ML19344D216
| ML19344D216 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/31/1980 |
| From: | Jensen S, Kayser W, Valentine P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19344D209 | List: |
| References | |
| TASK-04-01.A, TASK-4-1.A, TASK-RR XN-NF-79-078, XN-NF-79-78, NUDOCS 8003110551 | |
| Download: ML19344D216 (44) | |
Text
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<4 XN NF 79 78 F
r LOCA-ECCS ANALYSES SUPPORTING r
FOUR-PUMP OPERATION OF r
OYSTER CREEK REACTOR r
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i JANUARY 1980 1
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RICHLAND, WA 99352 I
I ERON NUCLEAR COMPANY,Inc.
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KN-NF-79-78 ISSUE DATE: 01/31/80 l
l LOCA-ECCS ANALYSES SUPPORTING FOUR-PUMP OPERATION OF 0YSTER CREEK REACTOR f
Prepared by:
SE Jensen PJ Valentine WV Kayser GC,Cooke f k.b bh7lN-Concur:
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JN'Morgsn, Manager Date Licensing & Safety Engineering Concur,:WlE<)
/A M
' Busselman,Wanager
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.tronics & Fuel Management Concur:
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/- 2 7 - Ett GA Sofer, Ma ga f'
Date i
Nuclear Fuels Tngineering Concur:
/
.2E-M LJ FEdEr ko, Manage 7 Date Nuclear Fuels Projects Approved:
[A
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WS Nechodom, Manager Date Licensing & Compliance SEJ:vb ERON NUCLEAR COMPANY,Inc.
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IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY r
o This technical report was derived througn research and development programs sponsored by Exxon Nuclear Company. Inc. It is being submitted by Exxon Nuclear to the USNRC as part of a technical contribution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear-fabricated reload fuel or other technical services provided by Exxon Nuclear for light water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.
Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:
A.
Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus. method, or process disclosed in this document will not infringe privately owned rights; or
- 8. Assumes any liabilities with respect to the use of, or fbr damages resulting from the use of, any information apparatus, method, or process disclosed in this document.
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1 1
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i XP-NF-79-78 TABLE OF CONTENTS Section Page
1.0 INTRODUCTION
AND
SUMMARY
I 2.0 COMPARISON OF FOUR-LOOP AND FIVE-LOOP ANALYSES........
7 3.0 IDENTIFICATION OF FOUR-LOOP GPERATION LIMITING BREAK....
22 3.1 INACTIVE LOOP BREAKS.................
22 l
3.2 SINGLE FAILURE STUJIES................
24 4.0 MAPLHGR LIMITS FROM ECCS ANALYSES.............
30
- 5. 0 REFERENCES.........................
37 l
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ii XN-NF-79-78
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LIST OF TABLES Table Pg 1.1 Oyster Creek Four-Loop Operation MAPLHGR Limits ENC 8x8 Fuel........................ 4 2.1 Blowdown Event Times for Four-Loop and Five-Loop Operation
........................10 3.1 Inactive Loop Break Results...............
25 3.2 Single Failure Study Results...............
26 4.1 Existing Four-Loop Operation MAPLHGR Limits 33 i
4.2 Oyster Creek Five-Loop Operation MAPLHGR Limits 34
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iii XN-NF-79-78 LIST OF FIGURES Figure page 1.1 Oyster Creek Four-Loop Operation MAPLHGR Limits, ENC 8x8 Fuel........................
5 1.2 Axi al MApLHGR Mul ti pl i e r..................
6 2.1 Oyster Creek 4-Loop Nodal.ization for Limiting Break Case.
11 2.2 Core Pressure, Limiting Break Fou r-Loo p Ope ra ti on....................
12 2.3 Hot Node Quality, Limiting Break Fou r-Loo p Ope ra ti on....................
13 2.4 Hot Assembly Inlet Flow, Limiting Break Four-Loop Opera tion....................
14 e
2.5 Hot Assembly Outlet Flow,. Limiting Break Four-Loop Opera ti on....................
15 2.6 Pump Side Break Flow, Limiting Break Four-Loop Operati on....................
