ML20214E477
| ML20214E477 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/06/1986 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20207B375 | List: |
| References | |
| DRF-L12-00754, DRF-L12-754, NEDC-31323-ERR, NUDOCS 8611240454 | |
| Download: ML20214E477 (3) | |
Text
w
, NUCLEAR ENERGY BUSINESS OPERATIONS e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 9512S GENER AL $ ELECTRIC APPLICABLE TO:
Nac- m rusuCATiON NO.
ERRATA And ADDENDA T. s. E. NO. N/A 5Hm TITLE TMenFA9m ennF FT N AE PADT- NO. T TAT. FFFDLTATFR HFATTNC AMAT YSTS FOR 11-6-86 OATE T.TMFRTrif CFM . 9TATTnM TNTT 1 CTCLE 1 NO TE: Correctallcopies of the applicable ISSUE OATE OCTORFR 1986 publication as speci/ led be/0w.
REFERENCES INSTRUCTIONS ITEM ,js,EcT[c,N, p A]NE) (cO AptCTIONS AND ADDITIONS) g
'I Page 1-1 Remove and repla'ce Page 1-1 with attached Page 1-1.
4 8611240454 861117 p4GE lof1 PDR ADOCK 05000352 P PDR
- 1. INTRODUCTION AND
SUMMARY
This report presents the results of a safety and impact evaluation for the operation of the Limerick Generating Station (LGS) Unit I with increased core flow (ICF) and/or partial feedwater heating (PFH). The ICF region provides an expanded operating envelope to permit operation to compensate for reactivity reduction due to exposure during an operating cycle. The ICF region is bounded by the 105% core flow line as illustrated in Figure 1-1.
Operation with PFH occurs in the event that certain stage (s) or string (s) or individual heater (s) (i) becomes inoperable during the fuel cycle resulting in feedwater heater out-of-service (FWHOS) or (ii) is intentionally valved out for operation at exposure beyond standard End-of-Cycle 1* (EOC1), resulting in ,
final feedwater temperature reduction (FFWTR). This report includes evalua- ,
tions to justify PFH operation from a normal rated feedwater temperature of about 420*F to a reduced feedwater temperature of approximately 360*F at 100%
power and reactivity coastdown to a minimum feedwater temperature of approximately 331*F (about 70% power) at EOC1.
l l
In order to evaluate operation with ICF and PFM, the limiting abnormal
' operational trans'ients reported in the Final Safety Analysis Report (FSAR),
Reference 1, for rated flow operation ware reevaluated at EOC1 at 105% core flow with and without PFH. The less-of-coolant accident (LOCA), fuel loading error accident, rod drop accident, and rod withdrawal error event were also reevaluated for ICF operation and/or PFH. These events were also reevaluated for EOC1 operation with ICF.
In addition, the effect of increased pressure differences (due to ICF) on the reactor internals components, fuel channels, and fuel bundles was also analyzed to show that the design limits will not be exceeded. The effect of
- EOC1 is defined as the core average exposure at which there is sufficient reactivity to achieve rated thermal power with rated core flow, all control rods withdrawn (beyond Rod Position 24), all feedwater heaters in service and equilibrium xenon.
1-1
NEDC-31323 DRF L12-00754 Class II October 1986 INCREASED CORE FLOW AND PARTIAL FEEDWATER HEATING ANALYSIS FOR LIMERICK GENERATING STATION UNIT 1 CICLE 1 J.M. Grau S. Wolf Approved: Approved:
A.E. Rogers, Manager D.J. Robare, Manager Plant Performance Engineering Licensing Services i
I f6MPt@v/6 .
NUCLEAR ENERGY DUSINESS OPERATIONS + GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 9512$
GENER AL $ ELECTRIC
NEDC-31323 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Caretully The only undertakings of General Electric Company respecting information in this document are contained in the contract between Philadelphia Electric Company (PECo) and General Electric Company, as identified in the purchase order for this report and nothing contained in this document shall be i
construed as changing the contract. The use or this information by anyone other than PECo or for any purpose other than that tor which it is intended, is not authorized; and with respect to any unauthorized use General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness ot the information contained in this document.
4
)
e NEDC-31323 CONTENTS E.*K*
l ABSTRACT iv ACKNOWLEDGEMENTS v
- 1. INTRODUCTION AND
SUMMARY
1-1
- 2. SAFETY ANALYSIS 2-1 2.1 Abnormal Operational Transients 2-1 2.1.1 Limiting Transients 2-1 2.1.2 Overpressurization Analysis 2-2
- 2.1.3 Rod Withdrawal Error 2-3 j 2.2 Fuel Loading Error 2-3 2.3 Rod Drop Accident 2-3 2.4 Loss-of-Coolant Accident Analysis 2-3 2.5 Thermal-Hydraulic Stability 2-4 2.6 Impact on Anticipated Transients Without Scram (ATWS) 2-5 4 3.
