ML20071N682
| ML20071N682 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/30/1982 |
| From: | Fischer D, Gridley R GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20071N666 | List: |
| References | |
| 82NED084, 82NED84, NEDO-22198, NUDOCS 8306070174 | |
| Download: ML20071N682 (21) | |
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! GENERAL ELECTRIC BOILING
! WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR PILGRIM NUCLEAR POWER STATION UNIT 1 CYCLE 6 l
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NEDO-22198 DRF L12-00547 82NED084 Class I September 1982 GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED LOAD LINE LIMIT ANALYSIS FOR PILGRIM NUCLEAR POWER STATION UNIT 1 CYCLE 6
- k Approved: > /d [E Z,_
Approved:
D. L. Fischer, Manager R. jf. Gridley,#Managfr Core Nuclear Design Fuel and Services Licensing NUCLEAR POWER SYSTEMS DIVISION + GENERAL ELECTRIC COMPANY SAN JOSE, CAllFORNIA 95125 GENER AL $ ELECTRIC
NEDO-22198 - -
IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Boston Edison Company (BECo) for BECo's use in supporting the operation of Pilgrim Nuclear Power Station. The information contained in this 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 General Electric Company Load Line Limit Analysis Proposal No. 424-TY591-HK1, October 29, 1981. The use of this information except as defined by said contract, 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 contributors 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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NEDO-22198 CONTENTS Page
- 1.
SUMMARY
1-1
- 2. INTRODUCTION 2-1.
. 3. DISCUSSION 3-1
- 3. l' Background 3-1
' 3. 2 . Analytical Basis 3-1
. 3.3 Existing License Basia 3-2 3.4- Analysis and Results 3-3 3.4.1 Stability 3 3.4.2 Loss-of-Coolant Accident. 3-4 3.4.3 Pressurization Transients 3-4 3.4.4 ASME Pressure Vessel Code Compliance 3-5 3.4.5 Rod Withdrawal Error 5 3.5 Conclusion 3-5
- 4. APPLICATION '4-1
- 5. REFERENCES- 5-1 l
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NEDO-22198 TABLES Table Title Page 3-1 ' Transient Input Data and Operating Conditions 3-6
- 2 GETAB Analysis Initial Conditions 3-7
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'3 Pressurization Transient Results 3-8' 3-4 ASME Pressure Vessel Code Compliance: MSIV Closure, Flux Scram 3-8 4
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NED0-22198 ILLUSTRATIONS Figure Title Page 1-1 Pilgrim Proposed Operating Power / Flow Map 1-2 3-1 Pilgrim Operating Map (FSAR) 3-9 4-1 Core Flow-Recirculation Flow Relationship for Jet-Pump Plants 4-2 vii/viii
NED0-22198 1.-
SUMMARY
This report justifies the expansion of the operating region of the power /
flow map for Pilgrim Nuclear Power Station Unit 1 Cycle 6. The operating envelope is modified to include the extended operating region bounded by the 108% APRM rod block line, the rated power line, and the rated load line, as shown in Figure 1-1.
The underlying-technical analysis is referred to as the Extended Load Line Limit Analysis (ELLLA).
The discussion and analyses presented show that the consequences of events initiated from within the extended domain are bounded by the consequences of the same events initiated from the license basis condition.
Therefore, it is shown that all safety bases normally applied to Pilgrim Nuclear Power Station Unit 1 are satisfied throughout Cycle 6 for operation within the extended operating region.
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NEDO-22198 140 l
100% POWER LINE 120 -
100% INTERCEPT POINT (100/87)
APRM PJO BLOCK LINE (108/100)
(PROPOSED) (0.58W + 50%)
0 00/100) 100 _
TYPICA L POW ER ASCENSION PATH 80 - ,
d ANALYSIS NEEDED TO c:
OPER ATE IN THIS REGION y
2 TYPICAL 100% POW ER/
100% FLOW LOAD LINE 60 -
I l MINIMUM PUMP SPEED LINE e -
I j NATURAL CIRCULATION
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10 20 30 40 50 60 70 80 90 100 110 CORE FLOW (%I Figure 1-1. Pilgrim Proposed Operating Power / Flow Map 1-2
NED0-22198
- 2. INTRODUCTION j
Two' factors which restrict the flexibility of a Boiling Water Reactor (BWR) during power ascension in proceeding from the low-power / low-core-flow condition to the high-power /high-core-flow condtiion are: (1) the Final Safety Analysis Report (FSAR) power / flow curve, and (2) Preconditioning Interim
. Operating Management Recommendations (PCIOMRs).
. If the rated load line control rod pattern is maintained as core flow
- is increased, changing equilibrium xenon concentrations will result in less than rated power at rated core flow. In addition, fuel pellet-cladding inter-
+ action considerations inhibit withdrawal of control rods at high power levels.
