ML20237D312
| ML20237D312 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 11/30/1987 |
| From: | Humphreys M, Talbert R, Wolkenhauer W WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | |
| Shared Package | |
| ML17279A737 | List: |
| References | |
| WPPSS-EANF-111, NUDOCS 8712230182 | |
| Download: ML20237D312 (14) | |
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WNP-2 FINAL FEEDWATER TEMPERATURE REDUCTION
.j
SUMMARY
REPORT l
3 Prepared By: d [ [ M
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W. C. Wolkenhauer, Principal Core Analysis Engineer Nuclear Fuel Reviewed By:
hhh R. J.' MTtierf, FTant Engineer Reviewed By:
/J7 8. 8 M. C. Humphreys, Plant E(gineer l
Concur With:
R. O. Vosburgh, (dpager, Safety Analysis & Simulator Engineering NLmM Concur With:
i M. R. Wuestefy, Supervisor, WNP-2 Reactor Engineer Concur With W
K. D. Cowan, Manager, WNP-2 Technical dkw H/n./97 Approved By:
D. L. Larkin, Manager, Engineering Analysis and Nuclear Fuel t
)
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NOTICE i
This.. report is - derived in part through information provided - to Washington Public Power Supply' System (Supply System) by Advanced Nuclear Fuels Corpora-a tion and General Electric ' Company.
It is being. submitted by the Supply System to the U.S. Nuclear Regulatory Commiss' ion in ' partial support of the WNP-2 Application For Technical Specification Changes Relating to Operation ~ With Final Feedwater. Temperature Reduction at the End of the Fuel. Cycle.
The information contained herein. is true and' correct to the best. of the Supply System's knowledge, information, and belief.
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i WNP-2 FINAL'FEEDWATER TEMPERATURE REDUCTION
SUMMARY
REPORT g
TABLE OF CONTENTS, Page
)'
1.0 INTRODUCTION
'l 2.0 : TRANSIENT ANALYSIS.....................
2 h
2.1 Resul ts of Transient Analysis.............
2 I
3 2.2 ' Safety Limit 3.0 ' MAXIMUM OVERPRESSURIZATION.................
4 4.0 STABILITY ANALYSIS......................
4 5.0 MECHANICAL EVALUATION OF REACTOR INTERNALS.........
5 5.1 Loads Evaluation 5
5 5.2' Loads Impact 6.0 FEEDWATER N0ZZLE AND FEEDWATER SPARGER FATIGUE USAGE....
6 6.1 Feedwater Nozzle Fatigue 6
6.2. Feedwater Sparger Fatigue...............
7
)-
7.0 LOCA/ CONTAINMENT ANALYSIS 7
8.0
SUMMARY
8
9.0 REFERENCES
8
)
)
i
)
4 q
3
WNP-2 FINAL FEEDWATER TEMPERATURE REDUCTION 1
SUMMARY
REPORT
1.0 INTRODUCTION
The Supply System is planning to utilize Final Feedwater Temperature Reduction (FFTR) and subsequent thermal coastdown from 100 percent power to 65 percent power at the end of Fuel Cycle 3 and subs,equent fuel cycles as a means of increasing electrical generation from a given fuel cycle.
This report summarizes the analyses performed by Advanced Nuclear Fuels Corporation (ANF) and General Electric Company (GE) in support of FFTR.
The FFTR analyses were performed consistent with the WNP-2 reload and transient analyses (References 9.1 and 9.2) employing the same method-ology (References 9.3 and 9.4).
This evaluation is applicable to core flow. values up to the maximum attainable with the recirculation' flow con-trol valve in its fully open position which is 106 percent of the rated core flow value at 100 percent power and for periods extending past the normal cycle with final feedwater temperature reductions up to 65"F.
Specifically, the analysis conditions considered included a 12 month fuel
)
cycle, at the end of which the feedwater temperature is reduced by 65 F to a value of 355'F followed by thennal coastdown (Reference 9.5 and 9.6).
The results of the analyses summarized here are considered to be applicable to WNP-2 Cycle 3 and subsequent fuel cycles containing ANF and I
GE 8x8 fuel of similar design.
)
The analyses summarized here and developed more thoroughly in the refer-ence documentation considered transient analyses, reactor safety limit, maximum overp pressurization, reactor stability, core internals loads, feedwater nozzle and sparger fa tigue, and the reactor LOCA analysis.