- MECHANICAL EVALUATION OF REACTOR INTERNALS AND FUEL ASSEMBLY 3-1
- 3.1 Loads Evalustian 3-1 i 3.2 Loads Impact 3-2 3.2.1 Reactor Internals 3-2 3.2.2 Fuel Assemblies 3-2
- 4. FLOW-INDUCED VIBRATION '
4-1
- 5.1 Method and Assumption 5-1 1 5.2 Feedwater Nozzle Fatigue 5-2
- 5.3 Feedwater Sparger Fatigue 5-3
- 6. CONTAINMENT ANALYSIS 6-1 i 7. OPERATING LIMITATIONS 7-1 7.1 Feedwater Heaters 7-1 7.2 Operating Map 7-1 7.3 MCPR Operating Limits 7-1 7.4 K g Factor 7-1 I 8. REFERENCES 8-1 i
l i
I 4
}
f l
TABLIS ,,
Table Title Page 2-1 Core-Wide Transient Analysis Results at ICF.and/or PFH 2-7 2-2 Required Operating Limit CPR Values at ICF and/or PFH ,
2-8 2-3 Overpressure Analysis Results 2-9 2-4 ATWS Analysis Results-Limiting MSIV Closure Event 2-10 5-1 Feedwater Nozzle Fatigue Usage for a 32-Year Seal' Refurbishment Period - FFWTR 5-5 5-2 Feedwater Nozzle Fatigue Usage for a 32-Year Seal ,
Refurbishment Period - FWHOS '5-6 5-3 Feedwater Sparger Fatigue Usage for a 32-Year Seal Refurbishment Period - FFWTR 5-7 5-4 Feedwater Sparger Fatigue Usage for a 32-Year Seal Refurbishment Period - FWHOS 5-8 ILLUSTRATIONS -
Fiaure Title Page 1-1 Limerick Unit 1 Power / Flew May 1-3 2-1 Generator Load Rejection with Bypass Failure at 104.3%
Power,105% Core Flow and Normal Feedwater Temperature 2-11 2-2 Feedwater Controller Failure, Maximum Demand, at 104.5%
Power, 105% Core Flow and Reduced Feedwater Temperature 12 2-3 Feedwater Controller Failure, Maximum Demand, at 100%
Power, 87% Core Flow and Reduced Feedwater Temperature 2-13 2-4 Feedvater Controller Failure, Maximus Demand, at 100% ,
Power, 87% Core Flow and Reduced Feedwater Temperature ;
at EOC1-2000 Exposure 2-14 2-5 Feedwater Controller Failure, Maximus Demand, with Turbine Bypass Inoperable at 100% Power, 87% Core Flow and Reduced Feedwater Temperature - 2-15 2-6 Feedwater Controller Failure, Maximum Demand, with Turbine Bypass Inoperable and RPT Inoperable at 104.5% Power, 105% -
Core Flow and Reduced 7sedwater Temperature L2-16 2-7 MSIV Closure Flux Scram, at 104.3% Power, 105% Core Flow and Normal Feedwater Temperature 2-17
-iii-
ABSTRACT i A safety evaluation has been performed to show that Limerick Generating
) Station (LGS) Unit 1 can increase core flow to operate within the region of the operating map bounded by the line between 100% power, 100% core flow (100, 100) and 100% power, 105% core flow (100, 105) throughout Cycle 1. This safety evaluation includes analysis to show that LGS Unit I can operate with partial feedwater heating (PFH) resulting from cartain stages or strings or i individual feedwater heaters out of service (FWHOS) during the cycle or valved
< out at the end of the cycle (final feedwater temperature reduction (FFWTR)),
up to a maximum feedwater temperature reduction of 160*F at rated power and up to a maximum core flow of 105% of rated. After reaching End-of-Cycle 1 (EOC1) +
exposure (depletion of full-power reactivity under standard feedwater I
conditions) with all control rods out, LGS Unit 1 can continue to operate in the region of the operating map bounded by the 105% core flow line between 100% power and the cavitation interlock power with or without FFWTR of 160*F at rated power.
I .
Th's minimum critical power r'atio (MCPR) operating limits will be changed from the values established by the'F'inal Safety Analysis Report licensing
! submittal to the appropriate values (Table 2-2) for increased core flow (ICF) and PFH operating conditions. All other operating limits established in the Cycle 1 licensing basis have been found to be bounding for the ICF and PFH operations as defined above.
a 4
1 i
t 3
! -iv-i
t "
T 4 .
\-
--. "i . ,
14,', ACKNOWLEDGEMENTS
!\
s The analyses reported in this r'eport were performed by the combined efforts of 3\ ,
- many individual contributors, including
I i w I g C.S. Chen, G.G. \ Chen, P.C. Gee, J.M. Grau, K. Jahanian, 3m i l' C.S. Kennedy, S.T. Lam, M.O.'Lenz, H.X. Nghies, C.R. Parker, O. Raza, G.L.sStevens, S. Wolf and W. M. Wong.
, s T
s i
'1
~
7 0
4 i #
\
k
\
\
s
-V-
.\. - -- .- .. - .- , , . . - - - _ - - . , . . __ _ . . . - . ., - _ . . . - - - , . . ., _ . .-
- 1. INTRODUCTION AND
SUMMARY
This report presents the results of a safety and impact evaluation for the operation of the Limerick Generating Station (LGS) Unit I with increased core flow (ICF) and/or partial feedwater heating (PFH). The ICF region provides an expanded operating envelope to permit operation to compensate for reactivity reduction due to exposure during an operating cycle. The ICF region is bounded by the 105% core flow line as illustrated in Figure 1-1.
Operation with PFH occurs in the event that certain stage (s) or string (s) or individual heater (s) (i) becomes inoperable during the fuel cycle resulting in feedwater heater out-of-service (FWHOS) or (ii) is intentionally valved out for operation at exposure beyond standard End-of-Cycle 1* (EOC1), resulting in final feedwater temperature reduction (FFWTR). This report includes evalua-tions to justify PFH operation from a normal rated feedwater temperature of about 420*F to a reduced feedwater temperature of approximately 360*F at 100%
power and reactivity coastdown to a minimum feedwater temperature of approximately 331*F (about 70% power) at EOC1.
- In order to evaluate operation with ICF and PFH, the limiting abnormal operational transients reported in the Final Safety Analysfa Report (FSAR),
Reference 1, for rated flow operation were reevaluated at EOC1 at 105% core flow with and without PFH. The loss-of-coolant accident. (LOCA), fuel loading error accident, rod drop accident, and rod withdrawal error event were also reevaluated for ICF operation and/or PTH. These events were also reevaluated for EOC1 operation with*ICF.
In addition, the effect of increased pressure differences (due to ICF) on the reactor internals components, fuel channels, and fuel bundles was also analyzed to show that the design limits will not be exceeded. The effect of
- EOC1 is defined as the core average exposure at which there is sufficient reactivity to achieve rated thermal power with rated core flow, all control rods withdrawn, all feedwater heaters in service and equilibruim xenon.
1-1
NEDC-31323 ICF on the flow-induced vibration response of the reactor internals was also evaluated to, ensure that the response is within acceptable limits. The thermal-hydraulic stability was evaluated for ICF/PFH operation, and the increase in the feedwater nozzle and feedwater sparger usage factors due to the feedwater temperature reduction was determined.- The impact of ICF/PFH operation on the containment LOCA response was also analyzed.
The results or the safety evaluation show that the current technical specifications with incorporation of the MCPR limits of Table 2-2 are adequate to preclude the violation of any safety limits during Cycle 1 operation of LGS Unit I within the expanded operating map, including the ICF region, as illus-trated in Figure 1-1, with or without PFH with the conditions assumed in the analysis. The CPRs and the minimum critical power ratio (MCPR) operating limits for plant operation are given in Tables 2-1 and 2-2. The EOC1 Option A and Option B MCPR limits (Reference 1) will be increased to the appropriate values as shown in Table 2-2.
i 1-2 I
NEDC-31323 ISO 140 -
WD = RECIRCULATION LOOP 120 -
DME W (100/871 (100/105) 100 -
7 g APRM ROD BLOCK UNE
- ANALYSIS (0.58WD + 50%)
80 -
ELLLA REGION
/
~
100% LOAD UNE
< s INCREASED CORE
/:'
FLOW REGION MINIMUM PUMP SPEED NATURAL CIRCULATION 20 -
CAVITATION PROTECTION O I I '
O 20 40 60 80 100 120 CORE FLOW (%)
Figure 1-1. Limerick Unit 1 Power / Flow Map 1-3
, NEDC-31323 9
- 2. SAFETY ANALYSIS 2.1 ABNORMAL OPERATIONAL TRANSIENTS 2.1.1 Limiting Transients All abnormal operational transients analyzed in the Cycle 1 FSAR I
licensing submittal (Reference 1) were examined for effects caused by ICF and/or PFH operation. Both ICF and PFH operation result in a decrease in core voids, which, in turn, results in a less bottom-peaked power distribution and a less effective scram reactivity. Two limiting abnormal operational transients were reevaluated in detail: generator load rejection with bypass failure (LRNBP) and feedwater controller failure to maximum demand (FWCF).
