of the feedwater from the vessel shell and increase in the heat transfer area of :
the fluid stream due to dispersal upon impact with the tube bundle shroud) !
were neglected.
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i Question 440.210 Explain why the calculated DNBR decreases twice before it increases monotonously.
Response 440.210 The DNBR transient of Figure 15.2.3-13 resulted from a separate parametric analysis which sought a minimum DNBR rather than a peak pressure. The parameters selected which minimized the DNBR resulted in a rapid pressurizer pressure response which opened the primary safety valves twice before the reactor was tripped on the second pressure rise. The core power and RCS pressure transients are shown in the attached figures.
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Figure 440.210-1 SYSTEM 80+ LOSS OF CONDENSER VACUUM i CORE POWER VS. TIME i i
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- i 0.0 10.0 20.0 30.0 40.0 50.0 TIME, SECONDS
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SYSTEM 80+ LOSS OF CONDENSER VACULH !
f RCS PRESSURE VS. TIME L r
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'l Ouestion 440.211-
- i FLB- Calculated Peak RCS Pressure i The calculated peak RCS pressure is 2785 psia on page 15.2-15 and is 2720 l psia in Table 15.2.8-2. Clarify the inconsistency.
Response 440.211:
The feedwater line break event has been reanalyzed and this section will i be updated in the next Amendment of CESSAR-DC. The response to RAI i 440.214 summarizes the reanalysis results for calculated peak pressures. .
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Ouestion 440.212: f FLB-DNBR f It is stated on page 15.2-16 that thg MDNBR is below 1.24 for the limiting FLB with a break size of 0.2 ft . The applicant is requested to l provide the calculated DNBRs as a function of time for the limiting FLB ;
cases. ,
Response 440.212:
A figure of minimum DNBR versus, time for the limiting feedwater line break for fuel performance, a 0.2 ft break, is attached as Figure 15.2.8-17.
- This figure represents the results of a revised analysis which will be included in the next Amendment to CESSAR-DC.
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July 15,'1993 l l FEEDWATER LINE BREAK Figure '
Jg LIMITING CASE ;
MINIMUM DNBR vs TIME 15.2.8-17 l
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Ouestion 440.213 ,
i Table 15.2.8-1: initial conditions Provide the criterion for selecting initial conditions in order to f maximize the peak RCS pressure resulting from the limiting FLB case. ,
i Response 440.213 The feedwater line break (FWLB) event, presented in Amendment N has been reanalyzed. The results of this revised analysis will be presented in the next Amendment to CESSAR-DC. In this revised analysis the initial :
conditions for maximizing the peak pressure for the limiting FWLB event result from extensive parametric variations of all system initial i conditions within their respective operating ranges. During the FWLB ,
event, the peak pressure is very sensitive to the timing of the high l pressurizer pressure trip (HPPT) with respect to the emptying of the affected steam generator (which results in a rapid loss of heat transfer) and the opening of the main steam safety valves (MSSVs). The initial conditions having a dominant influence on the peak pressure are the pressurizer pressure, the pressurizer water level, and the steam generator inventory. The peak pressure is generally obtained when the HPPT occurs ,
slightly (about I second) before the affected steam generator is empty, !
which would also allow a reactor trip. Peak primary pressure is increased when the MSSVs open after the affected steam generator is empty. The most ,
adverse timing, and the limiting peak pressure, is obtained by manipulation of the three aforementioned initial conditions. This i procedure is followed with the initial conditions of the other system ;
parameters in their most adverse' position in range. [
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6 Question 440.214:
Figure 15.2.8-1 :
ExplainwhatcausesthepeakRCSppessuredecreasingandthenincreasing for break sizes from 0.7 to 0.9 ft Response 440.214: !
The feedwater line break event has been reanalyzed and a revised Figure 15.2.8-1 is attached. The figure gives the peak pressure vs. break area, and is the same overall shape and has approximately the same peak pressure as the previous figure. The smoother. shape of the revised curve results '
from additional numerical iteration on the initial cc'ditions of !
pressurizer pressure, pressurizer water level and steam generator inventory in order to more closely achieve a- high pressurizer pressure trip at about the same instant that the steam generator dries out. i
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2,600 O 0.5 1.0 1.5 BREAK AREA, FT 2 l t
i Amendment R July 15,1993
- v. Figure ;
FEEDWATER UNE BREAK
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1 jQ MAXIMUM RCS PRESSURE vs BREAK AREA ,
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Ouestion 440.215: ;
On page 1S.3.8 and Table 15.3.3-2, it is stated that "a high primary system pressure and a low core inlet temperature were chosen to maximize ,
the amount of failed fuel." A higher pressurizer pressure and a lower ,
core inlet temperature would result in a higher initial DNBR. Explain why the transient would result in a maximum amount of failed fuel when the r initial DNBR is assumed at a higher value. -
Response 440.215:
The initial DNBR is assumed at the lower value. The high pressure and low i '
temperature initial conditions result in an initial DNBR being lower than that for the lower pressure and higher temperature initial conditions.
- This phenomenon occurs due to the operational characteristics of a !
COLSS/CPC plant. During operation, a specified portion of the total core thermal margin is reserved for the safety analyses and is called the Required Overpower Margin (ROPM). If this margin is violated, an alarm will sound alerting the operator that the Power Operating Limit (POL) was violated. Determination of-a violation of the R0PM is performed by COLSS 1 and all parameters affecting DNBR are considered in this determination.
If the plant is operating at a high pressure and low temperature other parameters, e.g., F,, will be allowed to vary in the more adverse direction as long as the required thermal margin is main?.ained. The locked rotor ,
event was initiated from this minimum thermal margin. The resulting ;
initial DNBR with high RCS pressure and low RCS temperature reflects this ;
minimum thermal margin. Thus, the benefit to DNBR of high pressure and low temperature are offset by the assumed variation of other DNBR related parameters in the more adverse direction. The analysis was therefore based on the combination of conditions yielding the minimum transient
- DNBR.
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. i Ouestion 440.216:
Page 15.5-5 The maximum charging flow is assumed to be 150 gpm decreasing from 250 gpm in the existing analysis (CESSAR-DC, Amendment H). Explain why a lower .
charging flow is assumed for the limiting case analysis. i Res'ponse 440.216:
l A 150 gpm charging pump flow is assumed for the limiting case analysis due- i to a change in the final design of the CVCS resulting in a decrease in the maximum charging pump runout flow. The charging pump runout flow has been '
reduced such that the maximum possible flow to the RCS is 150 gpm.
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Ouestion 440.217: i Page 15.5-6 It is stated that "a pressurizer absolute high level alarm at 65% of the '
pressurizer volume will prevent water from being discharged out of the '
pressurizer safety valves." Confirm that the appropriate procedures are ;
available and describe the operator actions to avoid water discharge from ;
the PSVs in responding to the alarm. ,
Response 440.217:
The words in CESSAR DC will be changed to read "A pressurizer absolute -
high level alarm at 65% of the pressurizer volume will alert the operator to the increase in water level." The operator will prevent water discharge from the PSVs by stopping charging flow and increasing letdown.