These items are discussed in turn here.
)
The results of these analyses have identified several technical specifi-cation changes.
These proposed technical specification changes which are attached are identified by title in Table 1.1.
)
TABLE 1.1 PROPOSED TECHNICAL SPECIFICATION CHANGES 1
Index
)
1.0 Definitions 3/4.1.6 Feedwater Temperature 3
3/4.2.3 Minimum Critical Power Ratio B 3/4.1.6 Bases; Feedwater Temperature l
B 3/4.2.3 Bases; Minimum Critical Power Ratio 2
l l
1 2.0 TRANSIENT ANALYSIS FOR THERMAL MARGIN I
1 ANF considers eight categories of potential system transient occurrences i
for Jet Punp BWRs in Exxon Nuclear Plant Transient Methodology For Boil-ing Water Reactors, XN-NF-79-71, Revision 2 (Reference 9.3).
The three 3
most thennally limiting events evaluated for Cycle 3 of WNP-2 have also been evaluated for the FFTR operating states.
These transients are:
o Load Rejection Without Bypass (LRWB) o Feedwater Controller Failure (FWCF) j i
o Loss of Feedwater Heating (LOFH) 2.1 Results of Transient Analysis Load rejection without bypass and feedwater controller failure transients were evaluated wi th the technical specification scram speed (TSSS) and normal scram speed (NSS) based on WNP-2 measured values (Reference 9.7) for FFTR of 65'F.
evaluations assumed that the recirculation pump trip (RPT) function j
was operable.
The limiting delta CPR for FFTR conditions at the end j
of cycle (E0C) for normal scram times is 0.27 for NSSS vendor fuel and 0.25 for ANF fuel as detennined by the LRWB, RPT operable trans-ient.
The limiting delta CPR for FFTR conditions at EOC for TSSS is
]
I 0.32 for NSSS vendor fuel and 0.29 for ANF fuel as determined by the LRWB, RPT operable transient.
For normal scram speeds, the effect of FFTR is to increase the delta CPR for the LRWB event by up to 0.02 and decrease the delta CPR for the FWCF event by as much as 0.01 for ANF fuel and 0.01 for fuel supplied by the NSSS vendor.
These changes in delta CPR results are applicable to future reloads where
)
the cycle is extended by the use of FFTR provided the assumptions made in performing the analysis remain applicable (Reference 9.5).
The loss of feedwater heating (LOFH) transient results for normal operation were obtained from WNP-2 specific analysis which were reviewed and determined to bound operation with FFTR for extended
)
cycle operation.
Due to the potential reactor power increases should an uncontrolled recirculation flow increase occur from a less than rated core flow state, the need exists for an augmentation of the operating limit MCPR (full or extended flow) for operation at lower flow conditions.
The reduced flow MCPR operating limit was evaluated for FFTR opera-tion utilizing the same methodology applied for the reload analysis.
A power distribution was chosen such that the MCPR equals the safety limit (1,06) at the final power / flow run-up point (the safety limit is discussed more fully in Section 2.2).
The reduced flow MCPRs were then calculated by XCOBRA (Reference 9.8) at discrete reduced f l
core flow and core power states, maintaining the power distribution constant.
However, for operation in the FFTR mode, the resulting subcooling increase and flow redistribution at the lower flows-increases the hot channel. MCPR making the FFTR analyses slightly more restrictive.
The reduced flow MCPR operating limit' while operating in the FFTR mode 'is tabulated in Table 2.1.
While in' the - FFTR mode, the MCPR operating limit for WNP-2 shall be the maximum of this reduced flow MCPR operating limit and the FFTR full flow MCPR operating limit.
This result is applicable to future cycles when the temperatures reduction is equal to or less than 65*F (Reference 9.5)..
TABLE 2.1 REDUCED FLOW MCPR OPERATING LIMIT FOR FFTR OPERATION Core Flow Reduced Flow MCPR
(% Rated)
Operating Limit t
100 1.07 90 1.13 80 1.19 l
l 70 1.26 60 1.34 50 1.45 40 1.59 2.2 Safety Limit The MCPR safety limit is the minimum value of the critical power ratio (CPR) at which the fuel could be operated where the expected number of rods in boiling transition would not exceed 0.1 percent of the fuel rods in the core.