16 e
2.7 Vessel Side Break Flow, Limiting Break Four-Loop Operation....................
17 2.8 Hot Node Heat Transfer Coefficient, Limiting Break Four-Loop Opera tion....................
18 2.9 Core Spray Flow, Limiting Break Fou r-Loop Opera ti on....................
19 2.10 Normalized Power, Limiting Break j
(s Fou r-Loop Ope ra ti on....................
20 2.11 Hot Assembly Inlet Enthalpy, Limiting Break Four-Loop Opera tion....................
21 3.1 Inactive Loop and Bypass Line Configuration........
27 3.2 Calculated Core Inlet Flow Inactive Loop Break Between Pump and Discharge Valve.................
28
iv XN-NF-79-78 LIST OF FIGURES (Conttnued)
Figure Page 3.3 Calculated Core Inlet Flow Inactive Loop Break Between Discharge Valve and Reactor Vessel........
29 4.1 Comparison of Four and Five-Loop MAPLHGR Limits for ENC 8x8 Fuel Type V....................
35 4.2 Comparison of Four and Five-Leop MAPLHGR Limits for ENC 8x8 Fuel Type VB...................
36 i
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XN-NF-79-78
1.0 INTRODUCTION
AND
SUMMARY
This report presents the results of LOCA-ECCS analyses considering the possible combinations for four recirculation loop operation of the Oyster Creek Plant.
The report provides the limiting break for LOCA-ECCS analysis with four-loop operation, and gives the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for the appropriate fuel types based on this limiting break. The four-loop operation MAPLHGR's are given in Table 1.1 and Figure 1.1.
This report also extends MAPLHGR limits for four-loop and five-loop operation out to exposures of 40,000 MWD /MTM. This has been done I
in support of high burnup demonstration programs involving additional irradia-j tion cycles for ENC 8x8 fuel.
The present analysis has been performed with the Exxon Nuclear Company (ENC)
WREM-Based NJp-BWR ECCS Evaluation Model(I) unchanged from previous analyses.
System input to the model was revised to reflect the four-loop operation con-figuration and conditions. The ECCS Evaluation Model was developed and approved as an integrated model for application to nonjet pump boiling water reactors (NJP-BWR's) such as Oyster Creek.
In 1977 this model was used to perform LOCA-ECCS analyses for the Oyster Creek Reactor assuming operation of all five operating loops, and the MAPLHGR limits in confonnance with 10 CFR 50.46 and Appendix K were established.(2) A major feature of the ENC WREM-based NJP-BWR Evaluation is the use of core heat transfer based on calculated i
core flow during blowdown instead of the more conservative "dryout correlation" approach where core heat transfer is based on initial liquid inventory.
l
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. XN-NF-79-78 The principal results of the present ECCS analysis for four-loop operation are:
e For active loop breaks with an emergency condenser connected to an L
active intact loop, the calculated system response is essentially that obtained for the previous five-loop analyses. Therefore,.the previous studies for break size, configuration, break location, and single failure done for five-loop operation remain valid for four-loop operation in this configuration.
e Other possible four-loop break configurations include inactive
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loop breaks and emergency condenser injection into an inactive loop. Analyses were performad for these possibilities, and the limiting break for four-loop operation was dettnnined to be the double-ended guillotine break in an active reci. culation loop 1
with 0.4 discharge coefficient and located between the reactor vessel and the venturi.
The worst single failure remains the failure of one emergency e
condenser valve to open, such that, the operable condenser discharges into an intact active loop.
i e
Peak cladding temperature (PCT) for four-loop operation exceeds
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that computed for the comparable five-loop operation case by 70*F due to a loss of pumping capability associated with four-loop operation.
The reduction in MAPLHGR limits from five-loop values varies with e
fuel exposure with the maximum impact being a reduction of 4.6%
for ENC 8x8 VB fuel at beginning-of-life.
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- XN-NF-79-78 e
The MAPLHGR limits supporting four-loop operation at Oyster Creek -
have been determined and are given in Table 1.1 and Figure 1.1.