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CESS AR !anflCATION
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The 2713 lbs of steam discharged by the pressurizer safety valve are contained within the in-containment refueling water storage tank with no releases to the atmosphere. The main steam safety valves discharge 134,009 lbs of steam to the atmosphere prior to 1800 seconds. At 1800 seconds, the operator stabilizes the plant and using the atmospheric dump valves. initiates plant cooldown, 15.5.2.4 Conclusions The peak RCS and steam generator pressures reached during the Pressurizer Level Control System malfunction with a loss of offsite power at turbine trip and the limiting single failure are 2682 psia and 1266 psia, respectively. These pressures are less than 110% of the design pressures. Since this transient is due primarily increase toinan increase in primary inventory which causes an RCS pressure, the DNBR increases until reactor / turbine trip at which time the loss of offsite power resulting in a decrease in reactor coolant flow causes the DNBR to decrease to a minimum of 1.62. Therefore, the acceptance criterion regarding fuel performance is met.
A pressurizer absolute high level alarm at 65% of the pressurizer T volume will prevent water from being discharged out of the j pressurizer safety valves.f An interval of 30 minutes is assured between the alarm and required operator action.
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l Amendment N 15.5-6 April 1, 1993
. .=- -- . ..- - . . . - ._. . ... . ..___. - __ _
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I INSERT A t i
A pressurizer absolute high level alarm at 65% of the pressurizer volume will alert the operator to the increase in' water level. j i
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Question 440.218:
Since the ADV block valves are credited for. the SGTR analysis and the pressurizer backup heaters are credited in the letdown line break analysis, confirm that both valves and backup heaters are the safety grade components.
Response 440.218:
The ADV block valves are safety grade and are credited in the SGTR analysis to mitigate the consequences of this event if a failure of an ADV to close is assumed. However, the purpose of assuming operation of the pressurizer backup heaters during the letdown line break event is to make j the consequences of the event more adverse. Therefore, they are not )
" credited" to mitigate the consequences of this event and as such are not j required to be safety grade. 1
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Ouestion 440.219:
Results Comparison The staff finds that the results of reanalysis are less severe than that
.of the existing analysis (Amendment H) event though the rated power was !
increased by 3%, a loss of offsite power (LOOP) was assumed without event l recategorization and the LOOP delay time was assumed to be zero second.
The applicant should identify the changes in system parameters - and l calculational methods for each case which was reanalyzed and assess the !
effects of the changes on the safety analysis results. For each change affected the safety analysis results, appropriate technical justifications should be provided.
Response 440.219: 4 i
Attached are slides which were presented at the June 16, 1993 meeting with I the NRC. The slides address the above request for the limiting Chapter 15 events.
Changes to the slides have been made and are identified by the lines in the right margin. Changes were made in order to provide additional information and clarification, and to make corrections.
i
SYSTEM 80+ CHAPTER 15 REANALYSIS FOR POWER UPGRADE-J I
JUNE 16, 1993 j
ABB-COMBUSTION ENGINEERING, INC.
t SYSTEM 80+ CHAPTER 15 REANALYSIS FOR '
POWER UPGRADE
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l M
O CHANGES INCLUDED IN THE SYSTEM 80+ REANALYSIS l i
i O DISCUSSION OF LIMITING EVENTS t i
-CHANGES.TO THE EVENT DEFINITION t
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-SIGNIFICANT ANALYSIS CHANGES j i
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-IMPACT ON RESULTS i
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SUMMARY
OF CHANGES INCLUDED IN THE SYSTEM 80+ REANALYSIS O THE NOMINAL REACTOR COOLANT INLET TEMPERATURE IS DECREASED FROM 558 F TO 555.8 F O CORE POWER WAS INCREASED FROM 3800 MWT TO 3914 MWT [
O THE INITIAL CONDITION SPACE FOR THE CHAPTER 15 ;
ANALYSES WAS CHANGED AS FOLLOWS: '
AMENDMENT H AMENDMENT N CORE POWER, % OF FULL POWER 0-102 0-102 ;
PRESSURIZER PRESSURE, PSIA 1905-2375 2175-2325 CORE INLET TEMPERATURE, I
< 90% POWER, F 543-565 543-561 90% - 100% POWER, F 553-563 550-561
- O AN INTEGRAL BURNABLE POISON WAS SUBSTITUTED FOR' SHIM RODS (WITH THE USE OF INTEGRAL BURNABLE POISON, THE 3914 MWT FULL POWER CORE AVERAGE HEAT FLUX IS LESS ,
THAN THAT FOR THE 3800 MWT CORE DESIGN) !
O THE MOST POSITIVE MODERATOR TEMPERATURE COEFFICIENT AT FULL POWER HAS BEEN REDUCED FROM 0. 0 TO -0.1 X 10-*
DELTA-RHO / F. AT ZERO POWER, THE MTC HAS BEEN '
REDUCED FROM +0.5X10-* DELTA-RHO / F TO 0. 0.
O THE CEDM HOLDING COIL DECAY TIME HAS BEEN DECREASED FROM 0.8 SEC TO 0.5 SEC O IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS ,
FOR AMENDMENT N ;
THE SHUTDOWN REACTIVITY WORTH VERSUS POSITION .
USED FOR AMENDMENT N WAS BASED ON A +0.3 ASI. I FOR AMENDMENT H A +0.6 ASI WAS USED.
THE 90% CEA INSERTION TIME HAS BEEN REDUCED FROM 3.66 SEC TO 3.5 SEC O THE UACERTAINTY ON THE MAXIMUM PRIMARY SAFETY VALVE OPENING SETPOINT HAS BEEN INCREASED FROM 25 PSIA TO 40 PSIA
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SUMMARY
OF CHANGES INCLUDED IN THE SYSTEM 80, REANALYSIS (CONTINUED) i O FINAL DESIGN DETAIL INCREASED THE ASSURED FLOW RATE ,
THROUGH THE PRESSURIZER SAFETY VALVES FROM 460,000 ;
LBM/HR TO 525,000 LBM/HR. THE MAXIMUM FLOW HAS BEEN INCREASED FROM 575,000 LBM/HR TO 630,000 LBM/HR O THE TIME DELAY FOR LOSS OF OFFSITE POWER FOLLOWING TURBINE TRIP WAS DECREASED FROM 3 SECONDS TO O SECONDS O THE MAXIMUM MAIN FEEDWATER SYSTEM FLOW WAS INCREASED l FROM 140% TO 160%
O THE MAXIMUM CHARGING FLOW TO THE RCS WAS DECREASED TO 150 GPM ,
i O THE STEAM GENERATORS HAVE BEEN REDESIGNED (E.G. THE NUMBER OF TUBES HAS INCREASED)
O FINAL DESIGN DETAIL INCREASED THE SURGE LINE LENGTH BY 25 FEET O THE DIAMETER OF THE DIRECT VESSEL INJECTION LINES WAS INCREASED NOMINALLY FROM 10 TO 12 INCHES O THE LETDOWN LINE FLOW RESISTANCE HAS BEEN INCREASED O THE REACTOR COOLANT PUMPS HAVE BEEN REDESIGNED (E.G., THE RATED HEAD HAS INCREASED)
O THE SITE ATMOSPHERIC DILUTION FACTORS, X/Qs, WERE CHANGED TO THE EPRI URD VALUES O THE OFFSITE DOSES FOR EVENTS INVOLVING FUEL FAILURE WERE COMPUTED USING THE NUREG-1465 SOURCE TERM O THE TORC COMPUTER CODE WAS USED INSTEAD OF THE CETOP CODE TO COMPUTE THE MINIMUM DNBRs FOR THE FEEDWATER LINE BREAK, LOSS OF CONDENSER VACUUM, LOCKED ROTOR,
., AND STEAM GENERATOR TUBE RUPTURE EVENTS
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ca 6
S 8
1>
't MOST UMITING h FUEL LICENSING o
UMIT REQUIRED TOLERANCE MARGIN '
(E.G. INSTRUMENT ERRORS.