The operating limit MCPR is established such that in the event of the most limiting anticipated operation transient, the safety limit will not be violated.
The safety limit for all fuel types in the WNP-2 Cycle 3 core was-j confirmed to be 1.06 The input parameters and uncertainties used i
to establish the safety limit were reviewed for application to the FFTR conditions.
It was concluded that a safety limit of 1.06 is applicable to the FFTR mode (Reference 9.5).
1 l L__ - - __ _
I 3.0 MAXIMUM OVEP, PRESSURIZATION
]
Maximum system pressure has been calculated for the containment isolation l
event (rapid closure of all. main steam isolation valves) with an adverse scenario as specified by the ASME pressure vessel code (Reference 9.5).
This analysis showed that with six safety valves out of service, the 1
remaining safety valves of WNP-2 have sufficient capacity and performance 1
.to prevent the system pressure from reaching the established transient I
system pressure safety limit of 110 percent of design pressure during the cycle extension period using increased core flow and final feedwater temperature reduction.
The maximum system pressures predicted are shown in Table 3.1.
As can be seen from review of this table, not only are the
-predicted values below the 110 percent design value (1,375 psig), they are also below the nonnal feedwater temperature case.
Therefore, the normal feedwater temperature ASME overpressurization results will be con-sidered to bound the FFTR conditions in future cycles.
TABLE 3.1 MAXIMUM VESSEL PRESSURE (PSIG)
Vessel Vessel Lower Steam Transient FFTR Dome Pressure Plenum Pressure Line Pressure MSIV Closure No 1,285 1,313 1,287 MSIV Closure Yes 1,273 1,306 1,280
)
4.0 STABILITY ANALYSIS 1
Cycle 3 stability calculations have been performed for WNP-2 for opera-tion with FFTR (Reference 9.5).
The calculations were perfonned at the two limiting power flow conditions for nominal feedwater temperature.
The calculated decay ratios for reduced and nominal feedwater temperature are shown in Table 4.1.
TABLE 4.1 CALCULATED DECAY RATIOS 1
l Decay Ratio
)
Power Flow Reduced Feedwater Nominal Feedwater I
% Rated
% Rated Temperature Temperature 65.0 45.0 0.52 0.49 48.0 27.6 0.70 0.84 i
I l
_4_
'The reduced feedwater - temperature ' stability calculations were performed -
for a 65*F decrease in the feedwater temperature.
The stability calcula-tions support the ' current Technical Specification stability detect and
~ suppress power flow bcundaries for the planned Cycle 3 final. reduced feedwater temperature operation (Reference 9.5).
The decay ratio changes for reduced feedwater temperatures shown in Table 4.1 are. either small or conservative such that the results of this analysis are applicable to l
future reloads where the cycle is extended by the use of FFTR provided the power / flow assumptions made in performing the analysis remain applic-able (Reference 9.9).
5.0 MECHANICAL EVALUATION OF REACTOR INTERNALS Evaluations were perfomed to determine bounding acoustic and flow induced loads, reactor internal pressure difference loads, and fuel sup-port loads for FFTR operations (Reference 9.6).
_S.1
-Loads Evaluation 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 ampli-tude is the sum of the initial pressure subcooling plus the experi-mentally determined pressure undershoot below saturation pressure.
FFTR operation increases the expected acoustic loads because this downcomer subcooling increases and, therefore, the total pressure wave amplitude increases.
The high velocity flow patterns in the downcomer resul ting from a recirculation suction line break also create lateral loads on the reactor vessel internals.
These loads are proportional to the square of the critical mass flow rate out of the break.
The additional subcooling in the downcomer resul ting
)
from FFTR operation leads to an increase in the critical flow and, therefore, to a corresponding increase in the-flow induced loads.
5.2 Loads Impact The reactor internals most affected by FFTR operation are the core plate, shroud support, shroud, top guide, shroud head, steam dryer, control rod guide tube, control rod drive housing, and jet punp.
These and other components were evaluated using bounding loads under normal, upset, emergency, and faulted conditions.
It is concluded that the stresses produced in these and other components are within 1
the allowable design limits given in the WNP-2 Final Safety Analysis Report (FSAR) (Chapters 3 and 4) or the ASME Code,Section III, Subsection NG.