In summa:7, operation of Oyster Creek within the limits defined by Table 1.1 and Figure 1.1, and the axial multiplier established previously(2)
(Figure 1.2), is in conformance to the NRC 10 CFR 50.46 criteria and Appendix K for all identified possible configurations of four-loop operation at the assumed full power and core flow conditions. MAPLHGR limits for four-loop operation for resident 7x7 fuel types are conservatively represented by existing heatup analysis values as given in Table 4.1.
The latest five-loop I
MAPLHGR limits are repeated and extended in Table 4.2.
i s
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.. XN-NF-79-78
. Table 1.1 Oyster Creek Four-Loop Operation MAPLHGR Limits ENC 8X8 Fuel J
t Type V-
.Burnup MAPLHGR MWR PCT F xF (GWD/MTM)
(kW/ft) a r
(%)
(*F) 10.9 10.35 2.234 16.0 2134 15.1 10.05 2.170 15.0 2124 23.5 9.91 2.140 15.0 2068 32.4 9.74 2.103 14.0 2038 l
40.0 9.30 2.007 13.5 2035 Type VB Burnup MAPLHGR MWR PCT F_xF (GWD/MTM)
(kW/ft) a p
(g)
(op) 0.0 9.88 2.133 7.5 2200 O.5 10.65 2.'00 8.2 2200 1.2 10.80 2.331 7.5 2157 1.9 10.80 2.331 7.8 2176 7.0 10.62 2.294 14.0 2200 11.1 10.55 2.278 15.0 2180 15.4 10.30 2.223 14.6 2092 23.7 10.03 2.166 15.0 2084 32.8 9.89 2.135 14.9 2061 40.0 9.43 2.035 14.5 2066 1
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, XN-NF-79-78 2.0 COMPARIS0N OF FOUR-AND FIVE-LOOP LOCA ANALYSES For the Oyster Creek Reactor, operation of four coolant pumps is suf-6 ficientEto provide 100% of the rated recirculation flow (61 x 10 lb/hr)and the 70% nominal core flow was conservatively used as in the five-loop LOCA-I ECCS analyses. Thus, the four-loop operation LOCA-ECCS analysis was performed for full power conditions with the initial core flow and fluid conditions assumed in the five-loop analyses.(2) The input speed of the four-active pumps was increased to provide the required flow, and a single inactive loop t
was modeled containing the bypass line resistance.
l The four-loop system nodalization is shown in Figure 2.1.
For the Oyster Creek Reactor design, a recirculation loop represents but a small part of the
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total primary system volume (2.2%), and would not be expected to influence greatly the LOCA behavior for breaks in other recirculation loops.
The first LOCA-ECCS analysis performed for four-loop operation was a recalculation of the previously identified limiting break (the double-ended guillotine with a discharge coefficient of 0.4, between the reactor vessel and the venturi with one emergency condenser operating on an intact loop). The other emergency condenser was assumed inoperable due to single failure of the t
l valve to open. The four-loop calculation assumed an active loop break and t
modeled four-loop operation with the intact inactive loop represented f
explicitly. The inactive loop has a small reverse flow (300 gpm) through a 2-inch bypass line around the pump discharge valve which is closed during four-loop operation.
These changes are required to represent four-loop
. XN-NF-79-78 4
operation, no other changes from previous analyses were made to either the Oyster Creek system representation or the ENC NJP-BWR ECCS Models.
Results of the four-loop calculation as given in Figures 2.2 through i
2.11 were compared with the results of the previous five-loop analysis.(2)
Event times, calculated core flow behavior, and fluid conditions as functions of time were determined to be basically similar to the values obtained from the previous five-loop analysis. Event times for both five-loop and four-loop cases are shown in Table 2.1.
Since r.) fundamental changes in behavior 3
has been observed from calculational results oetween five-loop operation and four-loop operation, the sensitivity studies performed to establish the limiting break for five-loop operation also apply to four-loop operation, for the configuration analyzed.
