ML20113H546
| ML20113H546 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 12/31/1984 |
| From: | Gridley R, Rogers A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML112241107 | List: |
| References | |
| NEDO-30603-1, NEDO-30603-1-R01, NEDO-30603-1-R1, NUDOCS 8501250202 | |
| Download: ML20113H546 (101) | |
Text
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NEDO 30603-1 REVISION 1 CLAS,SI l DECEMBER 1984 l
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l DUANE ARNOLD ENERGY CENTER POWER UPRATE 1
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0501250002 850
, I' O R ADOCK 0500 1 GENERAL ELECTRIC .'
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NEDO-30603-1 Revision 1 Class I December 1984 CENERAL ELECTRIC COMPANY DUANE ARNOLD ENERGY CENTER POWER UPRATE
/
. Approved: // ^ Approved M '~
'54t. L. Gridley, Aanager A.E. Rogers, Manager Application Engineering Fuel and Services Licensing Services Nuclear Safety and Licensing Nuclear Services Products Operation Department NUCLEAR ENERGY BUSINESS ff'[A J 1NS
- GENERAL ELECTRIC COMPANY
' S/ , a<d ; C o 'ORNIA 95125 GENERAL $ ELECTRIC
I o .
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for the use of Iowa
. Electric Light and Power Company. The information contained in thic report is
. believed by General Electric to be an accurate and true representation of the facts known,-obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting informa-tion in this document are contained in the Duane Arnold Energy Center (DAEC)
- Power Uprating Program Phase II Proposal No. 423-TY645-KEl (GE Letter No.
C-KE-3-084, dated July 22, 1983) and nothing contained in this document shall be construed as changing the under:akinga in that proposal. The use of this information except as defined by said proposal, or for any purpose other than-that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the Con-tributors to this document makes any representation or warranty (express.or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such-information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use'of such
-information.
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NED0-30603-1 CONTENTS
- 2. age, ABSTRACT xi
- 1. INIRODUCTION 1-1
- 2. HEAT BALANCE 2-1
- 3. POWER / FLOW MAP 3-1
- 4. INSTRUMENT SETPOINTS 4-1 4.1 Reactor Vessel High-Pressure Scram Setpoint 4-1 4.2 Recirculation Pump High-Pressure Trip Setpoint 4-2 4.3. Safety Relief Valve Opening Pressure Setpoints 4-2 4.4 Other Instrument Setpoints 4-3
-5. LOSS-OF-C00LANI ACCIDENT ANALYSES 5-1 5.1 Introduction 5-1 5.2 Acceptance Criteria for ECCS Performance 5-1 5.3 Input to Analysis 5-2 5.4 LOCA Analysis Models 5-3 5.5 Break Spectrum Results 5-3 5.6 LOCA Analysis Conclusions 5-4
- 6. REACTOR VESSEL OVERPRESSURE PROTECTION 6-1
- 7. TRANSIENTS AND ACCIDENTS 7-1 7.1 Transients 7-1 7.2 Rod Withdrawal Error 7-1
.7. 3 Control Rod Drop Accident 7-2
- 8. MCPR' LIMITS 8-1 8.1 Applicability of Current SLMCPR and Kr Curves 8-1 8.2 .Effect of Power Level Uncertainty on MCPR Operating Limit 8-2 8.3 MCPR Operating Limit 8-4
- 9. STABILITY 9-1 9.1 Channel Hydrodynamic Conformance to the Ultimate Performance Criterion 9-1 9.2 Reactor Conformance to the Ultimate Performance Criterion 9-1
.10. REACTOR INTERNAL PRESbURc DIFFERENCES / STRUCTURAL ZvALUATION 10-1 10.1 Reactor Internal Pressure Differences Analysis 10-1 in.? Structural Evaluation of DAEC at Uprated Power 10-3
- 11. ' CONTAINMENT EVALUATION 11-1 11.1 Short-Term Accident Response Analysis 11-1 11.2 Long-!erm Accident Response Analysis 11-2 iii-
I.
NEDO-30603-1 CONTENTS (Continued)
ESK'
- 12. IMPACT ON PLANT IMPROVEMENT PROGRAMS 12-1 12 . l~- Extended Load Line Limit Analysis 12-1 12.1.1 Introduction' 12-1 12.1.2 Analyses and Results 12-1 12.1.3 Conclusion 12-3 12.2 SINGLE-LOOP OPERATION 12-3 12.2.1 Loss-of-Coolant Accident 12-4 12.2.2 Stability 12-4 12.2.3 Transients and MCPR Limits 12-4 12.3 SRV LOW-LOW SET SYSTEM 12-5
- 13. SDDIARY AND CONCLUSIONS 13-1
- 14. REFERENCES 14-1 iv
NEDO-30603-1 TABLES Table- Title Page 12-1 Comparison of DAEC Reactor Heat Balance Parameters for Rated Power and Uprated Power cases 2-2
.5-l' LOCA Analysis Figure Summary 5-5 5-2a: MAPLHGR Versus Average Planar Exposure
.(P8DRB30lL) 5-6 5-2b MAPLHGR Versus Average Planar Exposure
.(P8DPB289) 5-6 5-2c! .MAPLHGR Versus Average Planar Exposure (P8DRB284H) 5-7 5-2d MAPLHGR Versus Average Planar Exposure (P8DRB299) 5-7
- 5-3 Summary of Break Spectrum Results '5-8 5-4 Significant. Input variables Used in the Loss-of-Coolant Accident Analysis 5-9 5-5
- Single-Failures Evaluated 5-12 6 . Transient Output Data Summary ~6-3 7-1 Core-Wide Transient Results 7-3 7-2 ' Local Rod Withdrawal Error (With Limiting Instrument Failure) Transient Susmary 7-4
,10 Comparison of Steady-State Operating RIPDs for Rated.
Power and Uprated Power Conditions 10-4 10-2 Comparison of Current Design RIPD Values Against New RIPD Values at Uprated Power 10-5 1: Comparison of DAEC Containment Responses to DBA for . , .-
. Rated Power and Uprated Power Conditions .11-3
.12-1 Comparison of LOCA Reflooding Characteristics for Single-Esop and two-Loop Operatior. 12-6 12-2' Summary of LLS Analysis Results for Limiting
. Transients Events (102% of Uprated Power)
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12-6 v/vi
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NED0-30603-1 ILLUSTRATIONS Figure Title Page 2-1 DAEC Reactor Heat Balance At Uprated Power 2-3 2-2 DAEC Reactor Heat Balance At 102% of Uprated Power 2-4 3-1 DAEC Power / Flow Map 3-2 5-la Water Level Inside the Shroud and Reactor Vessel Pressure Following a Design Basis Accident. LPCI Injection Valve Failure 5-13 5-lb Water Level Inside the Shroud and Reactor Vessel Pressure Following a Break of the Recirculation Line LPCI Injection Valve Failure, Break Area = 1.0 ft2 (Transition Break-Large Break Method) 5-14 lc Water Level Inside the Shroud and Reactor Vessel Pressure Following a Break of the Recirculation Line, LPCI Injec-tion Valve Failure, Break Area = 1.0 ft2 (Transition Break-Small Break Method) 5-15 5-ld Water Level Inside the Shroud and Reactor Vessel Pressure Following an Intermediate Break of the Recirculation Line, LPCI Injection Valve Failure, Break Area = 0.8 f t2 5-16 5-le Water Level Inside the Shroud and Reactor Vessel Pressure Following a Small Break of the Recirculation Line HPCI Failure, Break Area = 0.07 ft2 5-17 5-2a Peak Cladding Temperature Following a Design Basis Accident.
LPCI Injection Valve Failure 5-18
. 5-2b Peak Cladding Temperature Following a 1.0 ft2 Break of the Recirculation Line, LPCI Injection Valve Failure (Transition Break-Large Break Method) 5-19 5-2c Peak Cladding Temperature Following a 1.0 ft2 Break of the Recirculation Line, LPCI Injection-Valve Failure (Transition Break-Small Break Method) 5-20 5-2d Peak Cladding Temperature Following an Intermediate Break of the Recirculation Line, LPCI Injection Valve Failure, Break Area = 0.8 ft 5-21 5-2e Peak Cladding Temperature Following a Small Break of the Recirculation Line HPCI Failure Break Area = 0.07 ft2 5-22 vii
e-NED0-30603-1 ILLUSTRATIONS (Continued)
Figure Title Page 5-3a- Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for DBA 5-23 5-3b Fuel Rod Convective Heat Transfer Coefficient Duying Blowdown at the High Power Axial Node for 1.0 f t Break 5-24 5-4a Normalized Core Average Inlet Flow Following a Design Basis Accident 5-25 5-4b Normalized Core Average Inlet Flow Following a 1.0 ft
. Break 5-26 5-5 Peak Cladding Temperature Versus Break Area 5-27 6-1 .MSIV Closure with Flux Scram, Uprated Power 6-4 9-1 Reactor Core Decay Ratio vs Power 9-2 10-1 Reactor Internal AP Locations 10-6
- 10-2 Transient Pressure Differentials Following (Emergency Condition) at Uprated Power, Recirculation riov - Part A 10-7 10-3 Transient Pressure Differentials Following (Emergency Conditions) at Uprated Power, Recirculation Flow - Part B 10-8 10-4 Transient Pressure Differentials Following (Faulted Condition) at Uprated Power, Recirculation Flow - Pait A 10-9 10-5 Transient Pressure Differentials Following (Faulted Condition) at Uprated Power, Recirculation Flow - Part B 10-10 10-6 Transient Pressure Differentials Following (Faulted Condition) at Uprated Feedwater Flow, Recirculation Flow - Part A 10-11 10-7 Transicut Pressure Differentia 1c Follo,ing (Faulted Condition) at Uprated Feedwater Flow, Recirculation Flow - Part B 10-12
- General Electric Company Proprietary Information has been deleted.