The combination of these two factors can result in the inability to attain rated core power directly.
. This report provides the analytical basis for Pilgrim Nuclear Power Sta-tion Unit 1 operation during Cycle 6 under a modified operating envelope to permit improved power ascension capability to full power within the design bases previously applied.
The operating envelope is modified to include the extended operating region bounded by the 108% Average Power Range Monitor (APRM) rod block line, the rated power line, and the rated load line as shown in Figure 1-1.
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- 3. DISCUSSION
3.1 BACKGROUND
Operation of Pilgrim Nuclear Power Station Unit I utilizing the power /
flow map is described in Chapter 3 of the FSAR (Reference 1) . This section of
- the FSAR describes the basic operating envelope (Figure 3.7-1) within which normal reactor operations are conducted and provides the basic philosophy behind the power / flow curve. FSAR Figure 3.7-1 is reproduced as Figure 3-1 of this document. Justification for expansion of this operating region was pro-vided by a subsequent analysis as reported in References 2 and 3.
This analysis expands the operating domain beyond that of Reference 2 along the 108% APRM rod block
- line to 100% power at 87% flow. Rated power operation at any flow between 87% and 100% is acceptable within the constraints of the rod block monitor system. Figure 1-1 shows the proposed operating map.
3.2 ANALYTICAL BASIS To provide relief from the operating restrictions inherently imposed during ascension to power by the existing power / flow curve and PCIOMRs, a modified power / flow curve has been derived. In deriving this operating curve, five design basis objectives were specified:
- 1. For those transients and accidents that are sensitive to variations.
in power and flow, the 100% power /100% flow (licensing basis for BWR/2 and 3's) point must be shown to be a more limiting condition than any condition within the expanded operating region (i.e., the shaded region of Figure 1-1).
. 2. In no instance shall the ratio of power to flow intentionally exceed the ratio defined by the APRM rod block line.
- APRM Rod Block = 0.58 W+50% where W is recirculation flow in percent of rated.
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- 3. The slope of the APRM rod block line must be such that flow increases are capable of compensating for xenon buildup while increasing reactor power.
- 4. The consequences of all accidents and transients analyzed in the FSAR and subsequent amendments and the license submittals must remain within the limits normally specified for such events.
- 5. Reactor power ascension from minimum recirculation pump speed to full power shall be directly attainable through combined control rod move-ment and recirculation flow increase without violation of either the power / flow line or PCIOMRs.
The analyses to meet these objectives for tue Reference 2 expanded operat-ing domain were described in References 2 and 3. To meet these objectives for the Extended Load Line Limit Analysis region (Figure 1-1), additional analyses were performed for Cycle 6. All other analyses and conclusions reported in References 2 and 3 remain valid. From these additional analyses, conclusions are drawn concerning *.he safety consequences of operation in the extended operating region (shaded area of Figure 1-1).
3.3 EXISTING LICENSE BASIS The original reload licensing analysis for Pilgrim Cycle 6 supported oper-ation for a standard end-of-cycle target exposure distribution. The analysis results and Minimum Critical Power Ratio (MCPR) operating limits are given in Reference 4.
Subsequently, a revised end-of-cycle target exposure distribution was .
established. The reload licensing analysis was revised to reflect the new end-of-cycle target exposure shape. This revised reload analysis and MCPR
- operating limits are given in Reference 5.
This extended load line limit analysis was therefore performed using the revised end-of-cycle target exposure distribution and is consistent with the license basis of Reference 5.
3-2
'3.4 ANALYSIS AND RESULTS 3.4.1 Stability 3.4.1.1 Channel Hydrodynamic Conformance to the Ultimate Performance Criterion The channel performance calculation for Pilgrim Nuclear Power Station
. Unit 1 Cycle 6 is presented in Reference 5. The decay ratios are reproduced below:
i Extrapolated Rod Block Line* -
Channel Hydrodynamic Performance Natural Circulation Power Channel Type Decay Ratio P8x8R 0.18 8x8 0.22 j
- At this most responsive condition, the most responsive channels are clearly within the bounds of the ultimate performance criteria of <l.0 decay ratio at all attainable operating conditions.
3.4.1.2 Reactor Conformance to Ultimate Performance Criterion The decay ratios determined from the limiting-reactor core stability con-ditions are presented in Reference 5. The limiting condition for this analysis is the intersection of the extrapolated.108% APRM rod block line* and the natural circulation flow line.
1 Extrapolated Rod Block Line* -
Reactor Core Stability Natural Circulation Power
. Decay Ratio, X /Xg 0.65 These calculations show Pilgrim to be in compliance with the ultimate performance criteria, including the most responsive condition.
- APRM Rod Block < 0.58 W+50%, where W is recirculation flow in % of rated.
