ML20045B333
| ML20045B333 | |
| Person / Time | |
|---|---|
| Site: | 05200002 |
| Issue date: | 06/11/1993 |
| From: | Mike Franovich Office of Nuclear Reactor Regulation |
| To: | Miraglia F, Murley T, Russell W NRC |
| References | |
| NUDOCS 9306170202 | |
| Download: ML20045B333 (10) | |
Text
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UNITED STATES
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June 11,1993 Docket No.52-002 MEMORANDUM FOR:
T. Murley G. Lainas R. Zimmerman F. Miraglia J. Roe B. Boger W. Russell J. Zwolinski R. Gallo J. Partlow E. Adensam F. Congel D. Crutchfield B. Grimes E. Butcher W. Travers J. Richardson W. Bateman, EDO A. Gody B. D. Liaw A. Vietti-Cook S. Varga A. Thadani Operations Center J. Calvo M. Virgilio C. Rossi THRU:
R. W. Borchardt, Acting Director Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation FROM:
Michael X. Franovich, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
SUBJECT:
DAILY HIGHLIGHT - FORTHCOMING MEETING WITH ABB-COMBUSTION ENGINEERING (ABB-CE) FOR SYSTEM BO+ (REACTOR SYSTEMS OPEN ISSUES) t DATES AND June 15, 1993 (10:15 a.m. - 6 p.m.)
TIMES:
June 16, 1993 (8:30 a.m. - 5 p.m.)
LOCATION:
ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor, Connecticut 06095-0500 PURPOSE:
Discussion of remaining open issues from the CE System B0+
draft safety evalaution report (DSER) in the reactor systems area. contains issues related to the power upgrade for CE System 80+ that will also be discussed.
A proposed agenda is provided in Enclosure 1.
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. June 11, 1993 PARTICIPANTS *:
NRC ABB-CE M. Franovich S. Ritterbusch M. Rubin J. Longo S. Sun M. Cross D. Diec F. Carpentino M. Volodzko J. Rezendes
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Michael X. Franovi~ch, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures:
See next page
- Meetings between the NRC technical staff and applicants or-licensees are open for interested members of the public, petitioners, intervenors, or other parties to attend as observers pursuant to "Open Meeting Statement of NRC Staff Policy," 43 Federal Reaister 28058, S/28/78. Members of the public who wish to attend should contact me at_(301) 504-1121.
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Docket File PDST R/F SBajwa, 12G18 RBorchardt PDR TEssig TVWambach MXFranovich PShea OPA EJordan, MNBB 3701 GGrant, ED0 ACRS (11)
P0' Dell, PTSB JMoore, 15B18 GBagchi, 7H15 NRR Mailroom, 12G18 AChaffee, EAB MRubin, 8E23 SSun, 8E23 AAttard, 8E23 SMagruder RJones, 8E23
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i ABB-Combustion Engineering, Inc.
Docket No.52-002 cc:
Mr. C. B. Brinkman, Acting Director Nuclear Systems Licensing ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Manager Washington Nuclear Operations i
ABB-Combustion Engineering, Inc.
12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Licensing ABB-Combustion Engineering 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.
20585 Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.
Washington, D.C.
20503 Mr. Raymond Ng 1776 Eye Street, N.W.
Suite 300 Washington, D.C.
20006 Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge r
2300 N Street, N.W.
Washington, D.C.
20037-1128 Mr. Regis A. Matzie, Vice President Nuclear Systems Development ABB-Combustion Engineering, Inc.
1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500-i l
CE System 80+
Reactor Systems DSER Open Items Meeting June 15-16,1993 Windsor, Connecticut i
Tuesday. June 15.1993 i
10:15 - 10:30
Introductions
10:30 - 12:00 Discuss DSER issues that the staff considers technically resolved and begin discussion of remaining DSER open items 12:00 - 1:00 Lunch 1:00 - 6:00 Continue discussion on DSER open issues:
- Intersystem LOCA followrp/ status (feedback on heat exchanger submittal)
- Instrumentation and controls diversity (start as time permits) t Wednesday. June 16.1993 1
08:30 - 9:30 Instrumentation and controls diversity (Enclosure 2 questions) continued I
i 9:30 - 12:00 Shutdown risk follow up and staus 12:00 - 1:00 Lunch 1:00 - 5:00
- Power upgrade questions (Enclosure 2)
- Nitrogen 16 main steam line monitors (Enclosure 2) 4
- RCS boron dilution status (time permitting) i i
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CE SYSTEM 80+
Ouestions on CESSAR-DC Chapters 6 and 15 Reanalysis:
440.197 Technical Specifications (TSs)
Provide the revised TSs that reflect the assumptions and results of Chapters 6 and 15 reanalysis.