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m:
. ILLUSTRATIONS (Continued)
Figure Title Pm 11-1 Containment Pressure Versus Time After Event DAEC Recircu-lation Suction Line DEGB at Uprated Power 11 11-2 Short-Term Containment Temperature Versus Time After Event, DAEC Recirculation Suction Line DEGB at Uprated Power 11-5 11-3 (General Electric' Company Proprietary Information) 11-6 ix/x L
e NEDO-30603-1 ABSTRACT The Iowa Electric Light and Power Company (IELP) has implemented the-Duane Arnold Energy Center-(DAEC) Power Uprate Program with the objective of cperating DAEC at the full licensed power of 1658 MWt. Changes to the plant Technical Specifications will be necessary to attain this objective. Results of Nuclear Steam Supply System (NSSS) and containment systems safety analyses are: presented. These results demonstrate that the revised Technical Specifi-cations will satisfy the established DAEC licensing criteria, when supported by appropriate auxiliary systems analyses, and that the plant should operate at
- 1658 MWt without undue risk to the public health and safety. These NSSS-and containment systems. analyses include the following:
- a. Plant heat balance at 1658 MWe and 1691 MWt b .- Power / flow map at 1658 MWt
- c. Instrument setpoints at 1658 MWt
- d. Loss-of-coolant accidents at 1691 MWe
- e. Reactor vessel overpressure protection at 1658 MWt g f. Ainormal operational transients at 1658 MWt "g. Control rod drop accident
- h. MCPR operating limits at 1658 MWt
- i. Stability corresponding to operation at 1658 MWt
- j. Reactor internal pressure differences at 1658 MWe and 1691 MWt
- k. Short- and long-term containment response to accidents at.1691 MWt
- 1. . Extended load line limit operation at 1658 MWt
.m. Single-loop operation
- n. Impact'on the Low-Low Set System-The results of~the analyses are compared to the established DAEC licens-ing criteria. When the accompanying Technical Specification changes are impler ated, 4tl-the criteria will be e:6Asfied.
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- 1. INTRODUCTION The original design basis power level of the Duane Arnold Energy Center (DAEC) was 1658 MWe, which represents an approximate 4% margin over the original commercial basis (or rated power) of 1593 MWt. This higher design basis power is reflected in the DAEC operating license which was granted at 1658 MWt and by the plant safety analyses, including reloads, which are per-formed at 1658 MWe, although the Plant Technical Specifications have restricted operation to the " rated power" of 1593 MWt.
This unused plant capacity and licensing basis presents an opportunity to increase the generating capacity of the Iowa Electric Light and Power Company (IELP) System with no change in calculated safety margins.
This report presents the results of analyses of the DAEC Nuclear Steam Supply System (NSSS) and concainment systems necessary to demonstrate safe operation of the plant at the design basis power level of 1658 MWt. These results, when supplemented by verification analyses of plant auxiliary systems, provide justification for changing the Technical Specification maximum allow-able power level to 1658 MWt.
The entire IELP program, with the ultimate objective of increasing the Technical Specification thermal power limit to 1658 MWt, is known as the DAEC Power Uprate Program. In this report, the licensed or design basis power of-1658 MWt will be referred to as the uprated power, and the current power level of 1593 MWe will be referred to as the rated power. The DAEC Technical Speci-fications will be revised to redefine rated power (100%) as 1658 MWt.
This increase in thermal power level should add approximately 4% to the net electrical output of DAEC at small incremental cost and should provide sig-nificant economic benefits to the IELP rate payers.
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1 NEDO-30603-1 l 1
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.lium scope of the NSSS and containment systems analyses, which support the increase in thermal power, was determined from reviews of the original plant licensing documentation (Reference 1), the Standard Review Plan (Refer-ence 2), the updated FSAR (Raference 3), and selected license amendment sub-mittals including reloads (Reference 4).
These analyses include the following:
- a. A plant heat balance at 1658 MWt and 1691 MWt to define steady-state operating parameters and to provide inputs and initial conditions
-for the plant safety analyses.
- b. A new power / flow map which reflects the impact of power uprate.
- c. Determination of a new reactor high-pressure scram setpoint and eval-uation of impact on safety analyses.
- d. Determination of a new anticipated transient without scram (ATWS) recirculation pump trip (RPT) high-pressure trip setpoint and eval-uation of impact on transient analyses.
- e. Determination of new safety relief valve (SRV) setpoints, demonstra-tion of compliance with ASME code requirements and evaluation of impact on safety analyses.
- f. Updated loss-of-coolant accident (LOCA) analyses at 1691 MWt incor-porating the increased number of fuel bundles with drilled lower tie plate holes added since the last LOCA analysis was performed.
- g. New reactor vessel overpressure protection analysis at 1658 MWt reflecting the new SRV setpoints and the higher operating pressure which demonstrates compliance with the ASME Code requirements.
- h. Abnormal operational transient analyses performed to support the Cycle 8 reload. f i
n 1-2
- 1. Control rod drop accident (CRDA) analysis reflecting core character-istics due to operation at 1658 MWt.
- j. Minimum critical power ratio (MCPR) operating limits, including con-firmation of the validity of the 1.07 safety limit for power uprate.
- k. Stability analysis conservatively reflecting the higher power follow-ing two pump trip from the average power range monitor (APRM) rod block line.
- 1. Reactor internal pressure differences (RIPD) analyses at 1658 MWt and 1691 MWt and structural evaluation of reactor internals, demonstrat-ing that design stresses will not be exceeded.
- m. Short- and long-term containment responses to accidents, demonstrat-ing that drywell and wetwell temperature and pressure limits are not violated, including the effect of a 15% raduction in the residual heat removal (RHR) service water flow rate.
- n. Extended load line limit analysis (ELLLA) based on the uprated power, demonstrating that the consequences of transients and accidents ini-tiated from the ELLLA region are bounded by the consequences of tran-sients and accidents initiated from the uprated condition.
- o. Evaluation of single-loop operation (SLO) demonstrating that current
. SLO restrictions are adequate for SLO with power uprate.
- p. Evaluation of impact of power uprate on SRV cycling frequency, demonstrating that the time between closing and reopening of valves is within the established criteria, i.e., verification of low-low sr.L logic.
The results of the NSSS and containment systems safety analyses are com-pared to established DAEC licensing criteria. In all cases, the criteria are satisfied when the accompanying Technical Specification changes are implemented.
1-3/1-4
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e N EDO-30603-1
- 2. HEAT BALANCE The plant heat balance defines the initial conditions and input parameters for the plant safety analyses. To determine the necessary initial conditions and input parameters, heat balance analyses of the DAEC at the uprated power level of 1658 MWt and at 102% of the uprated power level (1691 MWt) were per-formed. The analyses relate the reactor thermal / hydraulic parameters to the plant steam and condensate flow conditions at 1658 MWt and at 1691 MWt. The reactor heat balance parameters were used as input parameters in various analyses described in this report, including LOCA, reactor pressure vessel overpressure protection, thermal limits, reactor internal pressure differences, and containment evaluation.
The reactor heat balance data for the DAEC at the uprated power level of 1658 MWt is given in Figure 2-1. A comparison of the reactor heat balance parameters at 1593 MWt (rated power condition) with those at the uprated power conditions is also shown in Table 2-1. The reactor heat balance parameters used in the LOCA analyses and containment evaluations (102% of uprated power, 1691 MWt) are given in Figure 2-2.
Uprating the DAEC thermal power from 1593 MWe to 1658 MWt will require an increase'in the reactor dome pressure from 1020 psia to 1040 psia. This increase in reactor pressure will compensate for the additional pressure drop caused by the increased steam flow to the turbine and will provide sufficient control margin for the turbine control valves, such that continuous stable operation will be maintained. The new dome pressure has been considered in the reactor equipment setpoints that are pressure dependent, such as: reactor scram setpoint on high pressure, ATWS reactor recirculation pump trip, and SRV setpoints. The new pressure also was considered in determining the revised initial conditions for accident and transient analyses.
2-1
NEDO-30603-1 Table 2-1 COMPARISON OF DAEC REACTOR HEAT BALANCE PARAMETERS FOR RATED POWER AND UPRATED POWER CASES Parameter Rated Power Uprated Power Units
'. Thermal Power 1593 1658 MWt 6 6 Core Flow 49.0 x 10 49.0 x 10 lb/hr Dome Pressure 1020 1040 psia Steam Flow 6,842,969 7,172,184 lb/hr Core Inlet Enthalpy 526.3 528.6 Btu /lb Feedwater Enthalpy 397.6 402.3 Btu /lb
+
l 1
2-2 l
e NEDO-30603-1 LEGENO e = F LOW, LS/HR F = TEMPER ATUR E, *F H,h = ENTH ALPY, Stu/HR M = % MOISTURE P = PRES $URE. PSI A ASSUMED SYSTEM LOSSES "P THERMAL 1.1 MW N
\
_, f,, ,
MAIN STEAM FLOW
,, / / 7.172.184 a f --e - e' -
1190.8 0.J M 1002 P MAIN FEED FLOW 20M,000 e
" '*'*# 7,151,184 e 534 F S29.4 h 424 F,402.4 h 'b 424 F 402J h 9
2 REclRCULATION LOM ' TOTAL COR E 18 INTERNAL JET PUMPS pg 49.0 a 100 m oc s2s,e h I
ah = 0.81 JIllll 438F i 418.4 h l CLEANUP OEMIN ER A LIZER SYSTEM i ! ji 70.000 m l
li 21,000
- 534 P l ROD ORIVE $28.6 h FEEDFLOW 80 F l 48 h l
' FROM CONOENSATE
( l STOR AGE TANK 1
Figure 2-1. DAEC Reactor Heat Balance At Uprated Power i
I L
2-3
NEDO-30603-1 LEGENO e a F LOW. LS/HR F
- TE MPE R ATUR E. *F H.h
- ENTH ALPY. Stu/HR M * % MOISTURE P
- PR ESSURE. PSI A ASSUMED SYSTEM LOSSES N
- ISOLAflON VALVES 1055 P THERMAL 11 MW N
N
~
f jh h~ p MAIN STE AM F LOW 7.343.428 m
(__ t :_ ::, _
m ,,,,;
0.3 M 1017.0 P MAIN FEEO FLOW
- 7.3 2.4n . 7.322.428 ,
427 F 404,9 h II 427 F 404.8 h 536F 531.1 h U
2 RECIRCULATION LOOPS s/
'I N/
TOTAL CORE II 16 INTERN AL JET PUMPS ptow 49 0 a 100 e T
O c S30.3 h an
- 0 81 b llIl 439 F 418.1 h
%= #
CLEANUP OE MINE R All2E R SYSTE M I I jl 70.000 e r
ROO ORivt 21.000 e 63$ F PEED F LOW 80 F 530.3 h 44 h r
FROM CONDE NSATE STOR AGE TANE Figure 2-2. DAEC Reactor Heat Balance At 302% of Upsetad Power 2-4 i
- 3. POWER /FIDW MAP The power / flow map for the DAEC at uprated power conditions is given in Figure 3-1. The map is a plot of core thermal power (in percent of uprated) versus core flow rate (in percent of rated) for various operating conditions.