The resul ts of these evaluations are fuel cycle independent.
l
6.0 FEEDWATER N0ZZLE AND FEE 0 WATER SPARGER FATIGUE USAGE 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.
The system cycle stresses are based on limiting cycles that use the maximum temperature range possible to show expected worst condi-tions.
Rapid cycling is caused by small, high frequency temperature fluctuations caused by mixing of relatively colder nozzle annulus water with the reactor coolant.
The introduction of FFTR 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 themal stress analysis.
GE has developed stan-dardized rapid cycling duty maps for each BWR plant that cover the design l
basis rapid cycles in the same manner that themal cycle diagrams cover the design basis thermal transients (systeu cycling).
The methodology used to develop the duty maps is based on the results of extensive test-ing of feedwater nozzles by GE.
FFTR is analyzed by modifying the design cycles in order to gauge its effect on fatigue usage.
The reduced feedwater temperature will tend to increase fatigue usage due to an increase in themal stress.
An evaluation of the effect of FFTR on the feedwater nozzle and feed-water sparger fatigue for WNP-2 was performed for the following condi-tions (Reference 9.6).
As the last step in a 12 month fuel cycle, FFTR to a feedwater temperature of 355 F (65'F reduction from nominal rated feedwater temperature) at rated power for 18 days was followed by a 3 percent per week coastdown over 12 weeks to a final power of 65 percent.
The coastdown was initiated from a reduced feedwater temperature of 355 F.
The associated feedwater temperature at the end of the coastdown was 321 F.
6.1 Feedwater Nozzle Fatigue The original stress analysis of the feedwater nozzle showed that the maximum system cycling fatigue usage factor for the nozzle blend radius region was 0.6524 for emergency and faulted conditions (Ref-
)
erence 9.10).
The usage factor for rapid cycling using the design basis (unmodi fied) duty map is 0.2047, providing a total 40 year j
usage factor of 0.8571.
The usage factor for rapid cycling includ-ing FFTR operation is 0.2796 providing a total 40 year usage factor of 0.9320.
This result is based on FFTR operation during every 12 month cycle for the life of the plant.
This is equivalent to 0.0019
)
fatigue damage per cycle of FFTR operation.
The 40 year total usage j
factor remains below the ASME Code Limit of 1.0 with FFTR operation i
and is thus considered acceptable.
l l
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6.2 Feedwater Sparger Fatigue Feedwater sparger fatigue usage.is calculated in' the same manner as j
feedwater nozzle fa tigue usage.
- However, since the feedwat'er sparger 'is not an ASME Boiler and Pressure Vessel Class 1 Code com-ponent, a fatigue analysis was not originally required. WNP-2 has a welded' single thermal sleeve design which does not allow leakage of feedwater flow to occur at the safe.end as do other thennal sleeve designs.
This leakage flow is the primary contributor to sparger
{
fatigue usage.
Therefore, sparger fatigue usage is much less affected by changes in feedwater flow and temperature for the welded single sleeve design.
The sparger is made from stainless steel material which is less susceptible to high cycle fatigue than the j
low alloy steel of the nozzle as evidenced by the - differences in their respective fatigue curves.
Small changes in the (high cycle)-
portion of the fatigue curve can cause very significant changes in 1
fatigue usage (i.e., a relatively small change in stress can cause a j
very significant change in the allowable number of cycles).
- Thus,
)
it becomes evident that the sparger fatigue damage is much less j
severe than nozzle fatigue damage during feedwater condition changes l
like FFTR for the welded single sleeve design.
Since the nozzle j
fatigue damage is so low (0.0019 per cycle), the sparger damage will be insignificant and, therefore, can be neglected.
7.0 LOCA/ CONTAINMENT ANALYSIS l
f LOCA analysis performed by GE for WNP-2 show that operation without FFTR bounds operation with FFTR (Reference 9.6).
FFTR decreases core inlet coolant temperature and increases the initial coolant mass in the reac-tor. This would result in a reduction in peak cladding temperature.
The impact of feedwater temperature reduction operation on the contain-ment LOCA response was evaluated by the reactor vendor (Reference 9.6).
l Operation with FFTR causes a slight increase in the initial drywell pressurization rate over the rate reported in the WNP-2 FSAR.