3-3
WEDO-22198 3.4.2 Loss-of-Coolant Accident The standard Loss-of-Coolant Accident (LOCA) analysis for Pilgrim 1 is applicable for plant operation in the power / flow domain bounded by the most limiting of the following:
- a. 100% core power, ,
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- b. 100% core flow, and .
- c. APRM rod block line.
The proposed operating domain for the Pilgrim Nuclear Power Station is within the above limits. In addition, according to Reference 6, Pilgrim has a 0.95 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) multiplier when operating at core flows of less than 90% of rated flow, and is restricted to a minimum value for MCPR operating limit of 1.24 due to Emergency Core Cooling System considerations.
A discussion of low-flow effects on LOCA analyses for all operating plants (Reference 6) has been presented to and was approved by the United States Nuclear Regulatory Commission (USNRC) (Reference 7). The LOCA analysis for.
'- Pilgrim Nuclear Power Station (contained in Reference 8) is applicable in the power flow domain discussed in this report.
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3.4.3 Pressurization Transients i
As shown in Reference 5, the most limiting pressurization transient for Pilgrim Cycle 6 is the Load Rejection without Bypass event. The Reference 5 .
results were calculated using a revised end-of-cycle target exposure distribu-
- tion which is more bottom peaked than that used in Reference 4. ,
l' For the extended load'line limit analysis, the Load Rejection without
! Bypass and the Feedwater Controller Failure events were analyzed at the limit-
. ing point in the extended operating region, which is the 100% intercept point, f
, (100%P, 87%F). These analyses were performed using the nuclear parameters I
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NEDO-22198 resulting from the same End-of-Cycle (EOC) target exposure shape used in the Reference 5 analyses.
The results for both the licensing basis case and the reduced flow case are shown in Table 3-3. The initial conditions for these cases are shown in Tables 3-1 and 3-2. As shown in Table 3-3, the (100, 87) point results are
. bounded by the Table 3-3, licensing basis (100,100) point results.
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. 3.4.4 ASME Pressure Vessel Code Compliance All Main Steam Isolation Valve (MSIV) closure with no scram event is used to determine compliance to the American Society of Mechanical Engineers (ASME) pressure vessel code. This event was analyzed at the 100% intercept point (100%P, 87%F) using the nuclear parameters resulting from the same EOC carget exposure shape used in the Reference 5 analyses. The results are compared to the results for the licensing basis point in Table 3-4. As shown, the licensing basis point results are bounding.
3.4.5 Rod Withdrawal Error The effective Rod Block Monitor (RBM) setpoint is a function of power and flow. Above the rated rod line, the rod block will occur with less rod with-drawal. Thus, the evaluation at rated is conservative for operation above the rated load line.
3.5 CONCLUSION
' -The results of the limiting transients for the limiting point in the
,citended: operating regio 2 (100, 87) are less severe than the same transients l
f or 'tih'he license basis point. The overpressure protection analysis results are n s* a s
' ulso less severe.for thel,100, 87 point. The stability results are the same 1
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N 'as those reported in the reload license submittal and the MAPLHGR results are
. , , . ;g unchanged.by the extended operating region.
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Therefore it is, concluded that all safety bases for the extended operating region are bounded by the license basis condition and therefore operation in
- the ' extended region is justified.
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NEDO-22198 Table 3-1 TRANSIENT INPUT DATA AND OPERATING CONDITIONS Licensing Basis Point 100% Intercept Point (100/100) (100/87)
Thermal Power (Mwt/%) 1998/100 1998/100 Steam Flow (Mlb/hr/%) 7.98/100 7.97/100 .