Nuclear transient data for 104.5% power *, 105% core flow (104.5, 105) with and without PFH were developed based on the Haling method at rated power for EOC1. Analyses were also performed to integrate the combination of Extended Load Line Limit Analysis (ELLLA)** and PFE.
The limiting abnormal operational transient based on the ICF/PFH/ELLLA 4 combinations was also analyzed at a fuel exposure of EOC1 minus 2000 mwd /st in order to demonstrate that it is acceptable for Limerick to operate with the FWHOS option during the cycle.
! The limiting abnormal operational transients were also analyzed for two additional equipment out-of-service cases: (1) turbine bypass (BP) 1 inoperable; and (2) both the BP system and recirculation pump trip (RPT) system inoperable. These results are provided for use in modifying the MCPR operating limits for applicable systems inoperable as contained in the current technical specifications described in Reference 1.
- The power level corresponding to this condition will vary from 104.5% to
- . 104.3%, depending on whether tinal feedwater heaters are in service. The j 104.5% power level provides a 5% steam flow margin to the 100% power operating condition to be consistent with the original FSAR analysis.
- (100, 87) is the limiting ELLLA condition, per Reference 2.
l 2-1
_ _ , _ .-- _- _ _ .. _ _ _ _ = . _ _ _ _ _ _ _ -
The results of the transient analyses are presented in Tables 2-1 and 2-2 with the limiting transient results previously documented in the FSAR licensing submittal (Reference 1). The transient performance responses are presented in Figures 2-1 through 2-6.
The operating limit critical power ratios (OLCPRs) for operation with the RPT and BP systems operable are included in Table 2-2. The Option A OLCPR is based on the LRNBP event. In this case, ICF without PFH bounds ICF with PFH and the Option A OLCPR = 1.23 exceeds the value determined for the FSAR licensing submittal, The Option B OLCPR, based on the loss of feedwater heating (LOFH) event, remains the same as determined for the FSAR.
For the BP system inoperable, both the Option A and Option B OLCPR values are determined by the FWCF event at ELLLA (100, 87) conditions. For the WCF event, conditions with PFH bound corresponding conditions without PFH. For both the BP and RPT systems inoperable, the FWCF event at the limiting ICF condition (104.5, 105) bounds the ELLLA condition and determines the Option A
,- and B OLCPR values.
Decreasing the power from the 100% rated condition along the 105% core flow line will result in an increase in transient ACPR for some events. This increase is less than the increase in operating CPR due to power decrease, and hence such operation will not result in violation of the safety limit MCPR in the event of a transient (Reference 3, Page 2-12).
2.1.2 Overpressurization Analysis l
The limiting transient for ASME code overpressurization analysis, main steam isolation valve (MSIV) closure with flux scram (direct scram failure),
was evaluated for the extended EOC1 conditions with ICF without PFH (Table 2-3 and Figure 2-7). For this evaluation, ICF without PFH is more severe than ICF with PFH. The ICF for the LRNBP event results in a less severe overpressure transient than MSIV closure with flux scram. The overpressurization analysis (Table 2-3) for the ICF region produced a peak vessel pressure of 1273 psig, which is below the ASME Code upset limit of 1375 psig and is, therefore, acceptable.
2-2
NEDC-31323 2.1.3 Rod Withdrawal Error The rod withdrawal error transient was evaluated under ICF and/or PFH conditions. When ICF is employed, the rod block monitor (RBM) setpoint (which is flow biased) increases, giving higher MCPR limit. Thus, the RBM should be clipped at flows greater than 100% of' rated so that the CPR values (Refer-ence 1) determined without ICF apply. The clipping procedure includes an adjustment to the RBM circuit so that the high RBM trip setpoint at flows greater than 100% of rated is equal to the value at 100% rated flow.
2.2 FUEL LOADING ERROR This event is not adversely affected by ICF and/or PFH operation. Based on a General Electric generic study, the impact of ICF and/or PFH or CPR is expected to be small compared with the margin to the OLCPR. Thus' the FSAR OLCPR is not affected by this event under ICF and/or PFH conditions.
2.3 ROD DROP ACCIDENT 1
Limerick 1 uses banked position withdrawal sequence (BPWS) for control rod movement. Control Rod Drop Accident (CRDA) results from BPWS plants have been analyzed statistically. The results show that, in all cases, the peak fuel enthalpy in a CRDA would be much less than the corresponding design limit even with a maximum incremental rod worth corresponding to 95% probability at the 95% confidence level. Based on these results, it was proposed to the USNRC, and subsequently found acceptable, to delete the CRDA from the standard GE-BWR reload package for the BPWS plants (Reference 3, Section 5.2.5.1.3 Item 1). Hence, the CRDA is not analyzed specifically for Limerick 1.
2.4 LOSS-OF-COOLANT ACCIDENT (LOCA) ANALYSIS Generic LOCA analyses performed for operation with ICF alone show that it bounds operation with ICF and PFH.
The effect of increased core flow on LOCA analyses is not significant ,
because the parameters which most strongly affect the calculated peak cladding 2-3
m NEDC-31323 temperature (PCT), i.e, high power node boiling transition time and core reflooding time, have been shown to be relatively insensitive to increased core flow.
Results of the LOCA analysis performed for Limerick show that the PCT for ICF increases by less than 10*F for the limiting break compared to the rated core flow condition. PCT changes throughout the remainder of the break spectrum will be of a similar magnitude and thus will not alter the limiting break size or failure.
Therefore, it is concluded that the LOCA PCT is acceptable and that the current maximum average planar linear heat generation rates (MAPLHGRs) for Limerick are applicable for ICF.
2.5 THERMAL-HYDRAULIC STABILITY The General Electric Company has established stability criteria to demonstrate compliance to requirements set forth in 10CFR50 Appendix A, General Design Criteria (GDC) 12. These stability compiiance criteria consider potential limit cycle response within the limits of safety system and/or operator intervention and assure that for GE BWR fuel designs this operating mode does not result in specified acceptable fuel design limits being exceeded. Furthermore, the onset of power oscillations for which corrective actions are necessary is reliably and readily detected and suppressed by operator actions and/or automatic system functions. The stability compliance of all licensed GE BWR fuel designs including those fuels contained in the General Electric Standard Application for Reactor Fuel l
(CESTAR II, Reference 3) is demonstrated on a generic basis in Reference 4 (for operation in the normal as well as the extended operating domain with ICF and/or PFH). The NRC has reviewed and approved this in Reference 3; there-fore, a specific analysis for each cycle is not required. The LGS Unit 1 Cycle 1 core contains licensed GE BWR initial core fuel and, hence, the generic evaluation in Reference 4 is applicable to LGS Unit 1.