440.198 Figures 6.3.3.2.11 and 6.3.3.3.12 I
Provide the results of sensitivity study on reanalyses for the break size spectrum to show that the limiting cases are the indicated DEG/PD for large break (LB) loss-of-coolant accidents.
l (LOCAs) and the break of 0.1 ftz for small break (SB LOCAs).
440.199 Table 6.3.3.2-2 Why are the core and the reactor coolant system (RCS) flow rates assumed to be the same?
440.200 Tables 6.3.3.2-2 and 6.3.3.3-2 Why are the initial core outlet temperatures different for LB and SB LOCAs?
440.201 Table 6.3.3.3-2 Table 6.3.3.2-2 shows that the calculated maximum peak clad temperature (PCT) for LB LOCAs is at the fuel burnup of 26,000 MWD /MTU. Confirm whether the calculated maximum PCT for.
SB LOCAs also depends on fuel burnups.
Include in Table 6.3.3.3-2 the limiting burnup resulting in the highest PCT for SB LOCAs.
440.202 Assumptions I, Page 6.3-35 Provide the basis for the boric acid precipitation of 27.6 percent at the containment pressure of 14.7 psia.
440.203 Table 6.3.3.4-1 Provide an explanation for the meaning of the note for shutdown cooling system (SCS) entry conditions (temperature and pressure) with consideration of the associated instrumentation errors.
440.204 Figure 6.3.3.4-5 Provide analytical results to show that for SB LOCAs with break z
sizes smaller than 0.01 ft, 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 draft safety evaluation report (DSER) Open Items 6.3.3-1 (see ABB-CE letter, LD-93-048, dates March 17,1993.)
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440.205 RCS Flow Rates The design RCS flow rate is 444,650 gpm as 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 1
and revise the standard safety analysis report (SSAR) reflecting the correct RCS flow rate.
440.206 The initial departure from nucleate boiling ratios (DNBRs) for the transients are significantly higher for the reanalyzed cases than those in the existing Chapter 15 analyses.
For example, the initial DNBR for the inadvertent opening of a steam generator (SG)
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atmospheric dump valve (ADV) increased from 1.24 to 1.36, 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.
Please provide the technical basis for the significant increase in initial DNBRs for all the events reanalyzed.
440.207 Steam Line Break (SLB) - Control Rod Worth The control rod worth of 10 percent was used in the SLB analysis.
This value is significantly increased (from 8.86 percent used in the previous analysis).
Provide the basis for the increase.
Clarify whether the change is due to the control rod design change or the calculational method change for the control rod worth.
440.208 Steam Line Break - AFW Actuation (Page 15.1-15) i The staff agrees that an early actuation of the auxiliary feedwater (AFW) will maximize the cooldown effect and is a conservative assumption for the post-trip SLB analysis.
For an SLB with a loss-of-offsite power (LOOP), assuming failure of a 1
main steam isolation valve (MSIV) 'n the intact SG to close, a 1
delay of AFW actuation could result in a complete depletion of the i
l water inventory from both SGs. Under this condition, the injection of cold AFW could cause significant thermal stress on SGs and potentially result in damage to the SGs. The applicant is requested to provide an analysis showing that a delay of AFW actuation (such as AFW on an automatic mode) will not result in a complete depletion of inventory from both SGs and a complete loss i'
of SG heat removal 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.
440.209 Water Level Instrumentation Errors i
In the analysis, the AFW is actuated on the low SG water level j
signal when the automatic mode is assumed.
In the SG tube rupture (SGTR) analysis, the high SG water level trip signal is credited for the reactor trip.
The water level instruments are subject to significant measurement errors. The applicant is requested to j
provide an assessment of effects of water level instrumentation errors on the results of the safety analysis, which rely on the i
level instruments for accident mitigation.
440.210 Loss-of-Condenser Vacuum (LOCV) (Figure 15.2.3-13)
Explain why the calculated DNBR decreases twice before it substan-tially increases.
440.211 Feedwater Line Break (FLB) - Calculated Peak RCS Pressure The calculated peak RCS pressure is 2785 psia on page 15.2-15 and is 2720 psia in Table 15.2.8-2.
Clarify the inconsistency.
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440.212 FLB - DNBR Itisstatedonpage15.2-16thattheminimumDNBR(MDNBR)js below 1.24 for the limiting FLB with a break size of 0.2 ft. The j
applicant is requested to provide the calculated DNBRs as a function of time for the limiting FLB case.