The limiting break size and configuration for active loop breaks for four-pump operation will be the break identified for five-pump operation (0.4 DEG between the reactor vessel and the venturi) with one emergency condenser injecting into an intact active loop. The final determination of the limiting break for four-loop operation then need only address break locations and single failure assumptions not considered in the
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previous five-loop analysis, such as, inactive loop breaks and emergency condenser injection into the inactive loop.
Results of these analyses are contained in Section 3.0 of this report.
t The peak cladding temperature (PCT) resulting from the above described four-loop configuration was found to exceed the five-loop calculated PCT by i
approximately 70*F even though the calculated fluid behavior results were
. X N-N F-79-78 essentially the same.
The cause of this adverse effect was studied in some detail. ' The increase in PCT results from a slightly earlier critical heat 1
flux due to a slightly earlier calculated core flow reversal.
This early flow behavior is attributed directly to four pump versus five pump effects.
In the analysis above, the initially operating pumps provide the assumed 70%
initial core flow in both five-loop and four-loop cases; however, the pump in the broken loop is effectively lost at the time of the break.
For five-loop operation, loss of the broken loop pump represents a 20% loss in pumping capabili ty.
For four-loop operation, loss of the broken loop pump is a 25%
decrease in pump capacity.
Since the pumps oppose the early flow reversal, the greater loss of pumping capacity for four-loop operation results in the i
earlier flow reversal and CHF.
Since this effect is adverse and the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits are computed to give PCT's only slightly below the 2200*F limit, a reduction in MAPLHGR limits is required to conform to NRC 10 CFR 50.46 criteria.
The revised limits are discussed in Section 4.0.
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,. XN-NF-79-78 Table 2.1 Blowdown Event Times for Four-Loop and Five-Loop Operation 5
I (0.4 DEG Break)
I Evant Time (SEC)
Event 5-Loop Operation 4-Loop Operation Break Initiation 0.0 0.0 Reactor Scram No Scram No Scram I
Low-Low Water Level 4.1 4.1 j
Low-Low-Low Water Level 6.2 about 6 i
Emergency Condenser On 14.1 14.1 A
[
Spray Starts 36.5 36.1 Rated Spray 52.3 51.9 t
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-Il-XN-NF-79-78 28 EMERGENCY CONDENSER ADS STEAM RELIEF
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PUMP 11 33 PUMP 12 BROKEN LOOP INTACT ACTIVE LOOPS Figure 2.1 Oyster Creek 4-Loop Nodalization For Limiting Break Case l
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, XN-NF-79-78 3.0 IDENTIFICATION OF FOUR-LOOP OPERATION LIMITING BREAK As noted in Section 2.0, ECCS analysis results do not change signficantly from five-loop operation to four-loop operation when breaks and emergency I
condenser injection are modeled to occur in active loops.
For five-loop operation, all loops are active.
For four-loop operation, this situation encompasses all,90ssibilities except inactive loop breaks, and single failure combinations in which the emergency condenser injects into an inactive loop.
The similarity of the calculated results between four-loop and five-loop operation assures that the identified limiting break location, break size, t
break configuration, and the worst single failure determined for five-loop operation will also be limiting for corresponding four-loop break cases involving this configuration.
Therefore, determination of the limiting break for four-loop operation need only consider the additional possibilities introduced by the inactive loop. The analyses of inactive loop breaks is i
presented in Section 3.1, and inactive loop emergency condenser injection and single failure combinations are discussed in Section 3.2.
3.1 INACTIVE LOOP BREAKS The inactive recirculation loop canponents and configuration is important in defining possible break locations in the inactive loop.
The inactive loop pump is turned off, the discharge valve is closed, the suction valve is open and the two-inch bypass line around the pump discharge valve is open. A small -everse flow (300 gpm) exists through the inactive loop and the bypass line.
The configuration is shown in Figure 3.1.
j
. XN-NF-79-78 The bypass line represents a very restricted flow area in the in-active loop, which is not present in the active loops.