The power / flow map contains information on expected system performance and limits of recirculation system operation for cavitation-free operation of the recirculation pumps and the jet pumps. The shaded area indicates the region expanded by the ELLLA (Subsection 12.1).
a 3-1
I NEDO-30603-1 120 o'
UPRATED POWER = 1868 MWt
'Y f RATED FLOW a 4S MHe/iw
/
~
' APRM ROD BLOCK = 0.58W + 50
/
f
/
/
/
(100,87)
/
/
80 -
/
E
/
/
__q g 100% LOAD LINE 5
8 a.
O U
t l SHADED AREA INDICATES REGION EXPANDED 8Y i 40 -
l EXTENDED LOAD LINE LIMIT ANALYSIS I
NATURAL f MINIMUM PUMP SPEED -
CIRCULATION f 20 -
j
/
/ CAVITATION
/ PROTECTION
/
/
0 " ' ' ' I O 20 40 60 80 100 CORE FLOW (% OP RATED)
Figure 3-1. DAEC Power / Flow Map 3-2
- 4. INSTRUMENT SETPOINTS Instrument setpoints are those values of sensed variables which result in initiation _of protective actions and are specified in the plant Technical Specifications. Some instrument setpoints are dependent on the maximum ther-
- mal power and must be modified to maintain equivalent safety margins for operation at the uprated power level. Only those instrument setpoints that are affected by the reactor thermal hydraulic parameters (Section 2, " Heat Balance")
.will be modified; all other setpoints will remain the same.
The determination of reactor instrument setpoints is based on plant operating experience and conservative analyses. The settings are selected high' enough to preclude inadvertent initiation of the protective action but low enough to assure that a significant margin is maintained between the safety systems settings and the actual safety limits.
As a result of increasing the reactor dome pressure (Section 2, " Heat Balance")_and enhancing the SRV simmer margin for the DAEC,* three instrument
.setpoint modifications are required before increasing reactor output to 1658 MWt. .These modifications are as follows:
- c. Safety Relief Valve Opening Pressure l The' bases for the changes are discussed in Subsections 4.1 through 4.3.
- f. ' 4.1 REACIOR VESSEL HIGH-PRESSURE SCRAM SETPOINI l-I' During a pressure increase transient not terminated by direct scram or high-t .
p flux .,cr a.> the nigh-pressure se'": will cerminate. the tramient and prevent .
( ' damage to the reactor vessel and the primary coolant pressure boundary (Sec-l
- tion 6, " Reactor Vessel Overpressure Protection").
l *Sismer margin is the difference between normal plant operating pressure and j the.SRV setpoint. Increase in valve simmer margin reduces the probability of l' steam leakage past the SRV pilot disc and thus precludes degradation of SRV performance. Analysis of the impact of_ increased simmer margins on plant ,
accident and transient response is contained in Reference'5.
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NEDO-30603-1 The reactor vessel high-pressure scram setting is maintained slightly above the reactor vessel maximum normal operating pressure. The setting per-mits normal operation without spurious scram, yet in the unlikely event of failure of other scrams provices a wide margin to the maximum allowable reactor vessel pressure.
For the DAEC operation at 1658 MWe in Cycle 8, the nominal reactor aper-ating pressure will be 20 psi greater than the nominal reactor operating pres-sure in Cycle 7. As a result of this pressure increase, the Technical Speci-fication scram setpoint on reactor high-pressure will be set at 1055 psig.
In order to account for an allowable instrument setpoinc drift of psi, all analyses were performed assuming a scram setpoint on high-pressure at psig.
4.2 RECIRCUI.ATION PUMP HIGH-PRESSURE TRIP SETPOINT The DAEC is equipped with two RPT Systems, one of which would actuate
.during plant transients associated with increases in reactor vessel dome pres-sure, the other would actuate on turbine stop valve closure or turbine control valve fast closure. The RPT on high-pressure is designed to provide a prompt negative reactivity effect during the initial part of an ATVS event.
As a result of the increase in reactor dome pressure of 20 psi at uprated power, the ATWS RPT setpoint on high-pressure will be increased from 1120 psig to 1140 psig.
4.3 SAFETY RELIEF VA1.VE OPENING PRESSURE SETPOINTS The third modification to the instrument setpoints of the DAEC at uprated power will involve raising the SRV opening pressurc setpoints by 30 psi. The incremental LP of 10 psi between the various valve groups will be maintained.
The raising of-the SRV setpoints will result in improving the SRV simmer margin (difference between reactor dome pressure and the SRV opening pressure),
while maintaining a significant margin to ASME pressure limits during plant transients.
- General Electric Company Proprietary Information has been del" ed.
4-2
NEDO-30603-1 The resulting nominal opening pressure setpoints for the SRVs will be as follows:
Valve No. valve Nominal Setpoint Value (psig) 1 1110 2 1120 3 1130 4 1U0 5 1140 6 1140 The above modifications in the pressure setpoints have been verified by analysis to meet the necessary design limits of the DAEC (Reference 5). This was done through the performance of the reactor vessel overpressure protection analysis of Section 6 and of the fuel thermal analysis of Section 8.
4.4 OTHER INSTRL' MENT SETPOINTS The following instrument setpoints that relate to plant conditions at full.
power need not be redefined, since they are referenced to rated conditions.
- b. APRM Control Rod Block Setpoint
- c. Main Steam Line High Radiation Scram Setpoint
- d. High Steam Flow tiSIV Closure Setpoint E. Turbine Stop Valve Closure and Turbine Control Valve Fast Closure Scram Bypass Setpoint .
For. the DAEC, Cycle 8, following approval of this submittal, rated power will be redefined as 1658 HWt. This change will be incorporated in the modi-fications of the DAEC Technical Specifications (Reference 6).
4-3/4-4
- 5. LOSS-OF-COOLANT ACCIDENT ANALYSES
5.1 INTRODUCTION
Ths purpo==. of thi. section is to provide *he resulta of the loss-of-coolant accident (LOCA) analysis for the DAEC. This analysis of the LOCA is provided to demonstrate conformance with the Emergency Core Cooling System (ECCS) acceptance criteria of 10CFR50.46 at uprated power conditions.
The performance of the ECCS is demonstrated through application of the 10CFR50 Appendix K evaluation models (Reference 7) and by showing conformance to the acceptance criteria for 10CFR50.46. The ECCS performance is evaluated for the entire spectrum of break sizes for postulated LOCAs.
5.2 ACCEPTANCE CRITERIA FOR ECCS PERFORMANCE The. applicable acceptance criteria, extracted from 10CFR50.46 are listed and, for each criterion, applicable figures, tables and references where con-formance is demonstrated are indicated.
The evaluated LOCAs and the applicable figures are summarized in Table 5-1. The response of water level inside the shroud and reactor vessel pres-sure during the accident analyzed is shown in Figures 5-la through 5-le. Peak cladding temperature is dealt with in Figures 5-2a through 5-2e. The heat transfer coefficient is presented in Figures 5-3a, 5-3b, and 5-2c through 5-2e.
Core average inlet flow is given in Figures 5-4a and 5-4b. Peak cladding tem-perature versus break area is represented in Figure 5-5.
Criterion 1: Peak Cladding Temperature "The Calculated maximum fuel element cladding temperature shall not exceed
-2200'F." Conformance to Criterion 1 is shown in Figures 5-2a through 5-2e and specifically in Tables 5-2a through 5-2d (MAPLHCR, maximum local oxidation, and peak cladding temperature versus exposure).
5-1 L
NEDO-30603-1 Criterion 2: Maximum Claddina Oxidation "The calculated total local oxidation of the cladding shall nowhere axceed 0.17 timas the total cladding thickness bafora oxidation." Conformance to Criterion 2 is shown in Tablen 5-2a through 5-2d (local oxidation versus t
-exposure) and Table 5-3 (break spectrum summary).
' Criterion 3: Maximum Hydrogen Generation i
"The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react." Conformance to Criterion 3 is shown in Table 5-3 (Core-Wide Metal-Water Reaction) .
Criterion 4: Coolable Geometry
" Calculated changes in core geometry shall be such that the core remains
' amenable to cooling." As described in Reference 7,Section III.A. conform-ance to Criterion 4 is demonstrated by conformance to Criteria 1 and 2.
Criterion 5: Lona-Term Coolina "After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by i the long-lived radioactivity remaining in the core." Conformance to Criterion.
5-is demonstrated generically for General Electric BWRs in Reference 7,Section III.A. Briefly summarized, the core remains covered to at least.the jet pump suction elevation and the uncovered region is cooled by spray cooling- !
and/or.by steam generated in the covered part of the core.