2-4
\
l NEDC-31323 l
. i For operation in the ICF region, the stability margin (defined by the core decay ratio) is increased as flow increases for a given power. ICF operation is bounded by the fuel integrity analyses in Reference 4.
Similarly, operation in the PFH mode is bounded by the fuel integrity analyses in Reference 4. In general, the effect of reduced feedwater tempera-ture results in a higher initial CPR which yields even larger margins than those reported in Reference 4. The fuel integrity analyses are independent of the stability margin, since the reactor is already assumed to be in limit cycle oscillations. Reference 4 also demonstrates that even if neutron flux limit cycle oscillations did occur just below the neutron flux scram setpoint, fuel design limits nre not exceeded for all licensed GE BWR fuel designs including those fuels contained in General Electric Standard Application for Reactor Fuel (GESTAR II, Reference 3). These evaluations demonstrate that substantial thermal / mechanical margin is available for the GE BWR fuel designs even in the unlikely event of very large oscillations.
) To provide assurance that acceptable plant performance is achieved during i operation in the least stable region of the power / flow map, as well as during all plant maneuvering and operating states, a generic set of operating recom-mandations has been developed as set forth in Reference 5 and communicated to all GE BWRs. These recommendations, which have been implemented in the LGS Technical Specifications, instruct the operator on how to reliably detect and suppress limit cycle neutron flux oscillations should they occur. The recom-mendations were developed to conservatively bound the expected performance of all current product lines and are applicable to operation with PFH (feedwater temperature of approximately 360*F at rated power).
2.6 IMPACT ON ANTICIPATED TRANSIENTS WITHOUT SCRAM (ATWS)
I The limiting Limerick ATWS event analysis is presented in Reference 2.
The ATWS MSIV closure event at the limiting ELLLA condition of (100, 87) was reanalyzed with PFH operation for both the two-pump and three-pump Standby i .
2-5 i
l i
- -- -,._,e--c -g,. - - - . ,----- -- --
NEDC-31323 Table 2-2 REQUIRED OPERATING LIMIT CPR VALUES AT ICF AND/OR PFH Initial Initial Core Core Transient Fuel Power Flow Description" Exposure c
(%NBR) (%NBR) ACPR OLCPR OLCPR **'
LRNBP (FSAR) EOC1 104.3 100 0.08 1.19 1.11 LRNBP 8 h EOC1 104.3 105 0.11 1.23 7,34 WCF (FSAR) EOC1 104.3 100 0.06 1.17 1.14 i
WCF EOC1 104.5 105 0.10 1.21 1.18 WCF EOC1 100.0 87 0.11 1.22 1.19 WCF w/o BP EOC1 100.0 87 0.16 1.27 1.24 WCF w/o BP, EOC1 104.5 105 0.24 1.36 1.28 w/o RPTd .
CPR OLCPR L NH (FSAR) 0.16 1.22
- s. LRNBP = Load rejection with bypass failure; WCF = feedwater controller failure to maximum demand; w/o RPT = RPT. inoperable; w/o BP = turbine
't bypass inoperable; L WH = loss of feedwater heating.
- b. EOC1 = End-of-cycle 1.
- c. ODYN results without adjustment factors, based on initial CPR which yields a minimum CPR of 1.06.
- d. Includes Option A adjustment factors.
l e. Includes Option B adjustment factors.
- f. Option A and B adjustment factors are specified in the NRC safety evaluation report on ODYN (NEDO-24154 and NEDE-24154P).
- g. For LRNBP, ICF w/o PFH bounds ICF with PFH.
- h. Required OLCPR using Option A.
- i. For WCF, condition with PFH bounds condition w/o PFM.
- j. For WCF w/o BP w/o RPT, limiting condition with ICF (104.5% power,105%
core flow) bounds ELLLA condition (100% power, 87% core flow).
- k. Required OLCPR using Option B.
2-8 I
,- .- , , _ . . . , , _ _____-..m. ,_ _
Table 2-3 OVERPRESSURIZATION ANALYSIS RESULTS Maximum Maximum Initial Initial Steamline Vessel Power Flow Pressure Pressure Transient (%) (%) (psig) (psig) Figure No.
MSIV Closure - Flux Scram 104.3 100 1227 1260 Reference 1 (FSAR)
MSIV Closure - Flux Scram 104.3 105 1235 1273 Figure 2-7 (ICF w/o PFH) 2-9
NEDC-31323 Table 2-4 ATWS ANALYSIS RESULTS - LIMITING MS V CLOSURE EVENT 2-Pump SLCS 3-Pump SLCS 2-Min Timer 5-Min Timer Parameter Rated FW* PFH** Rated FW PFH Initial Power / Flow 100/87 100/87 100/87 100/87
(%NBR)
Feedwater Temperature 420 360 420 360
(*F)
Peak Neutron Flux 688 636 688 636
(%NBR)
Peak Average Heat Flux 146 143 146 143
(%NBR)
Peak Vessel Pressure 1349 1297 1349 1297 (psig)
Peak Bulk Suppression 189.8 179.4 188.9 181.2 Pool Temperature (*F)
Peak Associrted 11.2 9.0 11.0 9.3 Containment Pressure (p'sig)
- Rated feedwater temperature condition from Reference 2.
- Partial feedwater heating.
2-10
NEDC-31323 g=e .=
E_ *
- m m;[ad '
aA:8 _=::
== s 6f I e
M nd i
,m,
- H.
n* >*EE c
W2g/ '
I E
_~J.
wa==2 el3r!
s a
W g
a n - W b.
/t y
N.
.y 3 C
>= b
! -n . / V1
-Y
'/i 5
l l u d
b o.
.Q E
c.
, ). C
. R 8.
_ 8 4 i s ist 11ta!1:r3W $
c C.
555 km. .
R*
(_y&:_ o fi
-u
= )-- E c W3;FC I #
- F '
- gd atvi: @E m
- - *i- rE -
E 5E-4==L=
I -$
%.' ; tu
- - ' O ss !
2p-;t -
ec
=m =,- t . g
[2:2s
-3 '
=33 w j ;e
= t;-" *
=s IE Uz z, E j l ] "&
-~-_C",w ;
- ~
_~-., . . . em
. *- " M W k '
w N
~
E E D o
, ~E j wz W .** .*_
- 2 J
, I 2
[]
e-l
- dI 0 ~
L .
\
\ .
4
_l /,P~ m f
e
'_ , / , . $
.f. . 9
=
.c .
.t.
.=
s_ s_ s. g s. =.
s_
103112i A U33f3d!