440.213 Table 15.2.8-1: Initial Conditions Provide the criterion for selecting initial conditions in order to maximize the peak RCS pressure resulting from the limiting FLB case.
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440.214 Figure 15.2.8-1 Explain what causes the peak RCS prpssure to decrease and then 2
increase for break sizes from 0.7ft to 0.9 ft.
440.215 Locked Rotor - Initial Conditions
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On page 15.3.8 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 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 initial DNBR is assumed at a higher value.
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. Also, different values of the maximum charging flow rates (250 gpm on page 15.6-10 and 180 gpm on page 15.6.26) are i
assumed for the steam generator tube rupture (SGTR) analysis.
l Clarify these discrepancies.
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j 440.217 Page 15.5-6 It is stated that "a pressurizer absolute high level alarm at.
65 percent of the pressurizer volume will prevent water from being discharged out of the pressurizer safety valves." Confirm that the appropriate emergency procedure guidelines (EPGs) are available, and describe the operator actions and timing to avoid water discharge from the pressurizer safety valves (PSVs) in responding to the alarm.
t 440.218 Safety Grade Components i
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 safety grade components and receive power from emergency at sources.
440.219 Results Comparison The staff finds that the calculated minimum DNBRs for most tran-sients reanalyzed are higher than that of the existing analysis (Amendment H) even though the rated power was increased by 3 per-cent, a LOOP was assumed without event recategorization, and the LOOP delay time was assumed to be zero seconds.
The applicant should identify the changes in system parameters and calculational methods for each case which were reanalyzed and assess the effects of the changes on the safety analysis results.
For each change that affects the safety analysis results, appropriate technical justifications should be provided.
Reauests for Additional Information on Instrumentation & Control (I&Cl Diversity Analysis (LD-93-080):
440.220 In the I&C diversity analysis, the moderator and Doppler reactivity feedback functions are the main parameters to control the core power 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.
l 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 the flow rates and revise, if
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necessary, the analysis to reflect the correct flow rate consistent with the best estimate method proposed in the analysis.
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-inch LOCA initiation (page 75).
For the SB LOCAs of 3 inches 2
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 SB LOCAs.
440.223 It is stated in the conclusion of the SB LOCAs (page 77) that no core uncovery is calculated for the cases analyzed.
Figure 3.8-21 _
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shows that for the 3-inch break LOCA, the collapsed water level falls up to 4 feet below the 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 coolability can i
be maintained for the 3-inch break case.
440.224 To enhance defense in depth, ABB-CE is requested to provide an i
evaluation of the LB LOCA and MSLB inside the containment events, i
and to identify the time available for the operator to respond to these events.
440.225 In the I&C diversity analysis, best estimate parameters were assumed for plant initial conditions.
Provide the technical basis i
for the best estimate input parameters for the events analyzed in the report.
i 440.226 The applicant is requested to discuss indications and controls utilized to respond to events and their diversity rationale (including the normal I&C 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.
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 ABB-CE's emergency procedure guidelines (EPGs), the ATWS event diagnosis requires verification on the post-trip rod position as the sole diagnostic step.
However, a common mode failure (CMF) 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.
440.228 The applicant is requested to confirm that ABB-CE's EPGs are adequate to guide operators for response to CMF events.
If the applicant finds that the EPGs revision is necessary for the CHF event mitigation, the revised EPGs should be provided for the staff to review.
Other Issues:
440.229 During a recent NRC senior management trip to Europe, information was received on a system to be implemented in the Dutch "Borsselle" PWR, which relies on N-16 monitors on the main steam lines to initiate automatic actions during an SGTR event. The system operates at power >35 percent, and upon sensing N-16 in a main steam line, trips the reactor and initiates pressurizer spray to reduce primary system pressure. The high-pressure safety injection is initiated in this plant only on coincident low-r pressurizer level and low-pressurizer pressure; since the N-16 system reduces pressure while maintaining pressurizer level, high-pressure safety injection is inhibited, and the threat of SG overfill is reduced. Has ABB/CE considered such design features for System 80+7 If not, provide an evaluation of the benefits and downsides of such a design in limiting SGTR consequences.
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440.230 Existing experimental evidences and recent analysis results show that an inherent mechanism for boron dilution in the PWR reactor coolant pump (RCP) loop seals could exist for events (including SB LOCAs) that involve heat removal by reflux / boiler condensation during natural circulation. The deborated water in the RCP loop seals could be transported to the core through natural circulation processes or startup of RCPs.
Injection of the deborated water into the core has a potential risk that a reactivity induced accident could occur and result in damage to the core. The applicant is requested to address the applicability of this boron dilution event to the System 80+ design and provide resolutions to this issue if necessary.
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