5.3 INPUT TO ANALYSIS l A 11s't of the significant plant, fuel and ECCS system performance input parameters to the LOCA analysis is presented in Table 5-4.
5-2 h-- ~
e NEDO-30603-1 5.4 LOCA ANALYSIS MODELS All LOCA analyses were performed using 10CFR50, Appendix K evaluation models. The analytical models used are documented in Reference 7. The same models were used for this evaluation that were used for previous Duane Arnold LOCA a.alyses.
5.5 BREAK SPECTRUM RESULTS For convenience in describing the LOCA phenomenon, the break spectrum has been separated into three regions: small breaks, intermediate breaks and large breaks. The selection of the break sizes to be included in each region is dependent on the most limiting single failure and the ECCS evaluation method used. The potentially limiting single active failures evaluated in establishing the various break regions are given in Table 5-5.
The small break region is defined as that portion of the break spectrum where the high pressure coolant injection (HPCI) is the most limiting single failure. In this region, the small break methods (SBM) are used. The limit-2 ing break size in this region is 0.07-ft . The results of the 0.07-ft break analysis are shown in Figures 5-le and 5-2e.
The intermediate break region is defined as that portion of the break spectrum up to the transition break where the low pressure coolant injection (LPCI) injection valve is the most limiting single failure. The transition break is defined as the 1.0-ft 2 break size. This break size has been chosen to be consistent with previous analyses. Both large and small break calcula-tional techniques censervatively model the respective break range above and below the transition break. The transition break has been analyzed with both the large and small break methods with the same single failure to allow a com-parison between the methods. The analysis of the transition break is shown in Figures 5-Ib and 5-2b for the large break method and Figures 5-ic and 5-2c for the small break methods. In the intermediate break region, small break methods are used. The results of the 0.8-ft analysis are shown in Figures 5-id and 5-2d as being representative of an intermediate break analysis.
5-3
l NED0-30603-1 The large break region is defined as that portion of the break spectrum between the transition break and the design basis accident (DBA). The DBA is defined as the complete severance of the largest pipe in that portion of the system which yields the highest peak cladding temperature when the most limit-ing single failure is assumed. The most limiting single failure in this reg *on is the failure n* tba LPCI injetsion valve. In Gie largt oreak region, !
large break methods (LBM) are used. The DBA results are presented in Figures I 5-la and 5-2a.
5.6 LOCA ANALYSIS CONCLUSIONS Having shown compliance with the applicable acceptance criteria of Subsection 5.2, it is concluded that the ECCS will perform its function in an acceptable manner and meet all of the 10CFR50.46 acceptance criteria, given operation at or below the maximum average planar linear heat generation rates (MAPLHCRs) in Tables 5-2a through 5-2d. Low-flow effects on LOCA analyses have been presented to the United States Nuclear Regulatory Commission (Refer-ences 8 and 9) and these effects apply to the DAEC at uprated power.
t 5-4 i-
Table 5-1 LOCA ANALYSIS FICURE
SUMMARY
Large Break Method t; mall Break Method Limiting Break
- DBA - ' Transition Break Transition Break Limiting Break (LPCI Inj. (LPCI Inj. (LPCI Inj. (LPCI Inj. Limiting Break Valve Failure) Valve Failure Valve Failure) Valve Failure) (HPCI Failure)
(2.51 fc2) (1.0 ft2) (1.0 ft2) (0.8 ft2) (0.07 ft2)
Water Level: Inside 5-la 5-lb 5-Ic 5-Id 5-le Shroud and Reactor 2:
Vessel. Pressure Q
u Peak Cladding 5-2a 5-2b 5-2c 5-2d 5-2e O O Temperature 5-3a 5-3b
{
Heat Transfer 5-2c 5-2d 5-2e Y Coefficient Core Average 5-4a 5-4b Inlet Flow Peak Cladding Tem- 5-5 5-5 5-5 5-5 5-5 perature vs. Break
-Area
NEDO-30603-1 Table 5-2a MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DRB30lt Average Planar exposure MAPLHGR PCT Caidation (mwd /t) (kW/ft) (*F) Fraction 200 11.5 1947 0.011 1000 11.5 1936 0.011 5000 11.9 1927 0.011 10000 12.3 1935 0.011 15000 12.4 1959 0.011 20000 12.2 1941 0.011 25000 11.3 1854 0.008 35000 9.9 1679 0.004 45000 8.7 1559 0.002 Table 5-2b MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DPB289 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 11.2 1915 0.010 1000 11.2 1906 0.010 5000 11.8 1923 0.010 10000 12.0 1914 0.009 15000 12.1 1926 0.010 20000 11.9 1921 0.010 25000 11.4 1863 0.008 30000 10.8 1786 0.006 35000 10.3 1712 0.005 40000 9.6 1646 0.004 45000 8.9 1580 0.003 5-6
7-NEDO-30603-1 Table 5-2c MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DRB284H Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) T* action 200 11.2 1912 0.010 1000 11.2 1900 0.010 5000 11.7 1909 0.010 10000 12.0 1921 0.010 15000 12.0 1926 0.010 20000 11.8 1918 0.010 25000 11.1 1837 0.007 30000 10.4 1744 0.005 35000 9.8 1674 0.004 40000 9.1 1608 0.003 45000 8.5 1541 0.002 Table 5-2d MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: DAEC BUNDLE TYPE: P8DRB299 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 10.9 1874 0.009 1000 11.0 1869 0.009 5000 11.5 1874 0.009 10000 12.2 1929 0.010 15000 12.3 1950 0.011 20000 12.1 1948 0.011 25000 11.5 18')0 0.009 30000 11.0 1807 0.007 35000 10.3 1725 0.005 40000 9.7 1661 0.004 45000 9.0 1601 0.003 5-7 m
NEDO-30603-1 Table 5-3
SUMMARY
OF BREAK SPECTRUM RESULTS
-* Break Size Peak Local Core-Wide
-e Location Oxidation Metal-Water 4 e Single Failure PCT (*F) (%) Reaction (%)
e 2.51 ft2 (Da.*.)
e Recirculation Suction 1959 1.1 0.08 e LPCI Injection Valve e 1.0 ft2 (LBM) e Recirculation Suction 1765 Note 1 Note 2 e LPCI Injection Valve e- 1.0 ft (SBM) e Recirculation Suction 1358 Note 1 Note 2 7
e LPCI Injection Valve l
[
t e 0.8 ft e Recirculation Suction 1142 Note 1 Note 2 e LPCI Injection Valve
, e' 0.07 ft
- e Recirculation Suction 1566 Note 1 Note 2 1J e HPCI .
f-4 .
9
. NOTES:
- l. :Less than DBA (1.1%)
- 2. - Less than DBA (0.08%).. .
~!
5-8 i
\ .
l
F NEDO-30603-1 Table 5-4 SIGNIFICANT INPUT VARIABLES USED IN THE LOSS-OF-COOLANT ACCIDENT ANALYSIS A. Plant Parameters Variable Units Value __
Core Thermal Power MWt 1691.0 6
Vessel Steam Output- lbm/hr 7.344x10 Corresponding Percentage of % 102 Uprated Steam Flow Vessel Steam Dome Pressure psia 1055 Lower Tie Place --
Fully Drilled B. Emergency Core Cooling Systems Parameters B.1 Low Pressure Coolant Injection System Variable Units Value Vessel Pressure at Which Flow psid (vessel to 197 May Commence drywell)
Minimum Rsted Flow gpa 14,400 at Vessel Pressure psid (vessel to 20 drywell)
Initiating Signals:
_Lov Water Level ft above top of 1.0 or active fuel (TAF)
[ High Drywell Pressure' psig >2.0 Maximum Time Delay sec 40.0 1 from Initiating Signal to i -
Pumps at Rated Speed
. Injection Valve Fully Open seconds after_DBA <40.0 I
initiation signal
~5-9 c.
NEDO-30603-1 Table 5-4 (Continued)
SIGNIFICANT INPUT VARIABLES USED IN THE LOSS-OF-COOLANT ACCIDENT ANALYSIS ,
B.2 Low ?ressure Core Spray System Variable _.
Units Value-Vessel Pressure at Which Flow paid (vaasel to 264 May Conunc ..e drywell)
Minimum Rated Flow gpm 6040 at Vessel Pressure paid (vessel to 113 drywell)
Initiating Signals:
Low Water Level ft above TAF 1.0 or High Drywell Pressure psig >2.0 Maximum Allowed (Runout) Flow gpm 7800 r Maximum Time Delay from sec 27.0 Initiating Signal to Pumps at Rated Speed Injection Valve . Fully Open see after DBA $27.0 initiation signal B.3 -High Pressure Coolant Injection System Variable Units Value,
-Vessel Pressure at Which Flow psid (vessel to 1135 May Commence- pump suction) ,
Minimum Rated Flow Available gpm 3000' ,l at Vessel Pressure 'psid (vessel to pump 150 suction) a r-Initiating Signals:
. Low Water Level fC above TAF >10.5 or 7 High.Drywell Pressure psig >2.0 Maximum Delay Time sec 30.0
.. Initiating Signal to Rated.