MANUFACTURING TOLERANCES, EQUIPMENT CAPABluTIES, ETC.)
i.
E Ci g Qg REQUIRED MARGIN FOR MOST LIMITING UCENSING BASIS UMIT
%5
- a. g o av -
wo 3: < .
o Q- ,
o AVAILABLE MARGIN FOR OPERATIONS E (157. MINIMUM) e g o a
a RATED POWER - '
NORMAL FUEL OPERATING * '
RANGE I
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Figure 4.2-1 ALWR Fuel Thermal Margin .
Page 4.2 4 ~
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15.1.4 INADVERTENT OPENING OF A STEAM GENERATOR >
ATMOSPHERIC DUMP VALVE i
O EVENT DEFINITION CHANGE ZERO TIME DELAY FOR LOSS OF OFFSITE POWER l
- LOSS OF OFFSITE POWER CONSIDERED AS PART OF THE '
EVENT (NO RECATEGORIZATION)
O SIGNIFICANT ANALYSIS CHANGES
- ROPM INCREASED r
- INITIAL DNBR INCREASED FROM 1.52 (AMENDMENT H) TO 1.62 (AMENDMENT N) '
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM O.8 SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) j
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR ;
AMENDMENT N
- AMENDMENT H ANALYSIS CHOSE AN INITIAL DNBR SUCH ,
THAT THE RESULTING TRANSIENT MINIMUM DNBR AT 30 MINUTES WAS 1.24. AMENDMENT N 30 MINUTE DNBR 1.36 1
O IMPACT ON RESULTS
- ROPM, DECREASED COIL DECAY TIME, IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS AND CHANGE IN ^
l ANALYSIS METHODOLOGY ALLOWS EVENT TO ACCOMODATE LOSS OF OFFSITE POWER WITHOUT FUEL FAILURE l AMENDMENT H AMENDMENT R EVENT W/O A SINGLE FAILURE j MINIMUM DNBR 1.24 1.30 l EVENT WITH A SINGLE FAILURE MINIMUM DNBR 1.24 1.29 l
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i 15.1.5 CASE 5 ,
FULL POWER OUTSIDE CONTAINMENT STEAM LINE BREAK i
O EVENT DEFINITION CHANGE LOSS OF OFFSITE POWER ASSUMED COINCIDENT WITH l REACTOR TURBINE TRIP '
O SIGNIFICANT ANALYSIS CHANGES ,
- ROPM INCREASE i
- INITIAL DNBR 1.62 (AMENDMENT R) VERSUS AMENDMENT H VALUE OF 1.53 !
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 $
SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) ;
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR AMENDMENT N _j
- THE MODERATOR TEMPERATURE COEFFICIENT HAS CHANGED !
FROM -5.4 X 10-' DELTA-RHO / F (AMENDMENT H) TO -3.5 :
X 10 DELTA-RHO / F (AMENDMENT N) j
- NUREG 1465 SOURCE TERM
- EPRI X/Qs !
O IMPACT ON RESULTS <
- DNBR AND FUEL FAILURE IMPROVE DUE TO ROPM, COIL DECAY TIME, IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS AND MTC ,
- DOSE CHANGE SMALL DUE TO LOWER IODINE RELEASES I (NUREG-1465) AND HIGHER X/Qs {
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15.1.5 CASE 5 FULL POWER OUTSIDE CONTAINMENT STEAM LINE BREAK (CONTINUED)
AMENDMENT H AMENDMENT N MINIMUM DNBR 1.18 1.25 FUEL FAILURE O.5% 0.5% (ASSUMED) {
2 HOUR THYROID DOSE 67 REM 70 REM l
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15.1.5 CASE 6 ZERO POWER STEAM LINE BREAK OUTSIDE CONTAINMENT WITH LOSS OF OFFSITE POWER O EVENT DEFINITION CHANGE
- NO CHANGE 1
O SIGNIFICANT PRALYSIS CHANGES
- COLD LEG TEMPERATURE OF 565 F (AMENDMENT H) REDUCED TO 561 F (AMENDMENT N): LOWER INITIAL SG ,
PRESSURES
- EPRI X/Qs 4
l O IMPACT ON RESULTS ,
i
- LOWER MASS RELEASE DUE TO LOWER INITIAL SG PRESSURES
- DOSE INCREASES DUE TO INCREASED X/Q CHANGE: LOWER j MASS RERLEASE HAS SMALL EFFECT
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AMENDMENT H AMENDMENT N MASS RELEASED FROM AFFECTED SG 623,000 LBS 571,000 LBS (2 HOUR) 2 HOUR THYROID DOSE PIS 15 REM 28 REM GIS 13 REM 23 REM
k 15.2.3 LOSS OF CONDENSER VACUUM O EVENT DEFINITION CHANGE !
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER
- LOSS OF OFFSITE POWER CONSIDERED AS PART OF THE :
EVENT (NO RECATEGORIZATION)
O SIGNIFICANT ANALYSIS CHANGES
- ROPM INCREASED
- INITIAL DNBR INCREASED FROM 1.53 (AMENDMENT H) TO l 1.63 (AMENDMENT N) ;
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 l SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N)
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR AMENDMENT N '
- 3% POWER UPGRADE IMPACTS PEAK PRESSURE :
- PSV LIFT PRESSURE UNCERTAINTY INCREA. SED FROM 25 PSI (AMENDMENT H) TO 40 PSIA (AMENDMENT N)
- COLD LEG TEMPERATURE REDUCED 3 F i
- THE TORC COMPUTER CODE WAS USED TO CALCULATE MINIMUM DNBR AS OPPOSED TO THE CETOP CODE USED FOR THE AMENDMENT H ANALYSIS i
O IMPACT ON RESULTS
- DNBR IMROVES DUE TO LOSS OF OFFSITE POWER )
COINCIDENT WITH TURBINE TRIP '
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15.2.3 .
LOSS OF CONDENSER VACUUM i (CONTINUED) 1
- ROPM, DECREASED COIL DECAY TIME, IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS AND USE OF THE TORC {
CODE ALSO IMPROVE DNBR !
- DNBR EVENT WITH LOSS OF OFFSITE POWER IS' BOUNDED BY !
THE LOSS OF FLOW EVENT IN SECTION 15.3.1 -
- EVENT WITHOUT LOSS OF OFFSITE POWER IS LIMITING FOR l PEAK PRESSURE PEAK RCS PRESSURE INCREASES DUE TO INCREASED POWER ,
AND PSV LIFT PRESSURE, AND REDUCED COLD LEG {
TEMPERATURE ,
O IMPACT ON RESULTS i AMENDMENT H AMENDMENT N MINIMUM DNBR 1.07 1.26 !
FUEL FAILURE 1.8% 0.0% !
PEAK RCS PRESSURE 2707 PSIA 2726 PSIA f 1
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15.2.8 FEEDWATER LINE BREAK -
O EVENT DEFINITION CHANGE 1
ZERO TIME DELAY FOR LOSS OF OFFSITE POWER [
q O SIGNIFICANT ANALYSIS CHANGES i
- ROPM INCREASED l
- INITIAL DNBR INCREASED FROM 1.56 (AMENDMENT H) TO !