Because of the restricted flow area, choking is expected to occur in the bypass line which t
will limit the flow at this point and thus, will limit the flow to any break downstream of this point.
For all possible large inactive loop recirculation i
line breaks, the bypass line will limit the break flow from one direction or the other..Therefore, two principal break locations exist in an inactive loop.
They are breaks on either side of the bypass line resistance. LOCA-ECCS calculations were performed for both of these locations.
Results for the break between tr.a inactive pump and the discharge valve confirmed the expected choking behavior.
The calculated core flow for this break is predominantly upward and of increased magnitude compared to the active loop break case.
The calculated core inlet flow for this case is shown in Figure 3.2.
The second break calculation was assumed to be on the opposite side of the bypass line resistance between the venturi and the reactor vessel.
The results for this break also show choking in the bypass line and a pre-dominantly downward core flow which is again higher in magnitude than that calculated for the limiting active loop break. Calculated core inlet flow
)
for this case is shown in Figure 3.3.
Table 3.1 shows the change in calculated PCT from the limiting active loop break case. The conclusion is that inactive loop breaks result in enhanced core flow compared to the limiting active loop break and thus i
are not limiting for four-loop operation of the Oyster Creek Reactor.
l
. XN-NF-79-78 3.2 SINGLE FAILURE STUDIES From the previous five-loop analyses the worst single failure for the Oyster Creek Reactor was identified as failure of an emergency condenser valve to open such that the remaining one emergency condenser operates and injects into an active intact loop.(1,2) Again, due to the similarity between five-loop and four-loop analysis results, the five-loop sensitivity studies remain valid for four-loop operation for configuration where the inactive loop is not involved.
Single failure possibilities where the inactive loop is involved include:
(1) failure such that the available emergency condenser discharges into the inactive loop; and (2) a different failure such that both emergency condensers operate, one connected to the inactive loop, and the other connected to an intact active loop.
ECCS analyses were performed for the identified worst active loop break considering these two single failure possibilities.
Calculated PCT's for the two cases compared to limiting active loop break case are given in Table 3.2.
The conclusion is that the worst single failure remains failure of one emergency condenser valve to open such that the remaining emergency condenser injects into an intact active loop.
Thus, the identified limiting break for four-loop operation of the Oyster Creek Reactor is the 0.4 double-ended guillotine break in an active recirculation loop between the reactor vessel and the venturi.
The associated worst single failure is failure of an emergency condenser valve to open such that the operable emergency condenser injects into an intact active recircula-tion loop.
, XN-NF-79-78 i
Table 3.1 Inactive Loop Break Results Break a PCT 'F (1) Limiting Active Loop Break 0
i (2)
Inactive Loop Break Pump Side of Bypass Line Between Pump and Discharge Valve
-292 (3)
Inactive Loop Break Venturi Side of Bypass Line Between Venturi r
and Reactor Vessel
-176 4
e i
f e
i e
2 e
4
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. XN-NF-79-78 Table 3.2 Single Failure Study Results Single Failure Case A PCT *F (1) Limiting Active Loop Break 0
(One Emergency Condenser to Intact Active Loop)
(2) Limiting Active Loop Break
-141 (One Emergency Condenser to Inactive Loop)
[
(3) Limiting Active Loop Break
-104 (Two Emergency Condensers, One to Inactive Loop, One to Active Loop)
I s
a
,e
Open 2-inch Bypass Bypass Line AA m
U a300 gpm 1
Closed Discharge Valve Pump i
I 5
Figure 3.1 Inactive Loop and Bypass Line Configuration y
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o to 40 G0 30 100 2G 143 ao 30 200 TIME Figure 3.2 Calculated Core Inlet Flow Inactive Loop Break
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TIME Figure 3.3 Calculated Core Inlet Flow: Inactive Loop Break Between Discharge Valve and Reactor Vessel
. XN-NF-79-78 4.0 MAPLHGR LIMITS FROM ECCS ANALYSES The MAPLHGR limits obtained with the ENC WREM-Based NJP-BWR ECCS Evalua-tion Model for four-loop operation have been provided in Table 1.1 and Figure 1.1.