Core Flow (M1b/hr/%) 69.0/100 60.0/87 Dome Pressure (psig) 1025 1025 Turbine Pressure (psig) 979 979 Relief Valve Setpoint 1125 1125 (psig)
Relief Valve Capacity 4/41.1 4/41.1 (No./% NBR)
Safety Valve Setpoint 1240 1240 (psig)
Safety Valve Capacity 2/16.5 2/16.5 (No./% NBR) 3-6
NEDO-22198 Table 3-2 CETAB ANALYSIS INITIAL CONDITIONS Licensing Basis Point 100% Intercept Point (100/100) (100/87)
Core Power (MWt) 1998 1998 Core Flow (Mlb/hr) 69.0 60.0 Reactor Pressure (psia) 1065 1063 Inlet Enthalpy (Btu /lb) 526.6 522.7 Nonfuel Power Fraction 0.035 0.035 Axial Peaking Factor 1.40 1.40 8x8 Fuel Local Peaking Factor 1.22 1.22 Radial Peaking Factor 1.51 1.50 R-factor 1.098 1.098 Bundle Power (MWt) 5.096 5.039 Bundle Flow (10 lb/hr) 100.21 85.0 P8x8R Fuel Local Peaking Factor 1.20 1.20 Radial Peaking Factor 1.63 1.61 R-factor 1.052 1.052 Bundle Power (MWt) 5.483 5.411 Bundle Flow (10 lb/hr) 100.8 85.6 3-7
NEDO-22198 Table 3-3 PRESSURIZATION TRANSIENT RESULTS Initial , p p ACPR Power / Flow $ (% Q/A (% sl v
(% NBR) initial) initial) (psig) (psig) 8x8 P8x8R Load 100/100* 597 123 1301 1310 0.33 0.36 ,
Rejection w/o 100/87** 579 122 1293 1306 0.30 0.33 Bypass Feedwater 100/100 385 123 1189 1216 0.28 0.30 Controller Failure 100/87 393 124 1193 1215 0.26 0.28
- Licensing basis point
- 100% intercept point (Most limiting point in the extended operating region)
Table 3-4 ASME PRESSURE VESSEL CODE COMPLIANCE: MSIV CLOSURE, FLUX SCRAM Peak Peak Steam Line Vessel Peak Peak Pressure Pressure Initial Neutron Heat P P Power / Flow Flux 4 Flux Q/A s1
(% NBR) (% initial) (% initial) (psig) (psig) 487 128 1346 1360 100/100*
468 127 1344 1358 100/87**
- Licensing basis point l
- 100% intercept point (most limiting point in the l extended operating region) ,
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NATURAL CIRCULATION LINE 7 100 -
60*Y 20% PUMP SPEED LINE CONSTANT PUMP 00 SPEED LINE f*9 /
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20 - 2 RECIRC PUMP NPSH LIMIT LINE JET PUMP NPSH , )
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0 20 40 00 e0 100 120 PtiRCENT RATED CORE FLOW Figure 3-1. Pilgrim Operating Map (FSAR) l 3-9/3-10 1 , _ _ . _ . . _ _ _ , . . _ , _ . ___ _ _ _ _ _ _ _
- 4. APPLICATION The method of analyses described in this report in support of operation along the modified power / flow line are of a bounding type that can be applied to evaluate BWR/3 and BWR/4 plants whose operation is guided by a power / flow curve.
, The rod block intercept point of 100% power /87% flow lies along the APRM flow-biased rod block line having a slope represented by the equation:
0.58W + 50%
where W = recirculation flow rate in percent of rated The relationship between core flow and recirculation flow is shown in Figure-4-1.
1 Currently, most BWRs operate on the basis of a power / flow curve approximated by the equation:
0.65W + 35%*
with the APRM flow-biased rod block represented by the equation, 0.66W + 42%*
The less restrictive equation (0.58W + 50%) was approved by.the USNRC (Reference 7) and the analyses for this report were performed with this line as the upper bound of the proposed operating envelope.
Operation utilizing the current Pilgrim Nuclear Power ' tation Unit 1 tech-nical specification rod block line (0.66W + 42%) can be effected in the same manner as using the proposed APRM rod block line, except the intersection with the 100% power line would occur at slightly higher flow (Figure 1-1). This is within the analyzed envelope and, therefore, conforms with the bases and conclusions of this report.
- Several plants vary a few percent from these values.
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Figure 4-1. Core Flow-Recirculation Flow Relationship for Jet-Pump Plants 4-2
NED0-22198
- 5. REFERENCES
- 1. " Final Safety Analysis Report, Pilgrim Nuclear Power Station."
- 2. " Pilgrim Nuclear Power Station Load Line Limit Analysis License Amendment
- Submittal," General Electric Company, September 1977 (NED0-24058).
- 3. " Supplement 2 to Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1 Reload 4 (Load Line Limit Analysis Reverification)," General Electric Company, April 1981 (NEDO-24224-2, Supplement 2).
- 4. " Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1, Reload 5," General Electric Company, August 1981 (Y1003J01A28).
- 5. " Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1, Reload 5," General Electric Company, September 1982 (Y1003J01A28, Revision 1).
- 6. 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.
- 7. D. G. Eisenhut (NRC), letter to R. L. Gridley, enclosing " Safety Evaluation Report Revision of Previously Imposed MAPLHCR (ECCS-LOCA) Restrictions for BWRs at Less Than Rated Flow," May 19, 1978.
- 8. " Loss-of-Coolant Accident Analysis Report for Pilgrim Nuclear Power Station,"
General Electric Company, August 1977 (NED0-21696).
. 9. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 59 to Provisional Operating License No. DPR-19, Amendment No. 52 to Facility Operating License No. DPR-25, Amendment No. 70 to Facility Operating License No. DPR-29, and Amendment No. 64 to Facility Operating License No. DPR-30, Commonwealth Edison Company and Iowa-Illinois Gas and Electric Company, Dresden Station Unit Nos. 2 and 3, Quad Cities Station Unit Nos. 1 and 2, Docket Nos. 50-237, 50-249, 50-254, and 50-265.
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