2-11
l lllIlll ,1I lIjlI s
b v 2'Ute
) .
I iP=
lS Si P
8M y -
- f 2 N*
8 f EI iI sri 2 fvI I
5 ilI liVV vClI SK R5m k
, . lAlT . w Ef n 0 IfEE 0 o PLa f 2 fttf 2 l
[
l i f N F LI ,f K(RH I e
ELp.
S SM' f E 0PtWW lr8 r
o WSi I o1N#
I2 H '
5 1
I vI I23 a
\
5.
1 C
"
- C - 5
, 0 M
I
' 1 M r I
- e 0
1
.T g
i w
o E P 5
I 5
. a -
S 4
0 1
t a
u R
n u.
d n
. .O
. ~ ~
a
- 0. 0 0' 0 I'
. .O m 0 0 0 0 ,' 2- e 2 1 1
D
-lgWW m u
m i
TtI xe RSS ar Pn l(El tet M u
. Mu SfCil ,ta
. E' - N Nt e T'
N PIIPi W
- er R' ELLlOIi re O"'
I up mX,'
Ea L - fIuL FVVnF l m K
U L
IgF N
(t M H
L 0
IEI[N RLI7A N11DTL 0 i e F
aT ft;I
' 2 aMJl1 2 r Nia,5fj L iNfNri5 l'
i re ffI5 s et N
t R 5 nfgE 1M S lSSC1 l a k~F.
l 5 f f l w lf Ef
- L EvRHW 1' t1 od Nt;V l i .
LHNT0 re t e I , '
t 5 I2145 r 5 nF 1I C - -
'"1 o 2 Cd 1
M e I
. rc
^
0 1
.T t
l i
3# 0 1 d eu t d ae wR ed 5 en z"
- F a s
2 5 5. 2 e
p5
- r 8
8 u g
- i F
L. 0 0
0.
s i
0 0.
1 0
5 D
E*
I ,
,' 0
!b !
vC '
ll l
I NfuTRON FLOW . I VESSEL PtfS RISE IP$ll 2 l'l fM f t E l CL Nif R IEMP J IM Ll4 PHE5 Al5E IPSil gg* _
3 hyl 5 4W 6 cfitRLflUI 200~ ' I I " ' d IIDI3 i 4 filflHilill fl 0H *
~l t' Ita l "l2I4flutBl i JH IH 5 VLSML 51t HM F LOH u f f vi hWE ILOHl I i u u Wi iHTL PCII
' .M -
} ',
3 100.
- - , 100, r, E ,
, 1 4 IM N '- ' 5 I 6 5 O'
l"g( l .b ih _
- 0. *- ' ** 100. - '- - *
- 0. 5. 10. '--15. 20. O. 5. 10. 15. 20.
flM 15Ecl flME 15ECI N
?
u I LEvEt Blu H-af f-5EP-suini i vnin nEw livlir N W
2HASfwa n LEVELeit(HESI 2 Oneetta tfaCllylt 3 N H Sf WJ n iIVf . SfHIM Hf t filvil I2- 4Tn M af TTINabflNCK$1 5 BIPRSS $1[FM FLOHiel PCll [g MIChDCUS '
12h 100.
- 0. ' 7" -- C- U ,'
.- -~ y u u - -
G
$0 -
' M ... .I' 8 i 4 225:
5 5
N 't E
) N ' '
- o. m s ; . 2. .t. . a .
O. 5. 10. 15. 20. U. S. 10. 15. 20.
IINE 15LCl llHE 15[Cl Figure 2-3 Feedwater Controller Failure, Maximum Demand, at 100% Power, 87% Core Flow and Reduced Feedwater Temperature
I NfulRdN FLUX l VfSSEL PFFS AISE IPSil 2 t fitt fit t [ENif R ifMP 2 SIM L If( t1f5 815[ 1r511 3 HVI si ;Uit ALflux 31 ' Int;# trsil ISO- (Tlinwn!d ril sflru 200- + a' 1 Nil igs ftt(6I_U/181 I $tEl 1 5 VE55LI. Sltiet FLIN If f V6 ft0Hl.I 1 4 4 Id%fif PCI) e 100.
D
, , 100. ..
-/
7 y u if
$0. - -b '
O. .UI* * '6 6 5 b . b' _
14
- 0. '- -
-100. '- - *
- 0. S. 10. 15, 20. D. 5. 10. 15. 20.
TlHE ISEcl fitt ISECs Y
- ?
w I LEVEt flWit-Aff-S[r-5stl AT 1 VOID MfAf IIVITT U N
2 H A Sf fW II L E Vl t e lNCit $1 2 Onittilt i FACllWli). W 3 N H Sit 61 18ffV(IIINfHESI g- -
A_ 1 SIIBM 15 8 filVll F IM- tit 5f M fTTIN4' IFill tV IDIHL hIIf)ITTII 5 BTIRSS 51tfM ILOHiel O, DM- /
loo.
/
I u u we#6A
- W
! -I' 1
x r *
% j -
- 0. 5'- - 5 l * -2. a ' - i
- 0. 5. 10. 15. 218 D. S. 10. 15. 20.
IINE ISECl TIME IS[Cl Figure 2-4 Feedwater Controller Faflure, Maximum Demand, at. 1001 Power, 87% Core Flow and Reduced Feedwater Temperature at EOCl-2000 Exposure
- 9?d eO L
1 1II l5IB iPDt I SII I P ul I.I i-(E S
fiM t I T2 ITi SRII EIEf4O I4 N rVIr I
RSl4 I0" ilMV li vCl L _
iRTi/
SfC$GF i
EPHI M s 0 2
fECt Wffi 0
2 e
F IiMf f ~
' f Fr l -
EHI t PK"MIAiS LI! REh L b
a ELI S
SN' I i 0PE 1P r e
EI V
/I 56 1
1 Q i 5-5 1
C l
0 VOc Ya J
9'
/ "l1 5
i p
o n
- E 8
S '
s I s a
E pe
- yr
.II - g B u 01 I t ea E
- nr ie 1
' b p rm ue TT 5 5.
~
hr t e it waw
" " - ,d M - * -
d e
- 1. ne
.O - . . .( aF 0* 0. 0 0 5 O 1 2 m 0 0 0 -
ed 2 1 1 D e gW E c mu ud me iR T8I R5S x I[E ad PX MHHi Mn MU EL
. SCCil N iNt e *
,a Tf 8 5 "
Pll EiIiH S L O ew ro Ri EE _
EA T
N tHF 1L fVVsL FEE F ul l F EEJ O U([ft te l
RLlaM
- f s .
i ae L C E 0 H001E 0 F r l1 2 fIf tI sflS 2 N lwN 5 o Ntt n
' IA9w5 ltf LS S- S rC e
E l %
lM 5_D S iAt[S l VRR r rft - l 7 FtvfE E aT o8 i
LNNrB r N P f_f V I23r5 .