Flow Available and Injection
" Valve, Wide Open 5-10.
e NEDO-30603-1 Table 5-4 (Continued)
SIGNIFICANT INPUT VARIABLES USED IN THE LOSS-OF-COOLANT ACCIDENT ANALYSIS B.4 Automatic Depressurization System (ADS)
Variable Units Value Total Number of Relief Valves 4 with t.03 Fuaccion 6
Total Minimum Flow Capacity lb/hr 3.58x10 at Vessel Pressure psig 1125 Number of Valves Remaining 3 with One Valve Failed 0
Total Minimum Flow Capacity lb/hr 2.69x10 at Vessel Pressure psig 1125 Initiating Signals:
Low Water Level ft above TAF 3,1. 0 or High Drywell Pressure psig 1,2. 0 Delay Time from All Initiating sec <120 Signals Completed to the Time Valves are Open C. Fuel Parameters Variable Units Value Bundle Type (limiting case) ---
P8DRB30lL
, Fuel Bundle Geometry ---
P8x8R Number of Fueled Rods -62 Peak. Technical Specification kW/ft 13.4 Linear Heat Generation Rate Initial Minimum Critical ---
1.20
- Power Ratio Design' Axial Peaking Factor --- 1.40 s
5-11
NEDO-30603-1 Table 5-5 3 INGLE-FAILURES EVALUATED The following table shows the single, active failurec evaluated in the ECCS perfucmance eialuation.
Suction Break Assumed Failurd Systems Remaining LPCI Injection Valve ADS, 2 core spray (CS),
HPCI Diesel Generator (D/G) ADS, 1 CS, HPCI, 2 LPCI HPCI ADS, 2 CS, 4 LPCI t< ,
All other postulated single failures are bounded by one of the above; i.e., remaining ECCS capacity will be as great or greater than .one or more of the above.
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- 6. REACTOR VESSEL OVERPRESSURE PROTECTION
. The -Duane Arnold Energy Center pressure relief system was designed to pre-vent excessive overpressurization of the primary system process barrier and the
- t. pressure vessel co preclude an uncontrolled release of fission products. The l' DAEC oressara relief cystem includes ras spring safety val.ves (SS":) and olx dual. function SRVs. T:tese valves provide the capacity to limit nuclear system I
I overpressurization.
The ASME Boiler and Pressure Vessel Code requires that each vessel
. designed to meet Section III be protected from the consequences of pressure in
' excess of the vessel design' pressure (a more detailed discussion of the vessel pressure ASME Code compliance is contained in Reference 10, Subsection 5.2.3):
- a. A peak pressure of 110% of the vessel design pressure is allowed under upset conditions (1375 psig for a vessel with a design pressure of 1250 psig).
- b. The lowest qualified safety valve setpoint must be at or below vessel design pressure.
-c. The highest safety valve set point.must not be greater than 105% of vessel design pressure (1313 psig for a 1250-psig vessel).
~
The two SSVs are set to actuate at 1240 psig. The proposed setpoints for: the six dual function SRVs are 1110, 1120, 1130 (two valves) and 1140 (two valves) psig. These setpoints satisfy Requirements "b" and "c".
Requirement "a" is conservatively evaluated by considering the most severe isolation event with indirect scram. .The SRVs are assumed to be active. The event selected is the closure of all main steam line isolation valves with flux scram. ~The main steam isolation valve (MSIV) closure with failure of
~ direct' scram-is technically an emergency event and application of the
_ upset criterion demonstrates the large excess capacity of the DAEC pressure l
6-1 L__
NEDO-30603-1 relief system. The results of this analysis are given in Table 6-1 and shown
.in. Figure 6-1. An abrupt pressure and power rise occurs as soon as the reactor is isolated. Neutron flux reaches scram level in about 1.65 seconds,'initiat-ing reactor shutdown. The SRVs open to limit the pressure rise at the bottom of the vessel to 1275 psig. There is a 100-psi margin to the vessel code limit of 1273 psig. Requirement "a" *.: aasily satisfied and ad. equate over-pressure protection is provided by the pressure relief system.
6-2
NED0-30603-1 Table 6-1 TRAN3IENT OUTPUT DATA
SUMMARY
Peak Peak Peak Peak core Neutron heface Steam Line Vessel Powera Flow Flux Heat Flux Pressure Pressure Event (%) (%) (% of Ref) (% of Ref) (psig) (psig)
MSIV b
Closure 100* 100 620 122 1244 1275 (Flux Scram)
{
Uprated 100 psi margin to ASME Code limit of 1375 psig.
6-3
NED0-30603-1
{ 1 NEUTRON' FLUX 1 VESSEL PRESS RISE IPSI) 2 AVE SURFACE '4 r EAT FLUX 3 SAFt:TY/ RELIEF VALVE FLOW J CORE INLET FLOW 4 SYPASS VALVE FLOW 150.0 30s s -'
I
)1 r
100.0 ,
- , "_ W 200.0 N
h w
D a
r 50.0 100.0 0 0 0 0 ;
0.0 _, o,o , , , , , _ , ,
0.0 5.0 0.0 5.0 TIME (SECONOS)
TIME (SECONOS) 1 LEVEL (INCH 4EF.SEP. SKIRT) I 1 VOlO REACTIVITY 2 VESSEL STEAMFLOW 2 OOPPLER REACTIVITY M0 3 TURBINE STEAMFLOW 3 SCRAM REACTIVITY e 4 FEEDWATER FLOW 1.0 4 TOTAL REACTIVITY h
=
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Figure 6-1. MSIV Closure with Flux Scram, Uprated Power 6-4
l
- 7. TRANSIENTS AND ACCIDENTS This section documents the calculated consequences of the most limiting abnormal operational transients performed for the DAEC at uprated power. The licensing basis for the transient and accident analyses for Reload 7, Cycle 8, is contained in Reference 4, and the transient and accident analyses methods j are described in Reference 10.
7.1 TRANSIENTS The scope of the transient analyses performed is identical to the reload licensing submittal (Reference 4). Requirements for vessel overpressure protection were discussed in Section 6; the evaluation of limiting abnormal operational transients from the standpoint of thermal margin requirements are discussed below. These safety requirements are based on the protection of the fuel. The most limiting abnormal operational transients for DAEC are Load Rejection without Bypass (LRw/oBP), Turbine Trip without Bypass (TTw/oBP),
Loss of Feedwater Heating (LFWH), Feedwater Controller Failure (FWCF) and Inadvertent Startup of HPCI Pump (EPCI) . These transients were performed at the uprated conditions (effects of uncertainty in the measured power are discussed in Section 8). Key results are shown in Table 7-1.
7.2 ROD WITHDRAWAL ERROR The rod withdrawal error (RWE) for the DAEC, Cycle 8, at uprated condi-tions, is based on the generic bounding analysis. The generic bounding analy-sis database includes a number of plant.1 of different power densities and is valid for the DAEC at uprated power. The results.obtained are shown in Table 2. The rod block monitor (RBM) setpoint of 105% is selected to allow for' failed instruments in the worst allowable situation. It is demonstrated that even if the operator ignores all alarms during the course of this transient, the RBM will stop rod withdrawal when the critical power ratio (CPR) reaches the 1.07 minimum critical power ratio (MCPR) safety limit. A more detailed discussion of analysis methods is included in Reference 10.
7-1
c-NEDO-30603-1
-7.3 CONTROL ROD'DRO? ACCIDENT A rapid. removal of a high worth control rod could result in a potentially significant excursion (i.e. , insertion of reactivity). .The accident which has
-been' chosen to encompass the consequences of a reactivity excursion is the CRDA. T.e sequcace of events and methodale;7 tr *ec;ribed "a Reference 10.
. Specific analysis results for the DAEC for the uprated conditions are chosn in Reference 4. The resultant peak enthalpy values are within bounding limits.
I l
I l
l 7-2
NEDO-30603-1 Table 7-1 CORE-WIDE TRANSIENT RESULTS ACPR Flux Q/A .P8x8R
- Transient (% NBR) (% NBR) (Unadj usted)
I ' Exposure: BOC 8 to E00 8 447.5 113.4 0.16 L Load Rejection w/o Bypass f
Exposure: BOC 8 to EOC 8 387.4 111.7 0.15 Turbine Trip w/o Bypass Exposure: BOC 8 to EOC 8 121.1 116.6 0.14 Loss of Feedwater Heater Exposure: BOC 8 to EOC 8 121.1 116.4 0.14 Inadvertent Startup of HPCI Pump Exposure: BOC 8 to EOC 8 278.2 110.1 0.11 Feedwater Controller Failure l -
7-3
NEDO-30603-1 Table 7-2 LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(Generic Bounding Analysis Results)
ACPR Rod Block Reading (All Fuel Types) 104 0.13 105- 0.16 106 0.19 107 0.22 108 0.28 109 0.32 110 0.36-Setpoint Selected: 105
- . ~j ,'
7 e.
[-
y-NEDO-30 503-1
- 8. .MCPR LIMITS
' Operating limits are specified to maintain adequate margin to the onset l- of boiling transition during abnormal operational transients. The figure of
. merit.ittilized for plant operation is the CPR, which is defd.ned as the ratio
- of the critical power (bundle power at which some point within the assembly l
l experiences onset rf biling transition) to the operating bundle power. The critical power is detstIdined at the same mass flux, inlet temperature, and pressure which exists at the specified reactor condition.
8.1. APPLICABILITY OF CURRENT SIMCPR AND K CURVES
- F The safety limit minimum critical power ratio (SIRCPR) and the Kgcurves
-(MCPR'aultiplier) are limits that have been incorporated in the design to achieve the objective of maintaining nucleate boiling, and thus avoid boiling transition. .The SLMCPR for the DAEC is 1.07. The Kg curves, as a function
~ of core: flow, are given.in Reference 10.
The design calculation of the safety limit MCPR is based on a Monte Carlo
' analysis of core performance in a limiting configuration and takes into account both performance monitoring uncertainties and calculational unecrtain-
. ties. For uprated power applications, the .only change from the standard analy-sis is in the nominal plant conditions. The radial power distribution will
-remain within the bounds given in Reference 10. It is concluded.that the 1.07
',' reload SLMCPR~ applies to uprated power'as well as to the current Technical iSpecification power rating.