1.62 (AMENDMENT R)
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) f
- IMPROVED SHUTDOWN CEP. INSERTION CHARACTERISTICS FOR ,
AMENDMENT N -l
- 3% POWER UPGRADE IMPACTS PEAK PRESSURE
. PSV LIFT PRESSURE UNCERTAINTY INCREASED FROM 25 PSI .f (AMENDMENT H) TO 40 PSIA (AMENDMENT N) i
- SURGE LINE LENGTH WAS INCREASED BY 25 FEET ;
- INITIAL PRESSURIZER PRESSURE INCREASED FROM 1971.3- -
l PSIA (AMENDMENT H) TO 2250 PSIA (AMENDMENT N) FOR- '
PEAK PRESSURE CASE
- THE TORC COMPUTER CODE WAS USED TO CALCULATE :
- MINIMUM DNBR AS OPPOSED TO THE CETOP CODE l USED FOR THE AMENDMENT H ANALYSIS
- NUREG-1465 SOURCE TERM
- EPRI X/Qs l
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l 15.2.8 FEEDWATER LINE BREAK (CONTINUED)
O IMPACT ON RESULTS
- DNBR MORE ADVERSE DUE TO PSV BLOWDOWN
- PEAK RCS PRESSURE INCREASED DUE TO HIGHER PSV OPENING PRESSURE, HIGHER POWER LEVEL, AND INCREASED SURGE LINE LENGTH
- PEAK SG PRESSURE IS REDUCED DUE TO A LARGE BREAK 2 BEING THE LIMITING BREAK FOR M4ENDMENT N (0.7 FT 2
VERSUS 0.3 FT FOR AMENDMENT H)
- DOSE INCREASES DUE TO FUEL FAILURE AND X/Qs N4ENDMENT H AMENDMENT N AMENDMENT R MINIMUM DNBR 1.56 1.21 1.17 FUEL FAILURE 0.0% 0.15% 0.22%
PEAK RCS PRESSURE 2720 PSIA 2785 PSIA 2793 PSIA-PEAK SG PRESSURE 1251 PSIA 1189 PSIA 1237 PSIA 2 HOUR THYROID 8.3 REM 22.9 REM 4.7 REM DOSE
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15.3.1 1
TOTAL LOSS OF REACTOR COOLANT FLOW t l
i O EVENT DEFINITION CHANGE :
- NO CHANGE: LOSS OF OFFSITE POWER ORIGINALLY ,
(AMENDMENT H) CONSIDERED EVENT ;
INITIATOR O SIGNIFICANT ANALYSIS CHANGES i
- ROPM INCREASED
- INITIAL DNBR 1.62 VERSUS AMENDMENT H VALUE OF 1.53
- COIL DECAY TIME FOR SCRAM KODS DECREASED FROM O.8 SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) i
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR ;
AMENDMENT N '
)
- RCP REDESIGN l l
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- COLD LEG TEMPERATURE REDUCED 15 F FOR PEAK PRESSURE .i CASE: IMPACTS RCS AND SG PRESSURE
- INITIAL PRESSURIZER PRESSURE FOR PEAK PRESSURE l CASE REDUCED FROM 2425 PSIA (AMENDMENT H) TO 2325 {
PSIA (AMENDMENT N) 1 l
- PSV LIFT PRESSURE UNCERTAINTY INCREASED FROM 25 PSI (AMENDMENT H) TO 40 PSI (AMENDMENT N) i
- THE FULL POWER MODERATOR TEMPERATURE COEFFICIENT !
CHANGED FROM 0.0 (AMENDMENT H) TO -0.1 X 10-' DELTA- i RHO / F (AMENDMENT N) FOR DNBR CALCULATIONS l I
- THE INITIAL AXIAL POWER SHAPE CHANGED FROM AN ASI I OF 0.0 (AMENDMENT H) TO +0.3 (AMENDMENT N) i
- 3% POWER UPGRADE IMPACTS PEAK PRESSURE 4
5 l
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a 15.3.1 !
TOTAL LOSS OF REACTOR COOLANT FLOW !
(CONTINUED) l O IMPACT ON RESULTS ,
- MINIMUM DNBR IMPROVED DUE TO ROPM, DECREASED COIL DECAY TIME, RCP REDESIGN, IMPROVED SHUTDOWN CEA l
,- INSERTION CHARACTERISTICS AND MTC i i
- LOWER COLD LEG TEMPERATURE REDUCES PEAK SG PRESSURE i I
-HIGHER POWER AND PSV SETPOINT INCREASES PEAK RCS l PRESSURE j AMENDMENT H AMENDMENT N AMENDMENT R MINIMUM DNBR 1.24 1.27 1.27 !
i
) PEAK SG PRESSURE 1274 PSIA 1249 PSIA 1273 PSIA i
PEAK RCS PRESSURE 2636 PSIA 2652 PSIA 2665 PSIA i
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ROTOR SEIZURE i
O EVENT DEFINITION CHANGE i
?
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER !
FOR DNBR CALCULATION :
I O SIGNIFICANT ANALYSIS CHANGES i
- ROPM INCREASED
- INITIAL DNBR 1.62 (AMENDMENT R) VERSUS AMENDMENT H :
VALUE OF 1.57
- COIL DECAY TIME FOR SCRAJ4 RODS DECREASED FROM 0. 8 SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N)
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS l FOR AMENDMENT N !
- RCP REDESIGN <
l t
- PSV LIFT PRESSURE UNCERTAINTY INCREASED FROM 25 PSI (AMENDMENT H) TO 40 PSIA (AMENDMENT N)
- INITIAL PRESSURIZER PRESSURE FOR PEAK PRESSURE !
CASE REDUCED FROM 2400 PSIA (AMENDMENT H) TO 2325 ;
PSIA (AMENDMENT N) l
- COLD LEG TEMPERATURE REDUCED 20 F FOR PEAK PRESSURE l CASE
- AMENDMENT H PEAK PRESSURE CASE ASSUMED EARLY LOSS OF OFFSITE POWER: PRIOR TO REACTOR TRIP ;
i
- THE FULL POWER MODERATOR TEMPERATURE COEFFICIENT HAS CHANGED FROM +0. 056 X 10-* DELTA-RHO / F l (AMENDMENT H) TO - 0.1 X 10-* DELTA RHO / F (AMENDMENT N) FOR THE FUEL PERFORMANCE CASE l i
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i 15.3.3 l ROTOR SEIZURE !
(CONTINUED)
- THE INITIAL AXIAL POWER SHAPE WAS CHANGED FROM AN l ASI OF -0.3 (AMENDMENT H) TO +0.3 (AMENDMENT N) !
- NUREG-1465 SOURCE TERM ;
- LOWER AND CONSTANT SG INVENTORY FOR AMENDMENT H l DOSE ANALYSIS, HIGHER AND VARIABLE SG INVENTORY FOR l AMENDMENT N CALCULATION [
- EPRI X/Qs
- THE TORC COMPUTER CODE WAS USED TO CALCULATE :
MINIMUM DNBR AS OPPOSED TO THE CETOP CODE USED FOR THE AMENDMENT H ANALYSIS
~
O IMPACT ON 'ESULTS _i
)
- DNBR AND FUEL FAILURE IMPROVES DUE TO ROPM, COIL l DECAY TIME Ib' PROVED SHUTDOWN CEA INSERTION j
- CHARACTERISTICS, MTC, RCP REDESIGN AND USE OF THE !