I234S > I 5
1 s
5 1
t ,
nr l l h C oe I
aI i Y E
S C w k o 1
i u I rP M e
_ - 0 1
.T I
0 1
t%
d1 a0 w0 x~
s e
- u et F a u
~-
5
. - .5 5
2 e
i r s u u g a
.O
. . w. 0 i
F g 0 0 0 0. 0- 0 0 5 S-1 0
1 5
g 1
-l E E Ni0
- l1
1 FEUTRS4 FLUX l L PfES RISE IPSil
'EfM fift EENTER TEMP 2 :IM IE PftE5 RISE IP5Il El' l pyg
- C HEAT FLUX 200. q 150. k l y VESSEL 51EAM FLON g
" E s
100.
~
100. ,
a b
3 u u u {/ NU K
" f- "" ' "
- I*' B R O.
g 50. 3 s
E - 5 b -
'. I 1 uh -100. '- *'
O. -
- 5. 10. 15. 20.
- 5. 10. 15. 20. O.
-O.
itME ISECB TIE ISECl Y
Z ?
I LEVELilNCH-fief-SEP-5KIMT I EAC TIVITT "
- fVO!!i 00FF_ER FEncilVITT 2 H R SENSE 0 LEVELilNCHESI 3 N R SENSED LEVELfINCESI g*
H SCHA1 fiff CTJJiT V
150' II'ENETKE i FLIDI (PCil t 70TANiCTIVI 5 STPRSS SlEFH FLOHiel
\
-^ "-- I 100.
u u u O.
[ e
/
3
- 1.
50- -
-p
- 4 L__1_ -
fit - 5 5 2. t 4
- 0. - - -
20.
- 0. 5. 10. 15. 20. O. 5. 10. 15.
TIME 15EC3 llE 15ECl Figure 2-6. Feedwater Cont roller Failure. Maxirium Deamnd, with Turbine Bypass Tnoperable and ::PT Inoperable et 104.5% Power, 100 Core Flow and Reduced Feedwater Temperature
. I I weiinnu tita I wsst rvrs Alw ersil I 211let iIt: f.tNIEH TIMr 2 SIM LIM Nt3 RISE Ir$18 150-
' Ueve.sisti rf.!s tilflus 3ao- Un gsigtfyigfs(cits.:solutel peli sa 1 g istramille flow 1 5 vtSttL SIE8M iLfte 5 IUfB SitiM f tthe IPCII
')T kx, s 251. '
x la IM- c--
y g C- - - - - - - -
h
- g. DM -.
d 100. 5 ___
h u
\
4 -
/,~
M e
- 0.
- J I *
- 0. 3 a- ^
- 0. 2. 4. 6. 8. D. 2. 4 6. 8.
TIM ISEcl TIME SECB e.
W I LEvft alwH-MF-WP.SalRi volo MN fivlit U 2 H A 5f erd il Lt Vf t t 1NI et st OnPI'L t R ttACIlvlif N 2m. 1 N H 5t il 1 0 tivittleswSI 3 feel Mi ditt W frN CIMi T'rtonl'IPCIe 5 DHivE fia W I IPCil 3 **
100- N 0 ^ '
3 8 7
, m
- h. ,w 3 mv _ N- -m- *__:c = '. 2 ~
=
D. --
E .l.
.t y -
d 300. . ....... . 2. su.n.... .
D. 2. ti . t>. ft. U. I. 2. 3. 4.
TIM iStC ilE ISECl 4
Figure 2-7 MSIV Closure. Flux Scram, at 104.3% Power. 105% Core Flow and Normal Feedwater Temperature
- 3. MECHANICAL EVALUATION OF REACTOR INTERNALS AND FUEL ASSEMBLY 3.1 LOADS EVALUATICN Evaluations were performed to determine bounding acoustic and flow-induced loads, reactor internal pressure difference loads and fuei-support loads for ICF and/or PFH operation.
Acoustic loads are lateral loads on the vessel internals that result from propagation of the decompression wave created by a sudden recirculation suction line break. The acoustic loading on vessel internals is proportional to the total pressure wave amplitude in the vessel recirculation outlet nozzle. The total pressure amplitude is the sum of the initial pressure subcooling plus the experimentally determined pressure undershoot below saturation pressure. PFli operation increases the expected acoustic loads because this downcomer subcooling increases and, therefore, the total pressure wave amplitude increases. The high velocity fl'ow patterns in the downcomer resulting from a recirculation suction line break also create lateral loads on the reactor vessel internals. These loads are prbportional to the square of the critical mass flow rate out of the break. The additional subcooling in the downcomer resulting from PFH operation leads to an increase in the critical flow and, therefore, to a corresponding increase in the flow-induced loads. The reactor internals most impacted by acoustic and flow-induced loads are the shroud, shroud support and jet pumps.
A reactor internals pressure difference analysis was performed for the ICF region. The increased reactor internal pressure differences across the reactor internals were generated for the maximum core flow at normal, upset, and faulted conditions for the reactor internal impact evaluation.
Fuel-support loads and fuel bundle lif t for LGS Unit I were evaluated based on results from probabilistic fuel lift analyses performed at 105% of rated core flow following the procedure of Reference 6. Envelopes of 100-nil and 80-mil channel fuel-support loads and ' fuel bundle lift were determined for upset, faulted and fatigue load combinations. It was shown that the fuel
! 3-1
NEDC-31323 bundle lift is a small fraction of the applicable design criteria (established in the NRC Safety Evaluation Report to Reference 6) for the faulted event.
3.2 LOADS IMPACT 3.2.1 Reactor Internals The reactor internals most affected by ICF and/or PFH operation are the I
core plate, shroud support, shroud, top guide, shroud head, steam dryer, control rod guide tube, control rod drive housing and jet pump. These and other components were evaluated using the bounding loads, discussed in Section 3.1, under normal, upset, and faulted conditions. It is concluded that the stresses produced in these and other components are within the allowable design limits given in the Final Safety Analysis Report (Chapters 3 and 4) or the ASME Code,Section III, Subsection NG.
3.2.2 Fuel Assemblies .
The fuel assemblies, including fuel bundles and channels, were evaluated for increased core flow operation considering the effects of loads discussed in Section 3.1 under normal, upset, faulted and fatigue load combinations.
Results of the evaluation demonstrate that the fuel assemblies are adequate to withstand ICF effects to 105% rated flow.
The fuel channels were also evaluated under normal, upset, and faulted conditions for increased core flow. The channel wall pressure differentials were found to be within the allowable design values (Reference 7).