These results are for ENC 8x8 Type V and Type VB fuels. New four-loop operation MAPLHGR limits were not computed for ENC 7x7 fuel at Oyster Creek since the current four-loop operation limits for this fuel type provide satisfactory operating limits.
The existing four-loop operation MAPLHGR limits for ENC fuels at Oyster Creek are based on the "dryout correlation" approach implemented with ENC's approved NJP-BWR Fuel Heatup Model.(3) These limitsI4) are provided in Table 4.1 for comparison with the present results.
The new four-loop opera-tion HAPLHGR limit results for ENC 8x8 fuels are approximately lli, higher than those given in Table 4.1 from the Fuel Heatup Model analysis. Hence continued application of the existing four-loop operating limits (Table 4.1) to ENC 7x7 fuels is clearly conservative.
Table 4.2 repeats and extends the latest five-loop operation MAPLHGR limits for ENC fuel at Oyster Creek.(2) As noted these limits are based on the approved WREM-Based NJP-BWR ECCS Evaluation Model U) which has been applied identically without any changes in the present four-loop operation analysis. Also included in Table 4.2 are five-loop operation MAPLHGR limits for ENC 8x8 fuel at an exposure of 40,000 MWD /MTM.
These limits have been calculated in support of additional irradiation cycles for ENC 8x8 fuel at l
l I
. XN-NF-79-78 Oyster Creek.
The effects of burnup enhanced fission gas release for peak pellet exposures above 20,000 MWD /MTM have been included in the MAPLHGR calculations.
Figures 4.1 and 4.2 provide a comparison of the four-loop and five-loop MAPLHGR limits.
Except at low exposure the four-loop operating limits are about 2% less than those for five-loop operation. At beginning-of-life the four-loop MAPLHGR limit for 8x8 fuel type VB is 4.6% less thcn for five-loop operation. At exposures of 1200 and 1900 MWD /MTM for this fuel type the four and five-loop MAPLHGR limits are the same. However in the case of four-loop operation the calculated margin to the maximum allowable PCT of 2200*F is reduced by approximately 35*F.
I In summary, operation of the Oyster Creek Reactor within the new MAPLHGR limits provided in this report (Tables 1.1, 4.1 and 4.2) assures that the Oyster Creek Emergency Core Cooling System meets the Acceptance Criteria as presented in 10 CFR 50.46 and Appendix K.
That is:
1)
The calculated peak fuel element clad temperature does not exceed the 2200'F limit.
2)
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total amount of zircaloy in the reactor.
3)
The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching.
e XN-NF-79-78 4)
The system long term cooling capabilities provided for previous cores remains applicable for ENC fuel.
I
. XN-NF-79-78 Table 4.1 Existing Four-Loop Operation MAPLHGR Limits TYPE VA TYPE VB EXPOSURE MAPLHGR MWR EXPOSURE MAPLHGR MWR (GWD/MTM)
(kw/ft)
(%)
(GWD/MTM)
(kw/ft)
(%)
2.0 9.11 7.43 0.0 8.71 6.79 4.0 8.99 7.25 1.0 9.22 7.27 10.0 9.07 7.20 2.0 9.12 7.18 15.0 9.22 7.09 4.0 8.98 7.08 20.0 9.13 7.59 7.0 8.99 6.98 25.0 8.95 5.27 10.0 9.02 7.20 30.0 8.70 12.61 15.0 9.16 7.09 20.0 9.37 7.08 25.0 8.90 5.19 30.0 8.45 9.47 TYPE IIIE TYPE IIIF EXPOSURE MAPLHGR MWR EXPOSURE MAPLHGR MWR JGWD/MTM)
(kw/ft)
(%)
(GWD/MTM)
(kw/ft)
(%).