TORC COMPUTER CODE ;
i
- PEAK RCS PRESSURE Rhi>UCED DUE TO If0WER INITIAL l PRESSURIZER PRESSURE, TIMING OF LOSb OF OFFSITE l
, POWER AND COLD LEG TEMPERATURE .j i
- PEAK SG PRESSURE REDUCED DUE TO REDUCTION IN COLD LEG TEMPERATURE ll 1
- DOSE IS REDUCED DUE TO REDUCED FUEL FAILURE; !
.)
AFFECTED BY SG INVENTORY !
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15.3.3 ROTOR SEIZURE (CONTINUED) l l
AMENDMENT H AMENDMENT N AMENDMENT R ;
MINIMUM DNBR 0.83 1.09 1.09 ,
FUEL FAILURE 3.5% 1.2% 1.2% i PEAK RCS PRESSURE 2647 PSIA 2615 PSIA 2635 PSIA :
PEAK SG PRESSURE 1276 PSIA 1248 PSIA 1273 PSIA ,
2 HOUR THYROID DOSE 18.1 REM 2.9 REM 2.9 REM i
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i 15.4.1 ;
UNCONTROLLED CEA WITHDRAWAL FROM SUBCRITICAL '
O EVENT DEFINITION CHANGE ,
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER
- LOSS OF OFFSITE POWER CONSIDERED AS PART OF THE EVENT (NO RECATEGORIZATION) l O SIGNIFICANT ANALYSIS CHANGES .
i
- MODERATOR TEMPERATURE COEFFICIENT DECREASED FROM '
+0. 5 X 10-* DELTA-RHO / F (AMENDMENT H) TO 0.0 (AMENDMENT N) .
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) l
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR AMENDMENT N O IMPACT ON RESULTS I
- DNBR INCREASED DUE TO REDUCTION IN MTC, COIL DECAY TIME, AND IMPROVED SHUTDOWN CEA INSERTION ;
CHARACTERISTICS AMENDMENT H AMENDMENT N l MINIMUM DNBR 2.26 3.71 l i
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i 15.4.2 UUCONTROLLED CEA WITHDRAWAL AT POWER j i
i O EVENT DEFINITION CHANGE j
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER !
- LOSS OF OFFSITE POWER CONSIDERED AS PART OF THE !
EVENT (NO RECATEGORIZATION) l O SIGNIFICANT ANALYSIS CHANGES +
i
- ROPM INCREASED i i
- INITIAL DNBR INCREASED FROM 1.51 (AMENDMENT H) TO
- 1.62 (AMENDMENT N) ,
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 !
SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) ;
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR l AMENDMENT N i
O IMPACT ON RESULTS ;
- ROPM,. DECREASED COIL DECAY TIME AND IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS INCREASES DNBR; LOSS OF OFFSITE POWER DECREASES DNBR: NET !
CHANGE IS SMALL j i
AMENDMENT H AMENDMEITT N I MINIMUM DNBR 1.32 1.33 l
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15.4.3 l
SINGLE CEA DROP l l
O EVENT DEFINITION CHANGE: NOT LIMITING; NOT PRESENTED l IN SAR ,
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER i i
- LOSS OF OFFSITE POWER CONSIDERED AS PART OF THE l EVENT (NO RECATEGORIZATION) !
i i
O SIGNIFICANT ANALYSIS CHANGES: EVENT WITHOUT LOSS OF OFFSITE POWER (NO REACTOR / TURBINE TRIP) {
i
- ROPM INCREASED !
- INITIAL DNBR INCREASED FROM 1.58 (AMENDMENT H) TO l 1.62 (AMENDMENT N) !
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, - AMENDMENT H ANALYSIS CHOSE AN INITIAL DNBR SUCH THAT THE RESULTING TRANSIENT MINIMUM DNBR WAS 1.24 l O IMPACT ON RESULTS l
- ROPM, AND METHODOLOGY CHANGE INCREASES DNBR l l
AMENDMENT H AMENDMENT N ,
MINIMUM DNBR 1.24 1.35 t i
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15.4.6 l INADVERTENT DEBORATION O EVENT DEFINITION CHANGE
- F.O CHANGE i i
.) SIGNIFICANT ANALYSIS CHANGES i
- AMENDMENT N ANALYSIS UTILIZES THE BORON DILUTION l ALARM IN MODES 3-6; THE MAKEUP WATER FLOW ALAPJ4 IS :
A BACKUP IN MODE 6. AMENDMENT H ANALYSIS UTILIZED A COMBINATION OF THE BORON DILUTION AND MAKEUP WATER FLOW ALARMS
- THE SHUTDOWN MARGIN WAS INCREASED TO 5.75%
(AMENDMENT N) FOR MODE 5 IN THE DRAINED CONDITION. .
THE AMENDMENT H ANALYSIS USED 3% ,
- THE VOLUME OF WATER IN THE SHUTDOWN COOLING SYSTEM -)
WAS REDUCED FOR THE AMENDMENT N ANALYSIS
- THE CHARGING FLOW OF 180 GPM (AMENDMENT H) WAS REDUCED TO 160 GPM (AMENDMENT N)
O IMPACT ON RESULTS AMENDMENT H AMENDMENT N TIME TO REACH CRITICALITY 38 MINUTES 67 MINUTES i
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15.4.8 CEA EJECTION J
i O EVENT DEFINITION CHANGE t
- NO CHANGE i
PEAK RCS AND SECONDARY PRESSURE CASES: LOSS OF l OFFSITE POWER WITH A ZERO TIME DELAY l i
SECONDARY RELEASE CASE: LOSS OF OFFSITE POWER WITH I A ZERO TIME DELAY CONTAINMENT RELEASE CASE: LOSS OF OFFSITE POWER AT' T=0 FUEL PERFORMANCE CASES: LOSS OF OFFSITE POWER HQT j MODELED ,
O SIGNIFICANT ANALYSIS CHANGES
- ROPM INCREASED j i
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) l l
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR AMENDMENT N FOR PEAK PRESSURE AND SECONDARY l' RELEASE CASES l ;
- CORE AVERAGE LINEAR HEAT GENERATION RATE IS-LOWER l FOR THE 3914 MWT DESIGN AS COMPARED TO THE 3800 l !
~
MWT DESIGN DUE TO THE ADDITION OF INTEGRAL BURNABLE l POISON l 1
- PSV LIFT PRESSURE UNCERTAINTY INCREASED FROM 25 PSI J (AMENDMENT H) TO 40 PSI (AMENDMENT N) l i
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15.4.8 !
t CEA EJECTION (CONTINUED) {
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- THE FULL-POWER MODERATOR TEMPERATURE COEFFICIENT 3 HAS CHANGED FROM 0.0 (AMENDMENT H) TO -0.1 X 10'*
DELTA-RHO / F (AMENDMENT N)
- METHODOLOGY CHANGED FOR COMPUTING OFFSITE DOSES
- NUREG-1465 SOURCE TERM
- EPRI X/Qs O IMPACT ON RESULTS
- ROPM, DECREASED COIL DECAY TIME, AND NEGATIVE MTC l REDUCE AMOUNT OF FUEL FAILURE
- PEAK RCS PRESSURE DECREASES DUE TO DECREASED COIL l DECAY TIME AND IMPROVED SHUTDOWN CEA INSERTION I CHARACTERISTICS - INCREASED PSV OPENING SETPOINT l AFFECTS FINAL RESULT i
- DOSES INCREASE DUE TO METHODOLOGY CHANGE l l
AMENDMENT H AMENDMENT N l FUEL FAILURE < 10% (6. 8%) 4.4% l l
PEAK RCS PRESSURE 2742 PSIA 2733 PSIA 2 HOUR THYROID DOSE 2.95 REM 69.6 REM i
._.-.___---....~.___._.,i
r 15.5.2 PLCS MALUUNCTION O EVENT DEFINITION CHANGE
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER ;
- LOSS OF OFFSITE POWER CONSIDERED AS PART OF THE EVENT (NO RECATEGORIZATION)
O SIGNIFICANT ANALYSIS CHANGES !