2 3-2
. _ , _ . _ _ . - - _ . _ _ _ _ _ _ - _ . - ~-
- 4. FLOW-INDUCED VIBRATION 4
To ensure that the flow-induced vibration response of the reactor inter-nals is acceptable, a single reactor of each product line and size undergoes an extensive vibration test during initial plant startup. After analyzing the results of such tests and assuring that all responses fall within acceptable limits of the established criteria, the reactor is classified as a valid prototype in accordance with Regulatory Guide 1.20. All other reactors of the same product line and size undergo a less rigorous confirmatory test to assure similarity to the base test. The acceptance criteria used for vibration assessment is based on a maximum allowable alternating stress intensity of 10,000 psi.
The increased core flow vibration analysis was performed by analyzing the startup test vibration data for the valid prototype plant (Browns Ferry 1).
Based on the results of the analysis and a review of the test data, the reactor internals response to flow-induced vibration is expected to be within acceptable limits for plant operation in the ICF region (region bounded as shown on the power flow map, Figure 1-1).
4-1
- 5. FEEDWATER N0ZZLE AND FEEDWATER SPARGER FATIGUE USAGE 5.1 METHOD AND ASSUMPTIONS The fatigue experienced by the feedwater nozzle and feedwater sparger results from two phenomena: system cycling and rapid cycling. System cycling is caused by major temperature changes associated with system transients.
These transients are identified on thermal cycle diagrams. Thermal stresses due to these transients are calculated by determining inner and outer metal surface temperatures using conventional heat transfer and stress analysis methods. Fatigue usage is determined by dividing the number of design cycles for each transient by the number of allowable cycles for each stress calcu-lated. Cumulative system fatigue usage is determined by summing all of the respective transient fatigue usage factors.
Rapid cycling is caused by small, high frequency temperature fluctuations caused by mixing of relatively colder nozzle annulus water with the reactor
, coolant. The colder water impinging the nozzle bore originates from leakage past the thermal sleeve secondary seal and from the boundary layer of colder water formed by heat transfer through the thermal sleeve. The mixing region extends from the feedwater nozzle surface region to the feedwater sparger surface; therefore, rapid cycling applies to both of these components. Once thermal stress due to rapid cycling is determined, fatigue usage is calculated and the results are added to the cumulative system cycling usage factor to obtain the total usage factor.
The introduction of PFH will cause a change in calculated rapid cycling fatigue only. This is because the system transient is very mild (small temperature change and relatively long duration) and is bounded by the original design basis thermal stress analysis. General Electric has developed standardized rapid cycling duty maps for each BWR plant that cover the design basis rapid cycles in the same manner that thermal cycle diagrams cover the design basis thermal transients (system cycling). The methodology used to develop the duty maps is based on the results of extensive testing of feed-water nozzles by General Electric. PFH is analyzed by modifying the design cycles in order to gauge its effect on fatigue usage. The reduced feedwater 5-1
NEDC-31323 temperature will tend to increase fatigue usage due to an increase in thermal stress, while the reduced leakage flow (due to a decrease in feedwater flow) during coastdcun will tend to decrease fatigue usage because of reduced leakage to the sparger and' nozzle surfaces.
An evaluation of the effect of FFWTR and FWHOS on the feedwater nozzle and feedwater sparger fatigue was performed for the following conditions:
- a. FFWTR Case: As the last step in an 18-month fuel cycle, FFWTR to a feedwater temperature of 360*F (60*F reduction) for 18 days was followed by a 3% per week coastdown over ten weeks to a final power of 70%. The coastdown was initiated from a reduced feedwater temperature of 360*F. The feedwater temperature at the end of the coastdown was 330.9'F.
- b. FWHOS Case: A 60*F drop in feedwater temperature was assumed for various lengths of time during an 18-month fuel cycle, so that a relationship could be determined for incremental fatigue damage as a function of time spent with FWHOS.
The analysis was performed by simulating the feedwater temperature reduction during the coastdown period in four equal increments. The temperature and flow rates are set for each step to provide conservative results.
5.2 FEEDWATER N0ZZLE FATIGUE The analysis done for normal design duty indicated that refurbishment of the thermal sleeve seals after 32 years would be necessary to keep the 40-year total fatigue usage (system cycling plus rapid cycling) below a value of 1.0.
For a refurbishment schedule of 32 years, the 40-year total fatigue usage was calculated as shown in Table 5-1. The fatigue damage per cycle for FFWTR operation is conservatively estimated by taking the difference between the FFWTR fatigue and the normal operation fatigue and dividing that quantity by the number of cycles in 40 years.
5-2
O l
NEDC-31323 If FFWTR and coastdown were used for every cycle, fatigue usage less than 1.0 can be achieved by reducing the refurbishment interval to 28 years, as noted in Table 5-1. Although the refurbishment interval is impacted by four years assuming FFWTR after every cycle, only one refurbishment is required, as is the case for normal operation.
The same approach was taken for FWHOS operation and the results are as shown in Table 5-2 tor both of the cases analyzed. The' impact of seal returbishment for this mode of operation for the times assumed is less severe than for FFWTR.
In conclusion, the implementation of FFWTR and FWHOS at Limerick Unit I will have a minor impact on the feedwater nozzle. The incremental fatigue usage due to FFWTR is 0.027 per cycle and incremental usage due to FWHOS for 12% per year is 0.013 per cycle. These are very conservative numbers since the analysis assumes both a long coastdown for every cycle and conservative seal leakage.
5.3 FEEDVATER SPARGER FATIGUE Since the feedwater sparger is not an ASME Boiler Pressure Vessel Code component, a fatigue analysis was not originally performed. To gauge the effect of the FFWTR and FWHOS on the sparger, the fatigue usage in the sparger for the original thermal duty cycles and rapid cycle duty map was calculated.
The fatigue usage for system cycling is 0.26 for the 40-year life of the plant. The calculated fatigue usage for rapid cycling is 0.45, which makes a total 40-year usage factor of 0.71. This result shows that the usage factor will not exceed 1.0 for the 40-year life of the plant.
The results for FFWTR and FWHOS operation are summarized in Tables 5-3 and 5-4, respectively, along with the above results for normal operation. The results show that implementation of FFWTR and FWHOS at Limerick Unit I will have a minor impact on the feedwater sparger. The fatigue results are based on the assumed conservative leakage rates and operating modes described l
l l
l 5-3 l
NEDC-31323 by the design basis duty map. The actual impact of FFWTR and FWHOS operation on sparger fatigue usage will depend upon actual leakage rates and operating modes. For example, if a shorter coastdown is used, the contribution of FFWTR to total fatigue would decrease significantly.