3.0 10.58 7.02 1.0 11.07 7.27 5.5 10.50 6.69 2.0 10.89 7.02 10.0 10.68 7.09 4.0 10.77 7.10 14.0 11.39 16.72 10.0 10.84 7.26 18.0
'1.09
'5.56 15.0 11.46 16.01 22.0
'0.99
'4.44 20.0 10.98 11.37 27.5 10.78 16.78 27.5 10.54 10.96 i
32.0 10.38 11.39 32.0 10.38 14.31
- XN-NF-79-78
~
Tabic 4.2 Oyster Creek Five-Loop Operation MAPLHGR Limits j
8 X 8 Fuel Type V Burnup MAPLHGR F, x F MWR PCT r
]
(GWO/MTM)
(kW/ft) g (op) 7.0 10.66 2.30 15.6 2161 11.0 10.45 2.26 16.8 2130 15.4 10.25 2.21 16.9 2153 23.9 10.09 2.18 16.9 2099 33.2 9.99 2.16 16.5 2094
- 40.0 9.50 2.05 14.0 2039' l
8 X 8 Fuel Type VB Burnup MAPLHGR F xF MWR PCT a
r (GWO/MTM)
(kW/ft)
,{gy (op) 0.0 10.36 2.24 7.7 2196 0.5 10.8 2.33 7.2 2142 1.2 10.8 2.33 6.9 2123 1.9 10.8 2.33 7.2 2140 7.1 10.8 2.33 13.1 2184 11.2 10.69 2.31 16.9 2169 15.6 10.40 2.25 15.2 2100 24.1 10.19 2.20 16.6 2097
- 33.3 10.04 2.17 16.0 2081
- 40. 'O
'9.55 2.06 15.3 2067 7 X 7 Fuel Type III E Burnup MAPLHGR F xF MWR PCT i
a r
I (GWO/MTM)
(kW/ft)
(%)
(*F) 6.7 12.43 2.15 16.2 2196 10.7 12.35 2.13 16.2 2178 15.4 12.22 2.11 16.3 2178 23.6 11.74 2.03 16.0 2106 32.8 11.45 1.98 16.8 2138 7 X 7 Fuel Type III F i
Burnup MAPLHGR F, x F MWR PCT (GWD/MTM)
(kW/ft)
M
(*F) i 6.8 12.63 2.18 16.0 2189 10.8 12.45 2.15 16.1 2178 15.3 12.19 2.09 17.0 2156 23.6 11.75 2.03 16.7 2157 32.8 1,.44 1.98 17.0 2126 I
Core Wide' Metal-Water Reaction is Less than 1.0% for all Cases.
I MAPLHGR Values added in present analysis
a 11.0 5 Loop 10.0 m
j 4 Loop P
e n
U E 9.0 7
8.0 5
7.0 M
0.0 10,000 20,000 30,000 40,000 h
Burnup, MWD /MTM Figure 4.1 Comparison of Four-and Five-Loop MAPLHGR Limits for ENC 8x8 Fuel Type V
~ l 9
3 5
5 5
11.0 5 Loop i
gE10.0
?
4 Loop U
n E
a 9.0 8.0 Y
- n 7o 0.0 10,000 20,000 30,000 40,000 s
Burnup, MWD /MTM Figure 4.2 ' Comparison of Four-and F.ive-loop.MAPLHGR Limits for ENC 8x8 Fuel Type VB
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e XN-NF-79-78
5.0 REFERENCES
(1) Exxon Nuclear Company, The Exxon Nuclear Company WREM-Based NJP-BWR ECCS Evaluation Model and Application to the Oyster Creek Plant, XN-75-55, Revision 2 August 1976; XN-75-55, Revision 2, Supplement 1, September 1976, and XN-75-55 Revision 2, Supplement 2, December 1976.
(2) Exxon Nuclear Company, Oyster Creek LOCA Analyses Using the ENC NFP-BWR ECCS Evaluation Model, XN-NF-77-55, Revision 1, March 1978.
(3) "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option - User's Manual," XN-CC-33(A),
Revision 1, November 14, 1975.
(4) Amendment 25 to License No. DPR-16, Jersey Central Power and Light Company, Oyster Creek Nuclear Generating Station Docket No. 50-219.
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