- ROPM INCREASED ;
- INITIAL DNBR INCREASED FROM 1.57 (AMENDMENT H) TO !
1.62 (AMENDMENT N) :
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 ;
SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N)
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR AMENDMENT N i
- 3% POWER UPGRADE IMPACTS PEAK RCS AND SG PRESSURES ,
- PSV LIFT PRESSURE UNCERTAINTY INCREASED FROM 25 PSI (AMENDMENT H) TO 40 PSI (AMENDMENT N) [
O IMPACT ON RESULTS :
- DNBR INCREASES DUE TO ROPM i i
< - PEAK RCS PRESSURE INCREASES DUE TO INCREASED POWER AND PSV LIFT PRESSURE I
r
- PEAK SG PRESSURE INCREASES DUE TO INCREASED POWER l
AMENDMENT H AMENDMENT N ,
MINIMUM DNBR- 1.57 (INITIAL) 1.62 (INITIAL) ;
PEAK RCS PRESSURE 2639 PSIA 2682 PSIA PEAK SG PRESSURE 1247 PSIA 1266 PSIA i
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j 15.6.2 ;
LETDOWN LINE BREAK I
O EVENT DEFINITION CHANGE l
- LETDOWN LINE BREAK WITH LOSS OF OFFSITE POWER: FOR DNBR CALCULATION (NOT PRESENTED IN SAR)
O SIGNIFICANT ANALYSIS CHAEGES !
- MINIMUM DNBR FOR STEAM GENERATOR TUBE RUPTURE BOUNDS LETDOWN LINE BREAK: HIGHER AVERAGE MASS FLOW RATE OUT THE BREAK AND CONSEQUENTLY HIGHER DEPRESSURIZATION RATE FOR A TUBE RUPTURE ,
- DECONTAMINATION FACTOR IN THE NUCLEAR ANNEX REDUCED FROM 3 (AMENDMENT H) TO 1 (AMENDMENT N)
- SYSTEM 80+ LETDOWN LINE MODEL USED RATHER THAN i SYSTEM 80 SYSTEM 80+ ORIFICES CREDITED i
SYSTEM 80+ LETDOWN HEAT EXCHANGER IN CONTAINMENT CREDITED !
- FLASHING FRACTION MAXIMIZED (TO INCREASE OFFSITE DOSE) BY ASSUMING MINIMUM CHARGING FLOW RATE (44 l GPM); AMENDMENT H CHARGING FLOW RATE: 250 GPM j
- LOWER INITIAL RCS PRESSURE BY 50 PSI l
- LOWER INITIAL COLD LEG TEMPERATURE BY 2 F
- EPRI X/Qs O IMPACT ON RESULTS ;
- LOWER INTEGRATED MASS RELEASE AND FLASHING ,
FRACTION ENABLE DOSE ACCEPTANCE CRITERION TO BE l MET CONSIDERING LOWER DF AND INCREASED X/Qs :
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i 15.6.2 LETDOWN LINE BREAK :
(CONTINUED) !
AMENDMENT-H AMENDMENT N -l MASS RELEASE OUT ,
BREAK 107442 LBS 48617 LBS l FLASHING FRACTION 40.0 % 19.8 % -!
NUCLEAR A.NNEX DF 3 1 .!
2 HOUR X/Q (SEC/M') 4 . 97X10-4 1. 0X10-2 l 2 HOUR THYROID DOSE 22.O REM- 26.7 REM _
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STEAM GENERATOR TUBE RUPTURE >
O EVENT DEFINITION CHANGE i
- ZERO TIME DELAY FOR LOSS OF OFFSITE POWER: FOR DNBR ,
CALCULATION j l
O SIGNIFICANT ANALYSIS CHANGES .
- POWER INCREASE OF 3 %: IMPACTS DOSE <
i
- ROPM INCREASE ,
- INITIAL DNBR INCREASED FROM 1.53 (AMENDMENT H) TO i 1.68 (ARENDMENT N)
I
- COIL DECAY TIME FOR SCRAM RODS DECREASED FROM 0.8 i SEC (AMENDMENT H) TO 0.5 SEC (AMENDMENT N) ;
- IMPROVED SHUTDOWN CEA INSERTION CHARACTERISTICS FOR l AMENDMENT N
- NEW CPC TRIP ALGORITHM: COINCIDENT LOW PRESSURE LOW DNBR I
- EPRI X/Qs !
O IMPACT ON RESULTS ..
- FUEL FAILURE IS PREVENTED VIA NEW CPC TRIP ALGORITHM -l 1
- DOSES INCREASE DUE TO (1) INCREASED STEAM RELEASE-RESULTING FROM HIGHER DECAY HEAT AND (2) INCREASED ~
X/Qs AMENDMENT H AMENDMENT N TWO HOUR THYROID DOSE 44.9 REM 93.1 REM (LIMITING DOSE) 1 l
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Ouestion 440.220: '
In tl.e I&C diversity analysis, the moderator and Doppler reactivity feedback functions are the main parameters to control the core power i increase or decrease in the events analyzed. Provide the values of the moderator and Doppler reactivity coefficients assumed in the analysis and justify the adequacy of these values for each event analyzed. ,
Response 440.220:
The values assumed for the moderator and Doppler- reactivity ;
coefficients / functions in the I&C diversity analysis are provided in the attached Tables 440.220-1 and 440.220-2. For safety analyses of Design Basis Events, conservative bounding values lof parameters are employed.
For the diversity analyses, best estimate values are appropriate although '
in some instances the bounding values were conveniently employed. Table 440.220-1 describes the bias direction for the moderator and Doppler _
temperature coefficients employed. For coolant heatup events, the most positive value of moderator coefficient is justified and for cooldown :
events the most negative value is justified. For the LOCA event, Table j 440.220-2 is employed for moderator reactivity. Table 440.220-1 also .
gives the bias employed on the Doppler temperature coefficient. For events which heat the fuel, the least negative value ir justified and for ,
cooldown events, the most negative value is justified. Values are listed in CESSAR-DC Tables 15.0-5 and 15.0-6 as a function of fuel temperiture. '
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REACTIVITY COEFFICIENTS EMPLOYED IN EVENT ANALYSES
- TABLE 440.220-1 >
EVENT MODERATOR MTC BIAS DOPPLER DTC BIAS TEMPERATURE TEMPERATURE COEFFICENT COEFFICIENT **
(d @ / F) (MULTIPLIER)
LOSS OF FLOW - 1. 8X 10 MEDIAN VALUE LEA! :GATIVE MOST !