5-4
r NEDC-31323 Table 5-1 l FEEDWATER N0ZZLE FATIGUE USAGE FOR A 32-YEAR SEAL REFURBISHMENT PERIOD - FFVIR Normal Operation FFWTR 40-Year Total Fatigue Usage 0.9815 1.6996*
Additional Usage Due to FFWTR -
0.7181 Additional Usage per cycle Due to FFWTR - --
0.0270
- The total 40-year usage fac u r for FFWTR operation after every cycle can be kept to below 1.0 by refurbishing the seals after 28 years.
e -
F O
5-3
,-- , --- - - - , - - - - - - , - . , , - , , - , , , , , - - -- -- -- -- - ,..n- ,
NEDC-31323 Table 5-2 FEEDWATER N0ZZLE FATIGUE USAGE FOR A 32-YEAR SEAL REFURBISHMENT PERIOD - FWHOS 18-Month Cycle 18-Month Cycle FWHOS Operation FWHOS Operation for 657 Hours for 1577 Hours Normal Each Cycle Each Cycle Operation (5.0% per year) (12.0% per year) 40-Year Total Fatigue Usage 0.9815 1.1286* 1.3346*
Additional Usage Due to FWHOS --
0.1471 0.3531 Additional Usage ,
per cycle Due to FWHOS --
0.0055 0.0132
- The total 40-year usage factor for FWHOS operation after every cycle can be kept to below 1.0 by refurbishing the seals after 31 years for the 657 '
hours / cycle case, and bj refurbishing the seals after 30 years for the 1577 hours0.0183 days <br />0.438 hours <br />0.00261 weeks <br />6.000485e-4 months <br /> / cycle case.
9 l
i 5-6
4, ; '
i
- i
\
NEDC-31323 Table 5-3 FEEDWATER SPARGER FATIGUE USAGE FOR A 32-YEAR 3EAL REFURBISHMENT PERIOD - FFVIR
\
- h Normal l Operation FFWTR g1 40-Year Total
,jg Fatigue Usage 0.7125 1.3897*
%3 Additional Usage s i Due to FFWTR -
0.6773 f ,
N g
t
,s Additional Usage
' ( ,
per cycle Due to FFWTR --
0.0254
..) ,' !
4- -
, # .3 s
- The total 40-year usage factor for FFWTR operation after every cycle can be
, kept to below 1.0 by refurbishing the seals after 28 years (same as nozzle).
4 f
l l*
l-l r
9 e
s 5-7 s s i
\
-,a__L.- --, - -
- ~ "
NEDC-31323 . l I
, Table 5-4 FEEDWATER SPARGER FATIGUE USAGE FOR A 32-YEAR SEAL REFURBISHMENT PERIOD - FWHOS 18-Month Cycle 18-Month Cycle FWHOS Operation FWHOS Operation for 657 Hours for 1577 Hours Normal Each Cycle Each Cycle Operation (5.0% per year) (12.0% per year) 40-Year Total Fatigue Usage 0.7125 0.8449 1.0303 Additional Usage Due to FWHOS --
0.1324 0.3178 Additional Usage per Cycle Due to FWHOS -
0.0050 0.0119 o
O e
5-8
- 6. CONTAINMENT ANALYSIS The impact of feedwater temperature reduction and increased core flow operation on the containment LOCA response was evaluated.
The results show that the containment LOCA response for ICF operation alone is bounded by the corresponding FSAR results (Reference 1). The calculated peak values for drywell pressure suppression chamber internal pressure under ICF and/or PFH are bounded by the corresponding values for the FSAR (Chapter 6) conditions. The peak value for drywell deck downward differential pressure is bounded by the appropriate design limit. All other containment parameters (e.g., drywell and suppression chamber temperatures, drywell and suppression, chamber external pressures, drywell and suppression chamber maximum allowable leak rates) are bounded by the results reported in the FSAR.
The LOCA-related pool swell, condensation oscillation and chugging loads were evaluated at the worst power / flow conditions during ICF/PFH operation.
The conditions which establish pool swell loads under ICF/PFH are bounded by the corresponding FSAR (Chapt.er 6) conditions. Pool swell, condensation oscillation and chugging loads with ICF/PFH conditions are bounded by the appropriate design loads.
6-1 I
- 7. OPERATING LIMITATIONS Restrictions / limitations which are unique to ICF/PFH operation are identified below.
7.1 FEEDWATER HEATERS The PFH analyses have assumed that certain stages or strings of feedwater heaters are valved out-of-service (PFH 60*F at rated power). This may be done at any time during the cycle and/or after EOC1 whether or not ICF is used. Operation after EOC1 is done to help increase power after all control rods have been withdrawn at EOC1, and was accounted for in the safety analyses in Section 2.
7.2 OPERATING MAP The allowable operating domain of the normal power-flow map has been
, increased to allow operation at 100% power up to 105%. core flow. The minimum allowable power in this ICF region is bounded by the jet pump cavitation protection interlock, as shown in Figure 1-1. ICF reactor internal pressure differences and fuel bundle lift calculations were analyzed and are applicable for reactor operation within the power flow map in Figure 1-1.
7.3 MCPR OPERATING LIMITS l
Required MCPR operating limits applicable to ICF/PFH have been determined for LGS Unit 1 as given in Table 2-2.
l l
7.4 K FACTOR g
j For core flows greater than or equal to rated core flow, the Kg factor is equal to 1.0.
l
! 7-1 i
n -
- 8. REFERENCES
- 1. " Final Safety Analysis Report, Limerick Generating Station", as amended throegh Revision 45, December 1985.
- 2. " General Electric Boiling Water Reactor Extended Load Line Limit Analysis for Limerick Generating Station Unit 1, Cycle 1", April 1986 (NEDC-31139).
- 3. " General Electric Standard Application for Reactor Fuel (Supplement for United States)", May 1986 (NEDE-24011-P-A-8-US, as amended).
- 4. G.A. Watford, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria", October 1984 (NEDE-22277-P-1).
- 5. "BWR Core Thermal Hydraulic Stability", SIL No. 380 Revision 1, February 10, 1984.
Licensing Topical Report, Amendment No. 3, October 1984 (NEDE-21175-3-P-A and NEDO-21175-3-A).
- 7. "BWR Fuel Channel Mechanical Design and Deflection", General Electric Company, September 1976 (NEDE-21354-P) .
8-1
- - - - _ _ _ _ - _ __,_- _ , - - _ _ . - _ - - _ . , . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - ~
e NEDC-31323 DISTRIBUTION R.J. Brandon 779 C.S. Chen 147 S.S. Dua 769 R.S. Drury 763 D.W. Diefenderfer PRO J.M. Grau (2) 763 D.A. Hamon 763 G.R. Hull 174 G.V. Kumar 156 L.K. Liu 770 J.R. Palletta 763 H.J. Pearson 171
- 0. Raza 171 D.J. Robare , 682 f A.E. Rogers 763 1
L.F. Rubino (30) PECo G.J. Scatena 171 l
G.L. Sozzi 747 G.L. Stevens 769 C.A. Vondamm 682 J. Wallach 156 S. Wolf (2) 763 NEB 0 Library (2) 528 6