. Calculations of Kg curves are normally done for steam flow condition
- -(uprated power) even though a plant is only licensed to power. This
- lends inherent conservatism to theg K basis. The existing Kg curves are,
~cherefore,:directly applicable to uprated power operation.
s 8-1
- __- . _ , - . . ., _ . ,_ . .. . , - . .__j
t =-
l l
'8.2 EFFECT OF POWER LEVEL UNCERTAINIY ON MCPR OPERATING LIMIT Regulatory' Guide 1.49 (Reference 11) stares: "Some of the analyses in l support of the proposed licensed power level are made for a slightly higher assumed. pow---lavel to allow for possible errors in determining the power level. The Regu} story Staff has determined that a margin of two percent of the licensed power is adequate for this purpose." General Electric has chosen to incorporate allowances for power level uncertaintv into the ODYN transient analysis'and into_the GETAB thermal margin analysis. The acceptance of this method of ' accommodating power level uncertainty is found in the ODYN Safety Evaluation Report (Reference 12) and the Review and Evaluation of GETAB
-(Reference 13).
~
The ODYN Safety Evaluation Report (Reference 12) describes the methodology whereby a power level uncertainty equivalent to 2% has been included in the ODYN " adders":
" Uncertainties uin 'the ODYN Code need to be considered since these will affect the probability of exceeding the thermal-hydraulic design basis. One method of accounting -for the effects of ODYN code uncertainties...is to assure
. that the ODYN licensing calculation gives a sufficiently conservative value of ACPR to assure that the thermal-hydraulic: design basis is not exceeded. GE has chosen to use. . . (this) . . . method in demonstrating the acceptability of the ODYN licensing basis.
i
~" General Electric has provided statistical analyses of the loss of load L
~
CIurbine Trip) and feedwater controller failure transients. These analyses use Monte; Carlo calculations to predict ACPR1with a second order response surf ace which simulates ODYN calculations. The input parameters in the.
response surface are: initial power; ...
"The distribution functions of each of the input variables (initial power,
. . . ) were. reviewed.. .The uncertainty on initial power level used by CE was +2%.
We' requested additional'information to substantiate this value and were given extensive _information on the various elements in the plant energy balance and the uncertainties _ associated with each of these elements. The elements of the 8-2
+
w m. w . . _ - , - - , - , - , - - - -
f- energy balance were checked against the ASKE standard for deteenining energy output from a nuclear plant, "ASME Performance Test Codes, Test Code for Nuclear Steam Supply Systems (PTC 32.1-1969)." In addition, the uncertainty values for each element were ceviewed and found to be reasonable. We have concluded that the 2% uncertainty (at one standard deviation) is an acceptable value for power measurement uncertainty."
The effect of uncertainties in measured power level are also included in the GETAB/GEXL conclusion. The Review and Evaluation of GETAB (Reference 13) notes: "GE... proposes to combine the effect of the uncertainties in the GEXL correlation with the uncertainties in the reactor operating variables in determining the thermal limits...the direct incorporation of uncertainties in the proposed CPR thermal limit assures that uncertainties are considered
.during design and operation of the reactor."
Reference 13 further describes the statistical procedure: "The statisti-cal procedure uses a computer program which calculates the CPR of the bundles in the core assuming a given power distribution and values of the operating variables. Using the calculated values of CPR, the probability of boiling transition occurring is summed for all rods in the core. Successive trials using random variations in the opersting variables are performed until the mean and standard deviation of the probability of boiling transition occurring in the core is found.
"...The random variations in operating variables are based on estimates of the uncertainties in each . variable. (23,24)
- A review and evaluation of these variables has shown that the variables which contribute significantly to the overall uncertainty have been considered."
.An examination of the references shows that power level uncertainty is satisfied by accounting for the uncertainties in feedwater flow,,feedwater temperature and reactor pressure. The uncertainty of these variables- (one sigma) taken together is very close to the recommended 2%.
I'
- References 14 and 15 in this report.
8-3 L
- ; Reference 13 concludes: "We conclude that the proposed design basis
-(i.e., more than 99.9% of the fuel rods in the core would be expected to avoid
'a boiling transition caused by single operator errors or equipment malfunc-tions) is acceptable when applied to core-wide transients such as a turbine-trip or pump-coastdown transient. We also conclude that the method used ta calculate the MCPR thermal limit is an acceptable method by which power dis-
.tribution and uncertainties in the CEXL correlation and the reactor operating parameters can be included in the determination of whether the design basis is met."
8.3 MCPR OPERATING LIMIT The DAEC uprated MCPR operating limit has been established to ensure that the fuel cladding integrity safety limit is not exceeded for any abnormal operational transient. This operating requirement is obtained by addition of the absolute maximum ACPR value (including any imposed adjustments factors) for the most limiting transient postulated to occur at the plant from uprated conditions, J to the fuel cladding integrity safety limit.
The Option A MCPR operating limit for the DAEC, Cycle 8, is 1.28 (P8x8R fuel). and the Option B MCPR operating limit for cycle 8 is 1.26 (Reference 4).
4 i
O S
8-4
- 9. STABILITY Thermal hydraulic stability analyses for the. DAEC, Cycle 8, are presented in Reference 4 for the uprated conditions. The results are summarized in l
S"bcection 4.1 and 9.2.
L 9.1 CHANNEL HYDRODYNAMIC CONFORMANCE TO THE ULTIMATE PERFORMANCE CRITERION The channel performance calculation for the DAEC, Cycle 8, yields decay ratios as presented below:
Channel Hydrodynamic Performance At the Extrapolated Rod Block Line - Natural Circulation Power / Flow Conditions Channel Type Decay Ratio (x2 /*0 P8x8R 0.31 At this most responsive condition, the most responsive channels are clearly within the bounds of the ultimate performance criteria of a 1.0 decay ratio at all attainable operating conditions.
9.2 REACTOR CONFORMANCE TO THE ULTIMATE PERFORMANCE CRITERION The decay ratios determined from the limiting reactor core stability.
conditions are presented-in Figure 9-1. The most responsive case is the
-extrapolated rod block line - natural circulation condition.
, Reactor Total Core Stability At the Extrapolated Rod Block Line - Natural Circulation Power / Flow Conditions Decay Ratio, x2 /*0 0.84 ,
These calculations show the reactor to be in compliance with the ultimate
. performance criterion, including the most responsive condition.
9-1
NED0-30603 1
)
l PERFORMANCE LIMIT 1.00 EXTRAPOLATE APRM ROD BLOCK 0.75 -
/
NATURAL CIRCULATION o
i M
7 e
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o 0.50 -
1 I
100% ROD LINE 0.25 -
0 I I ' !
0 20 40 60 80 100 120 PERCENT POWER Figure 9-1. Reactor Core Decay Ratio vs Power 9-2
- 10. REACTOR INTERNAL PRESSURE DIFFERENCES / STRUCTURAL EVALUATION 10.1 REACTOR' INTERNAL PRESSURE DIFFERENCES ANALYSIS A reacto_r internal pressure difference (RIPD) analysis was performed for the DAEC at upratea power to confirm cnat the safety design bases for reactor internals are met. The analysis examined the responses of the reactor vessel internals to loads imposed during normal and accident events. The reactor internals of interest (Figure 10-1) are as follows:
a.
b.
c.
d.
e.
f.
g.
The objective of the evaluation was to determine the maximum pressure differentials across the components during steady-state and upset conditions and during certain emergency and faulted conditions.
At steady state, the RIPD values were calculated.at the reactor conditions specified in Table 2-1.
. A com-parison of RIPD values at rated and uprated power conditions is given in Table 10-1.
l 10-1 L-
F NED0-30603-1 10 -2
ci 1... ,
_ NEDO-30603-1 10.2 STRUCTURAL EVALUATION OF DAEC AT UPRATED POWER A structural evaluation was performed to compare the calculated peak RIPD i .val ues at uprated power to the current design values at rated power. Results at nar=>1, upset, emergency and faulted conditions are given in Table 10-2.
For any of the = components, where the 11PD at uprated power exceeded the current design value, the component was reanalyzed. The results of this analysis indi-
. cate that the stresses due to the new RIPD are within the allowable limits.
The impact of power uprate on the reactor vessel and the vessel nozzle design stresses were also calculated. For the reactor vessel, the vessel pressure and temperature at uprated power conditions were found to be within the values used in Reference 16. Therefore, power uprate will not violate the design stress limits for the vessel. For the vessel nozzles, only the feedwater nozzle was judged to be significantly affected by the power uprate.
The results of the analysis show that the increase in feedwater flow due to power uprate will have a negligible impact on the fatigue duty of the feed-water nozzles.
The effect of increased neutron fluence on the reactor pressure vessel at uprated power was also analyzed. The DAEC operating practices, contained in References 3 and'6, were reviewed to assure the adequacy of the fracture tough-ness operating limits for the' reactor vessel. Based on this review, it is concluded that the practices used to adjust the operating limit curve of Figure 3.6-1 of the Technical Specifications (using dosimeter measurement extrapolaticns to determine the fluence versus integrated power) are correct and would account for the increase in the reactor power from 1593 MWt to 1658 MWt.
6.
e 10-3
NEDO-30603-1 Table 10-1 COMPARISON OF STEADY-STATE OPERATING RIPDs FOR RATED POWER AND UPRATED POWER CONDITIONS Pressure Differential psi Reactor Component Rated Power Uprated Power 1593 MWt 1658 MWt 4
D 10-4
h NEDO-30603-1 l
Table 10-2 l_
COMPARISON OF CURRENT' DESIGN RIPD VALUES
- j. AGAINST NEW RIPD VALUES AT UPRATED POWER CURRENT DESIGN RIPD NEW RIPD AT UPRATED POWER
- Events" Events" Component N U E F _ [I_ U E F 4
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-U.- UPSET E --EMERGENCY F - FAULTED 1
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- 11. CONTAINMENT EVALUATION 11.1 SHORT-TERM ACCIDENT RESPONSE ANALYSIS Analyses to determine the DAEC containment short-term accident response at'uprated power conditions were performed in accordance with the load proce-dures of Reference 17, used in the Mark I containment DBA evaluation. In this event, an instantaneous double-ended guillotine break (DEGB) of the recircula-tion pump suction line is postulated to occur. Following the rupture of the recirculation line, the flow out of both sides of the break will be limited to the maximum allowed by critical flow considerr.tions. Initial conditions l- of the reactor, at the time of event initiation, correspond to 102% of uprated power.- Other assumptions usad in the analysis are identical to those of Reference 18.