(1.0) CONSERVATIVE i (WITHOUT UNCERTAINTY)
LOCKED ROTOR / - 0.1 X 10 MOST POSITIVE LEAST NEGATIVE MOST SHEARED SHAFT (0.85) CONSERVATIVE (WITH UNCERTAINTY)
CEA EJECT 10 - 1. 8 X 10 MEDIAN VALUE LEAST NEGATIVE MOST (1.0) CONSERVATIVE (WITHOUT UNCERTAINTY)
LETDOWN LINE (DNBR for this event is bounded by DNBR for SGTR Event.)
BREAK SGTR 0.0 REPRESENTA- MOST NEGATIVE REPRESENTATIVE .
TIVE VALUE: (1.38) VALUE: MARGINAL MARGINAL IMPACT t IMPACT STEAM LINE BREAK CESSAR-DC EXTREMELY MOST NEGATIVE MOST FIG. 15.1.5-0 CONSERVATIVE (1.38) CONSERVATIVE MODERATOR MTP-5.4X10 (WITH REACTIVITY VS. M/F UNCERTAINTY)
TEMPERATURE FEEDWATER LINE 0.0 MOST POSITIVE LEAST NEGATIVE MOST BREAK * (1.0) CONSERVATIVE (WITHOUT UNCERTAINTY)
LOCA SEE TABLE BE*-HF P MOST NEGATIVE MOST 440.220-2 MODERATOR (1.0) CONSERVATIVE DENSITY VS. (WITHOUT REACTIVITY UNCERTAINTY)
(TABLE 440.220-2)
Best estimate i
<
- Values are for reanalysis to be included in future Amendment.
" See CESSAR-DC Tables 15.0-5 and 15.0-6 for Doppler reactivity as a function of temperature.
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BEST ECTIMATE REACTIVITY VS. MODERATOR DENSITY FOR LOCA ANALYSES t
TABLE 440.220-2 s
CORE AVERAGE REACTIVITY 3
DENSITY (LBM/FT ) (of )
27.00000 -0.044248 28.00000 -0.038829 29.00000 -0.033939 j 30.00000 -0.029527 31.00000 -0.025552 !
32.00000 -0.021972 >
33.00000 -0.018755 ,
34.00000 -0.015867 35.00000 -0.013283 36.00000 -0.010977 37.00000 -0.008928 ,
38.00000 -0.007116 :
39.00000 -0.005523 40.00000 -0.004135 ,
41.00000 -0.002935 42.00000 -0.001911 43.00000 -0.001049 i
44.00000 -0.000335 l 45.00000 0.000245 46.00000 0.000704 !
47.00000 0.001059 48.00000 0.001325 49.00000 0.001522 50.00000 0.001671 i
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e Ouestion 440.221 The design RCS flow, as stated in Chapter 15 of CESSAR-DC, is 445,600 gpm.
The vessel flow is assumed to be 461,200 gpm in the analysis. Clarify the difference in flow rates and revise, if necessary, the analysis to reflect the correct flow rate consistent with the best estimate method proposed in the analysis. ;
Response 440.221 ;
The actual best estimate flow rate for the System 80+ plant ranges between >
477,099 gpm and 480,211 gpm. These flow rates were computed based on best estimate system resistances and reactor coolant puro performance. The -
461,200 gpm flow rate was chosen in order to be somewhat higher than the ,
Chapter 15 value, since the flow rate c:pected to Le seen at an actual j plant is normally higher than that stated in Chapter 15. Thus the value of 461,200 gpm was chosen for expediency rather than an actual best estimate value. The 461,200 gpm is approximately the average between i 445,600 gpm and the minimum best estimate flow rate of 477,099 gpm. See !
1 the response to RAI 440.225.
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O Ouestion 440.222:
To be consistent with the operator action analysis in Section 2.4, the RCP trip delay time is assumed to be 23 minutes after,the 6" LOCA initiation (page 75). For the SBLOCAs of 3" and 0.041 ft breaks, the RCPs are tripped much sooner (17 minutes following the LOCAs stated in pages 75 and 76). Clarify the discrepancy in the RCP trip times assumed in the SBLOCAs.
Response 440.222:
The estimate of reactor operator response described on page 74 (LD-93-080, dated May 19, 1993) indicates that the operator could be reasonably expected to trip two RCPs within 16 minutes and two more within 22 minutes. These times for operator action are consistent with ABB's " Trip 2 / Leave 2" (T2/L2) RCP trip Emergency Procedure Guidelines and represent the best estimate of the appropriate times for tripping the RCPs.
As described on page 74, the RCP trip delay time of 23 minutes for the 6" break in the top of the pressurizer was not selected to be consistent with the T2/L2 operator action analysis but to be conservative. As stated on page 74, it was found to be conservative for this break in the pressurizer to delay RCP trip. Therefore, it was assumed that all four RCPs were l tripped 23 minutes after event initiation. This time is one minute beyond the time for reasonable operator completion of the T2/L2 procedure.
As described on page 75, the RCP trip delay time of 17 minutes for the 3" break in the top of the cold leg was also not selected to be consistent with the T2/L2 operator action analysis but to be conservative. Early RCP trip was found to be conservative for this break in the cold leg.
Therefore, it was assumed that all four RCPs were tripped 17 minutes after event initiation. This time is one minute after the time of assumed HPSI j actuation and accounts for early recognition by the operator of the need '
to trip all RCPs during the LOCA event.
As descrjbed on page 76, the RCP trip delay time of 17 minutes for the 0.041 ft break in the top of the vessel upper head was selected. The calculated response for this event is relatively insensitive to the RCP trip sequence due to the break size and location. Parametric calculations for both trip sequences produced the same results for this event.
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l Ouestion 440.223:
It is stated in the conclusion of the SBLOCAs (page 77) that no core uncovery is calculated for the cases analyzed. Figure 3.8-21 shows that .
for 3" break LOCA, the collapsed water level falls up to 4 feet below the l' top of fuel for about 60 seconds. Clarify the inconsistency of the core covery in your conclusion and Figure 3.8-21, and demonstrate that the core l coolability can be maintained for the 3" break case.
Response 440.223:
The occurrence of core uncovery is based on the calculated two-phase mixture level in the inner vessel node not on the collapsed liquid level.
Refer to Figure 3.8-20 (actually mislabeled as 3.9-20 in the report) for the mixture level versus time graph. The collapsed liquid level graph is usually included in the results so that it can be compared to the two-phase mixture level graph. This comparison shows the void fraction in the L core two-phase region or the extent of core boiling or steam production being calculated during the transient. The collapsed liquid level is a ,
hypothetical level based on instantaneously releasing all the steam from the two-phase region and therefore does not indicate any true level of uncovery.
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Ouestion 440.225: ;
In the I&C diversity analysis, best estimate parameters were assumed for plant initial conditions. Provide the technical basis for the best estimate input parameters for the events analyzed in the report.
Response 440.225:
The best estimate input parameters and initial conditions assumed in the I&C diversity analyses for the events presented in the report are based on the design and operating conditions as follows:
(a) The pressurizer initial pressure of 2250 psia and the core inlet (cold leg) coolant temperature of 555.8 F are at the nominal values per CESSAR-DC Figures 5.4.10-5 and 5.4.10-3, respectively, for all events analyzed. ;
(b) The initial core thermal power of 3914 MWt is the design total core heat output, per CESSAR-DC Tables 4.4-2 and 4.4-9, for all events analyzed.
(c) For all events except LOCA, FWLB and CEA Ejection, the integrated radial peaking factor F, for the unrodded core is assumed at the design best estimate and a first core analysis maximum value of i 1.50. The radial peak has no significant impact on the LOCA and FWLB containment analysis results and, therefore, is not modelled.
The CEA Ejection event analysis used the pre- and post-ejected r peaking factors that were shown to be the most limiting for the Chapter 15 safety analysis.
(d) The axial shape index of -0.07 was assumed as the best estimate ;
value consistent with the axial power distribution typical for the j first core for all events except LOCA and FWLB. The LOCA assumed a minimum full power value of -0.3 per Technical Specifications, to yield more conservative results. The axial shape has no impact on '
the FWLB containment analysis results.
(e) The RCS flowrate assumed for all events other than LOCA is discussed in Response 440.221. A low value for the initial RCS flow rate was conservatively assumed for LOCA. The FWLB event was re-analyzed utilizing 461,200 gpm value for containment analysis (see CESSAR-DC). The FWLB results are not significantly sensitive to the value of the initial RCS flowrate.
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e Ouestion 440.226 The applicant is requested to discuss indications and controls utilized to respond to events and their diversity rationale (including the normal '
instrumentation and control systems). The discussion should cover each event presented in the report included in an ABB-CE letter, LD-93-083, and its References 2 and 6.
Response 440.226 ,
Section :.0 of ALWR-IC-DCTR-31 discusses the allocation of diverse '
software to the I&C equipment used for protection, control, and information displays in NUPLEX 80+. Diverse equipment, as discussed in Section 3.0 of ALWR-IC-DCTR-31, use diverse microprocessors, and diverse operating systems in the microprocessors, and the applications software is ,
developed by independent teams. !
1 The control and information display equipment which remain available with >
the postulated common mode software failure are identified in Section 2.0 of ALWR-IC-DCTR-31. The implications of the common mode failure on the ;
normal plant response and the indications and controls which remain '
available for coping are presented for each event in Section 4 of ALWR-IC- l
~
DCTR-31.
Discussions with the NRC staff subsequent to the Lawrence Livermoore review of the ALWR-IC-DCTR-31 evaluation determined that the capability of the diverse equipment to provide adequate protection had been demonstrated l for 19 of the 28 event initiators in Chapter 15. Subsequent discussion of :
the evaluation with NRC management determined that a revised evaluation ;
would be appropriate for the remaining 9 events, and would apply more l relaxed criteria than those applied in Chapter 15 and credit use of manual '
controls implemented in the design to comply with position 4 of the NRC l policy statement on common mode failure. !
In Section 2.2 of the revised evaluation (LD-93-080), a more detailed l description is provided of the instrumentation available to the operator. i In Sections 2.3 and 2.4 of LD-93-080, dated May 19, 1993, the specific j information available to support credited operator actions is identified ;
for each of the 9 events. i l
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e Question 440.227 It is indicated in Section 2.3, the determination of the required event diagnostic time of one minute is based on ATWS scenarios. According to CE-EPGs, the ATWS event diagnosis requires verification on the post-trip rod position as the sole diagnostic step. However, a common mode failure (LMF) event involves more complications which include conflicting indications and inoperable controls in combination with a plant transient.
The applicant is requested to provide bases that justify the response time of one minute as sufficient for the CMF event diagnosis.
Response 440.227 In discussing operator initiation of a manual reactor trip, Section 2.3 of LD-93-083 reports that the data discussed in the Appendix to ANS-58.8 and in Reference 8 of LD-93-083 indicate that, in the response time data determined for ATWS scenarios, the earliest time for operator action is typically less than 1 minute. The discussion in Section 2.3 goes on to identify the specific indications available to the operator to prompt his manual initiation of a reactor trip. The multiple familiar indications provided to the operator of the need for a reactor trip, the availability of the DIAS-P indications to confirm that a trip had not occurred and the empirical response data for ATWS scenarios provide the basis for concluding that operator action to initiate a manual reactor trip within two minutes of reaching an alarmed trip condition is considered reasonable for the purposes of the beyond design basis analysis provided in LD-93-083.
Section 2.3 of LD-9.3-083 also discusses the bases for response times considered reasonable for operator actions performed as part of the Standard Post-Trip Actions, i.e., subsequent to a reactor trip initiated manually or by the Alternate Protection System. That discussion identifies that if a step in the EPG involves simple verification of parameters which are readily available on a normal DPS display and can be verified on DIAS-P, and the value can be expected to be within the acceptable range, then that step in the EPG is estimated to take 1 minute per manipulation, which is consistent with the ANS-58.8 model for performing familiar actions.
Additional time was allocated in the response time estimates for indications which could not be readily verified and for actions for which the normal control interface was assumed inoperable. During the first nine minutes of the respor.se sequence, four minutes are allocated for the supervisor and two operators to identify that a global problem has occurred in the DIAS-N displays and to determine that the DPS displays
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Response 440.227 (continued) should be used instead. This is reasonable since there will be multiple alarms provided by the DPS displays and IPS0 indicating a significant plant problem, and the condition could be confirmed via the DIAS-P ,
displays. Since the DPS would alarm discrepancies between parameter values validated independently by the DPS and DIAS, and since the DIAS-N implements a different type of display device than used by the DPS and each DIAS-N and DPS display panel displays a rotating icon to indicate if the information on the screen is being updated, a system wide DIAS-N failure would be readily evident.
For actions for which the normal control interface would be inoperable, e.g. for ESF equipment actuation, additional time was allocated for the operator to determine that the manual initiation had not taken effect and then to perform manual initiation using the diverse manual actuation switches.
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l Ouestion 440.228 l
The applicant is requested to confirm that CE-EPGs are adequate to guide j operators for response to CMF events. If the applicant finds that the EPGs revision is necessary for the CMF event mitigation, the revised EPGs should be provided for the staff to review.
Response 440.228 ,
The EPGs do not need to be revised to support the assumptions made for operator responses in the LD-93-083 evaluation. The EPGs specify the parameters or information the operator is to use to establish the status of critical functions, and they specify the system or component responses '
to be actuated if required. The DPS provides alarms to indicate when ;
discrepancies are found between parameter values validated by the DPS and i those validated independently by DIAS. The evaluation assumes that the operators will have guidance to use the DIAS-P displays to verify either DIAS-N or DPS indications under circumstances where they are inconsistent, and to use the diverse manual actuation switches if they are unable to i establish that successful actuation of the associated equipment has occurred when using the ESF-CCS.
The LD-93-083 evaluation demonstrates that the implementation of diversity in System 80+ is sufficiently comprehensive for coping with common mode :
failures, that even if a failure were postulated to disable all of the protective functions normally performed by the PPS, the ESF-CCS and the monitoring capability of the DIAS; the plant would retain a sufficient combination of automatic controls, manual controls and indications to provide adequate protection even for low probability limiting fault events. ABB-CE does not consider such a scenario to be credible and, therefore, does not consider it appropriate as a design basis for the plant protective systems. Other aspects of the NUPLEX 80+ design assure the reliability of the protective systems and the extremely low probability of an impairment of their protective functions by any common mode failure. Most important among these is the emphasis on simplicity in the digital technology used for protective function actuation and control and the rigor of the verification and validation process. However, metrics are not currently available in the industry to quantify the benefit of this approach. Therefore, the LD-93-083 evaluation was performed as a demonstration of the extensive protection provided by the implementation of diversity in System 80+. Consideration of the other aspects of the NUPLEX 80+ design, in combination with this demonstration of the extensive benefit provided by diversity establishes the acceptability of the NUPLEX 80+ design for addressing the potential impact of common mode software failures.
,