- The results for the drywell and wetwell pressures and temperatures, under
'the uprated power design basis conditions, are shown in Figures 11-1 and
-.11- 2 . A peak drywell pressure of 42.7 psig occurs at 4.2 seconds af ter event initiation and is well below the containment design pressure of 56 psig. A peak drywell temperature of 286.8'F is noted.in Figure 11-2, which is below
~ the drywell temperature limit of 340*F.
t A comparison of the DAEC containment responses at uprated reactor power wich' those of the rated power case (Reference 18) is shown in Table 11-1. As noted, the increase in drywell pressure is on the order of 1.0 psi. This
' increase is negligible and considerable margin to the design limit has been maintained.
11-1 i
NEDO-30603-1 11.2 LONG-TERM ACCIDENT RESPONSE ANALYSIS The long-term containment response to transients or accidents is char-acterized by suppression pool temperature, containment temperature and con-tainment pressure. Containment temperature is less than 200*F (Reference 3) compared with a design value of 281*F. Containment peak long-term pressure is less than 25 psig (Reference 3), well below the 56 psig design value. Because of the large margins between predicted and design temperatures and pressures no reanalysis was done at uprated power. Pool temperature response was evaluated for the DAEC at uprated power for the two events which result in the highest maximum local pool temperature based on previous analyses (Refer-ence 20). This evaluation for.uprated power at 1691 MWt (102% of uprated
_ power) also' allows for a 15% reduction in residual heat removal (RHR) heat exchanger service water flow (4080) gpm which results in a ' reduction in heat exchanger effectiveness relative to previous analyses.
~
Since the limiting local pool temperature criterion is , the long-term contain-ment response shows that the DAEC conforms to the limit at uprated power.
i 11-2
C NEDO-30603-1 TABLE 11-1 COMPARISON OF-DAEC CONTAINMENT RESPONSES TO DBA l
FOR RATED POWER AND UPRATED POWER CONDITIONS PARAMETER RATED POWER UPRATED POWER UNITS (1593 MWt) (1658 MWt) t .
- Drywell Peak Pressure '41.6 42.7 pais Wetwell Pressure at 23.6 23.7 psig 30 seconds Drywell Peak Temperature 285.7 236.8 'F Wetwell Temperature at 121.0 121.8 'F 30 seconds 11-3 -
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DRYWELL PRES 3URE - PSIG
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Figure.ll-1. Containment Pressure vs Time After Event DAEC Recirculation Suction Line DECB at Uprated Power
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- 12. IMPACT ON PLANT IMPROVEMENT PROGRAMS 12.1 EXTENDED LOAD LINE LIMIT ANALYSIS 12.1.1 Incroductior, In order to reach 100% uprated power level at high power / flow conditions without control rod withdrawals (which may be restricted by the Preconditioning Interim Operating Management Recommendations (PCIOMRs)], operation above the rated load line* is required during power ascension. This approach will minimize the capacity factor loss during startup.
L The following analyses were performed as part of the Extended Load Line Limit Analysis (ELLLA) for the DAEC, Cycle 8 (Reference 22), to verify the safe operation within the proposed extended power / flow region above the rated rod line:*
- a. Thermal-Hydraulic Stability Ar.alysis
- b. Loss-of-Coolant Accident (LOCA) Analysis
- c. Containment Response Analysis
- d. Pressurization Transients Analysis 12.1.2 Analyses and Results 12.1.2.1 Stability The results of stability analyses performed for the DAEC, Cycle 8 (Refer-ence 4) bound the least stable condition in the ELLLA region. These calcula-tions also showed compliance with the ultimate performance criteria including the least stable point.
- " Rated load line" or " rated rod line" refers to the power vs flow relation-ship attained with a constant rod pattern that intersects 1658 MWe at rated core flow. See the DAEC Power / Flow Map of Figure 3-1.
12-1
~12.1.2.2 Lass-of-Coolant Accident The LOCA analysis for the DAEC is applicable for plant operation in the power / flow domain bounded by the most limiting of the to11owing:
- a. 100% of uprated core power
- b. APRK rod block line The proposed operating domain is within the above limits.
12.1.2.3 Containment Response Impact of plant operation in the proposed domain for the DAEC has been evaluated for the containment LOCA response. The operating condition was 102%
of Uprated Power and 87% Flow. The results show no impact on the containment
.LOCA tesponse. The maximum drywell pressurization rate observed is less than the value used in plant-unique testing for defining LOCA-related pool swell loads.
12.1.2.4 Pressurization Transients The most limiting pressurization transients for the DAEC are Load Rejec-tion Without Bypass, Turbine Trip Without Bypass, and Feedwater Controller Failure. The transients were analyzed at 100%'Uprated Power and 87% Flow.
The results show that the (100P; 87F)* point is bounded by the licensing basis (100P; 100F) point for the Load Rejection Without Bypass and the Turbine Trip
- Without Bypass transients. The Feedwater Controller Failure at (100P; 87F) is more limiting than at the (100P; 100F) point;-however, the MCPR operating .
limit is not affected by the ELLLA.
To determine compliance to the American Society of Mechanical Engineers
'(ASME) pressure vessel code,-the MSIV closure with a flux scram event is used.
This transient event was also analyzed at the 100% Uprated Power and 87% Flow
- P = % Power; F = % Flow
'12-2
p-NEDO-30603-1 point at end-of-cycle nuclear exposure. The results were compared to those shown in Section 6; they showed the peak vessel pressure well below the 1375-psig limit and bounded by the licensing basis point.
12.1.3 Conclusion The results of the limiting transients for the 100% uprated power inter-cept point (100P; 87F) do not affect the MCPR operating limit established by the analysis at the (100P: 100F) point (Reference 4). The overpressure pro-tection analysis results are within the ASME pressure vessel code allowable for the (100P; 87F) point. The stability results are within the bounds of the ultimate performance criteria 1.0 decay ratio and the maximum average planar linear heat generation rate (MAPLHCR) results are unchanged by the extended operating region.
Therefore it is concluded that all safety bases normally applied to the DAEC are satisfied for operation within the ELLLA region in Cycle 8.
12.2 SINGLE-LOOP OPERATION The capability to operate using only one of the two recirculation driving loops is highly desirable in the event maintenance of a recirculation pump or other component renders one loop inoperative. In single-loop operation (SLO) the reactor .would be operating at a reduced power and flow consistent with the
- latest recommendations (Reference 23). To evaluate the impact of uprated power on SLO the following analyses have been reviewed for one-pump operation
- a. Loss-of-Coolant Accident
- b. Stability
- c. Transients and MCPR Limits 12-3
-12.2.1 Loss-of-Coolant Accident The analyses from References 24 and 25 dictate the use of a single-loop
-MAPLHGR multiplier which relecca the SLO MAPLHCR to the standard two-lcop operation (TLO) MAPLHCR. This multiolier is based on the similarity of the reactor core uncovered time (refloed time minus recovery time) for SLO and TLO during a LOCA.
For the DAEC at uprated power, LOCA single-loop analyses (Design Basis Accident "DBA", 70% DBA and 40% DBA) were performed at 102% of 1658 MWt. The results of the analyses were compared with the LOCA two-loop analysis of Section 5 (Table .12-1) . The analyses confirm that the reflooding times of the SID and TID during LOCAs are similar. It is concluded that the SID MAPLHGR multiplier of 0.87 used in Cycle 7 rated power is still applicable to Cycle 8 uprated pover.
12.2.2 Stability The least stable power / flow condition occurs at.the intersection of the extrapolated APEN rod block curve and the natural circulation flow curve.
This power / flow condition is not affected by one-pump operation.
Current recommendations (Reference 23) for operation of BWRs.have con-servatively restricted single-loop power operation. These reconumendations were evaluated for the DAEC at uprated power, and it was concluded that they.
are applicable to the uprated power condition. Under these conditions, the n TLO stability results reported in.Section 9 bound SLO.
l 12.2.3 Transients and MCPR Limits Operation with one recirculation loop results in a maximum power output f which is 20% to 30% below that which is attainable for two-pump operation.
Therefore, as stated in Reference 25, the consequences of abnormal operational ]
transients from one-lo'op operation will be considerably less severe than those analyzed from a two-loop operational mode. It is concluded that the results i 1
W s
12-4
NEDO-30603-1 of Section 7, " Transients and Accidents", bound both the thermal and overpres-sure consequences of one-loop operation at uprated power.
Although the transients results for two-loop operation 'ouuud SLO. the MCPR operating limit is increased to account for the increase in tha 6 el cladding integrity safety limit. This more restrictive MCPR safety limit accounts for the increased uncertainties in core flow measurements and in traversing in-core probe (TIP) readings (Reference 25).
12.3 SRV LOW-LOW SET SYSTD1 The purpose of the Low-Low Set (LLS) System is to mitigate the induced thrust loads on the SRV discharge line resulting from SRV subsequent actuations during abnormal transients or small break LOCAs. The load reduction is accomplished by lowering the opening and closing setpoint of two non-ADS SRVs after an initial SRV opening. As noted in Reference 26, the DAEC LLS
^
extends the time between SRV actuations beyond the minimum value of 3.7 sec-onds for all the limiting events considered in the design analysis.
For the DAEC at uprated power, the two limiting events considered in Reference 26 were re-analyzed:
a.
b.
The analysis at uprated power was performed at 102% of 1658 MWt. The results of the analysis (Table 12-2) indicate that the minimum time between valve actuations of events a and b exceeds the minimum value of 3.7 seconds by a factor -f . It is concluded that there is no adverse effect on the LLS System cue to a power uprate.
12-5
NEDO-30603-1 Table 12-1 COMPARISON OF LOCA REFLOODING CHARACTERISTICS FOR SINGLE-LOOP AND TWO-LOOP OPERATION '
Parameter Single-Loop Operation Two-Loop Operation Bruk Siz. : DM 70: D",A 40: DBA ,DBA 70~ DBA *0%
. DBA Reflooding Time, 161.1 154.8 174.7 160.3 154.3 173.8 .
see Total Uncovered Time, 138.1 127.1 127.1 137.2 126.1 127.2 sec Table 12-2
SUMMARY
OF LLS ANALYSIS RESULTS FOR LIMITING TRANSIENTS EVENTS (102% of Uprated Power)
Ti=c Between Valve Actuations Event (sec)
Minimum Design Value 3.7 12-6 l
NED0-30603-1
- 13.
SUMMARY
AND CONCLUSIONS This report presents the results of analyses of the DAEC NSSS and contain-ment systems necessary to demonstrate safe operation of the plant at the design basic power level Of 1058 MWt. These resulr=, when supported by appropriate analyses of plant auxiliary systems, thus provide justification for uprating the Technical Specification power level to 1658 MWt. A summary of the NSSS and containment systems analyses results is given in the following paragraphs.
- a. Plant heat balances were performed at uprated power and 102% of uprated power. These heat balances defined steady state operating parameters and provided inputs and initial corditions for subsequent plant safety analyses. The heat balance also provides confidence that steady-state operation at uprated power can be achieved routinely,
- b. The power / flow map generated provides information on expected system performance and plant operating domain at uprated power.
- c. Plant instrument setpoiner aere examined to ensure that significant margins were maintained between the safety system settings and the actual safety limits. Certain setpoints were modified so as to main-tain equivalent operating margins at the uprated power level.
The reactor vessel high-pressure scram setpoint and the ATWS recir-culation pump high-pressure trip setpoint were adjusted to account for the increase in reactor operating pressure due to uprated power.
The SRV opening pressure setpoints were raised to account for the increased reactor operating pressure and to improve the SRV simmer margin while maintaining a significant margin to ASME pressure limits during plant transients. All the new setpoints were found to conform to all applicable safety criteria.
i 13-1
- d. The LOCA analyses were performed to demon 3trate conformance with the
, Emergency Core Cooling Systems (ECCS) acceptance criteria of 10CFR50.46 for uprated power conditions. The ECCS performance was evaluated for the entire spectrum uf 'uteak sizes for postulated LOCAs. Results of the LOCA calculations, at uprated power = hew that the ECCS will per-form its function in an acceptable manner and meet all of the 10CFR50.46 acceptance criteria, given the operation at or below the provided MAPLHCR limits.
- e. The pressure relief system was analyzed to ensure that adequate reactor vessel overpressure protection exists during plant operation at uprated power conditions. Results of the analyses demonstrate that the ASME Boiler and Pressure Vessel Code compliance criteria are met using the proposed uprated power setpoints for the SRVs and the SSVs.
- f. The most 1Lniting abnormal operational transients were evaluated for the DAEC at uprated power to ensure the integrity of the fuel clad-ding. Plant operation at or above the resulting MCPR limits will ensure that the fuel cladding integrity safety limit is not exceeded for any abnormal operational transient.
The CRDA was analyzed to determine the consequences of a reactivity excursion. Specific analysis results for the DAEC at uprated con-ditions are bounded by the rod drop design limit peak enti.alpy of 280 cal /gm.
- g. Thermal hydraulic stability analyses were performed for uprated conditions to determine channel hydrodynamic conformance and reactor conformance to the ultimate performance criteria. At the most responsive condition, the channels and the core are within the bounds of the ultimate performance criteria of a 1.0 decay ratio.
13-2
- h. A RIPD analysis was performed at uprated power to confirm that the design bases for reactor internals are met. The analysis examined r
the responses of the reactor vessel internals to loads imposed during steady-state and upset conditions and during certain ccergancy and faulted conditions. The results of the analyses show that the maxi-mum pressure differentials across the components are bounded by the design basis.
E
- i. Structural evaluation of the reactor internals, reactor vessel and vescel nozzles was performed at uprated power conditions. Results of this evaluation demonstrate tnat none of the component stresses will exceed the appropriate design ilmits.
- j. The containment was evaluated in terms of short-term and long-term
[ accident response at uprated power conditions. For both responses, the resulting temperatures and pressures are well below the appro-
- priate design basis limits.
[ .
- k. An ELLLA at uprated power was performed to extend the plant operating 1 domain above the rated rad line. Stability, LOCA, containment 2
response and pressurization transients were analyzed to verify safe operation within the extended power / flow region. From the results 5 of these analyses, it is concluded that all safety bases normally applied to the DAEC are satisfied for operation within the uprated power ELLLA region.
(
r L
_ 1. The SLO at uprated power was analyzed to ensure safe plant perform-ance in this mode of operation. It was concluded that the results k from two-loop operation analyses and previous SLO documentation
{ justify SLO at uprated power, using the latest recommendation.
- m. The effect of power uprate on the SRV Low-Low Set (LLS) System was also analyzed. Two 1Luiting events were considered
- , and it was concluded that there is no impact on the g LLS System due to power uprate.
P t 13-3
/
l NEDO-30603-1 These results demonstrate that the revised Technical Specifications will satisfy the established DAEC licensing criteria and, when sup-ported by appropriate auxiliary systems analyses, that the plant will operate at 1658 We without undue risk to the public healch and safety.
13-4
Z NED0-30603-1
- 14. REFERENCES r
b 1. " Safety Evaluation of the Duane Arnold "nergy Center", Docket No. 50-331, f January 1973.
=
I 2. NUREG-0800, " Standard Review Plan for the Review or Safety Analysis E
E Reports for Nuclear Power Plants".
- 3. " Updated Final Safefy Analysis Report - Duane Arnold Energy Center",
( Docket No. 50-331.
t E
4.
" General Electric Boiling Water Reactor Supplemental Reload Licensing m
- Submittal for Duane Arnold Energy Center, Reload 7", 23A1739, June 1984.
{
- 5. " General Electric Boiling Water Reactor Increased Safety / Relief Valve Simmer Margin Analysis for Duane Arnold Energy Center", NEDO-30606, May 1984.
n
- 6. " Technical Specifications and Bases for Duane Arnold Energy Center",
b e Docket No. 50-331.
D b 7. " General Electric Company Analytical Model for Loss-of-Coolant Analysis K
= in Accordance with 10CFR50, Appendix K", NEDE-20566P, November '.975.
=
1
- 8. R. L. Gridley (GE), Letter to D. G. Eisenhut (NRC), " Review of Low-Core Flow Effects on LOCA Analysis for Operating BWRs", May 8,1978.
- 9. D. G. Eisenhut (NRC). Letter to R. L. Gridley, enclosing " Safety Evaluation Report Revision of Previously Imposed MAPLHGR (ECCS-LOCA)
Restrictions for BWRs at Less Than Rated Flow", May 19, 1978.
%r
{ 10. " General Electric Standard Application for Reactor Fuel (Supplement for United States)", NEDE-24011-P-A-6-US , April 1983 and NED0-240ll-A-6-US ,
{
April 1983.
r-14-1
- 11. Regulatory Guide 1.49, " Power Levels of Nuclear Power Plants", Rev 1, December 1973.
- 12. Safety Evnluncion for the General Electric Topical Report Qualification of the One Dimensional Core Transient Model for BM11a;, Water Reactors NEDC-24154 and NEDE-24154-P, Volumes I, II, and III, Reactor Systems Branch, June 1980.
- 13. Review and Evaluation of GETAB (General Electric Thermal Analysis Basis) '
for BWRs, by Technical Review, Directorate of Licensing, United States Atomic Energy Commission, September 1974.
- 14. Letter: J. A. Hinds to W. R. Buller, " Responses to the Third Set of AEC Questions on the General Electric Licensing Topical Reports NEDO-10558 and NEDE-10958, General Electric BWR Thermal Analysis Basis (CETA3):
Data, Correlation and Design Application", July 24, 1974.
- 15. J. F. Carew, " Process Computer Performance Evaluation Accuracy",
NEDO-20340, 74 NED 32, Class I, June 1974.
- 16. "Feedwater Nozzle Safe End and Thermal Sleeve Stress Analysis Addenda for Duane Arnold", Chicago Bridge and Iron Company, March 1972.
- 17. " Mark I Containment Program Load Definition Report", General Electric Company Report NEDO-21888, November 1981.
- 18. " Mark I Containment Program Plant daique Load Definition, - Duane Arnold Energy Center, Unit 1", General Electric Company Report NEDO-24571, March 1982.
- 19. " Mark I Containment Program - 1/4 Scale Plant-Unique Tests", General
, Electric Company Report NEDE-21944-P, April 1979 and NEDO-21944, April 1979.
- 20. "Duane Arnold Energy Center Suppression Pool Temperature Response",
General Electric Company Report NEDC-22082-P, March 1982 and NEDO-22082, March 1982.
14-2
NEDO-30603-1 21.
- 22. "Ccncral Elcetric Boiling Water Reactor Extended Luad Line Limic Analysis for Duane Arnold Energy Center, Cycle 8", NEDC-30626, May 1984.
f
- 23. General Electric Service Information Letter (SIL) No. 380, Rev. 1, "BWR Core Thermal Hydraulic Stability", February 1984.
- 24. " General Electric Conpany Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K Anendnent No. 2 - One Recirculation ..
Loop Out-of-Service", General Electric Company Report NEDO-20566-2, July 1978.
- 25. "Duane Arnold Energy Center, Single-Loop Operation", General Electric Co=pany Report NEDO-24272, July 1980.
F *[
26.
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