The calcu-lated peak values for drywell pressure and wetwell pressure under FFTR are bounded by the corresponding values for the WNP-2 FSAR (Chapter' 6) l l
conditions.
The peak value for drywell floor differential pressure (download) is bounded by the appropriate design limit of 25 psid.
All 1
other containment parameters are bounded by the results reported in the WNP-2 FSAR.
The LOCA related pool swell, condensation oscillation, and chugging loads were evaluated at the worst power / flow conditions during FFTR operation.
Pool boundary pressure load during pool swell under FFTR conditions exceeds the load calculated based on WNP-2 FSAR conditions by less than 2.2 percent.
However, this load and all other pool swell loads are bounded by the appropriate design loads.
The condensation oscillation and chugging loads with FFTR conditions are also bounded by the appro-priate design loads. _ - - _ _
The GE analysis was reviewed by ANF with regard to the potential impact of FFTR on ANF-performed loss-of-coolant-accident for WNP-2.
As a result of this review, ANF concludes that the ANF ECCS analysis at normal feed-water temperatures al so bounds FFTR: conditions (Reference 9.11 ).
In addition, the reported pressure loads on the containment structure cal-culated by GEc remain applicable for' ANF reload fuel in Cycle. 3 and sub-sequent cycles for FFTR conditions.
8.0
SUMMARY
Analyses have been performed and are summarized here to describe the impacts of operation of WNP-2 with FFTR at the end of Fuel Cycle 3 and subsequent fuel cycles.
The analyses were performed consistent 'with the WNP-2 reload and transient analyses.
The analyses summarized here con-sider transient analyses, the reactor safety limit, maximum overpressure-ization, reactor stabili ty, core internal s, feedwater nozzle fa tigue, feedwater sparger fatigue, and the reactor LOCA analysis.
The results of these analyses have identified several required Technical Specifications changes which have been described and which are attached to this report.
The above analyses and the attached Technical Speci fication changes support operation of WNP-2 with FFTR at the end of Fuel Cycle 3 and sub-sequent fuel cycles.
9.0 REFERENCES
9.1 J.
E.
Krajicek, " Supply System Nuclear Project Number 2 (WNP-2)
Cycle 3 Reload Analysis",
XN-NF-87-25, Advanced Nuclear Fuel s l
Corporation, Richland, Washington 99352, March 1987 9.2 J.
E.
- Krajicek, "WNP-2 Cycle 3
Plant Transient Analysi s",
XN-NF-87-24, Advanced Nuclear Fuel s Corporation,
- Richland, Washington 99352, March 1987
)
9.3 R.
H., Kelley, " Exxon Nuclear Plant Transient Methodology For Boiling Water Reactors", XN-NF-79-71(P), Revision 2 (as supplemented), Exxon Nuclear Company, Inc., Richland, Washington 99352, February 1987 9.4 M. J. Ades, "XCOBRA-T:
A Computer Code For BWR Transient Thermal-Hydraulic Core Analysis", XN-NF-105(A), Volume 1,
Supplement 1,
Volume 1,
Supplement 2,
Advanced Nuclear Fuel s Corporation, Richland, Washington 99352, February 1987 l
9.5 J. E. Krajicek, "WNP-2 Plant Transient Analysis With Final Feedwater Temperature Reduction", XN-NF-87-92, Advanced Nuclear Fuels Corpora-tion, Richland, Washington 99352, June 1982 9.6 S.
Wol f,
" Safety Review of Washington Public Power Supply System Nuclear Project Number 2 at Core Flow Conditions Above Rated Flow l
Throughout Cycle 1 and Final Feedwater Temperature Reducti on",
NE-DC-31107, General Electric Company, San Jose, California
- 95125, narcn two i _ _ _ - _ -
i 9.7 W.
C.
Wolkenhauer, "WNP-2 Cycle 2
Reload Summary Report",
WPPSS-EANF-101, Washington Public Power Supply System, Richland, Washington 99352, February 1986 9.8. T. W. Patten, " Exxon Nuclear Critical Power Methodology For Boiling Water Reactors", XN-NF-524( A), Revision 1, Exxon Nuclear Company, Inc., Richland, Washington 99352, November 1983 1
9.9 J.
- Maryott,
" Personal Communica tion",
Advanced Nuclear Fuel s Corporation, Richland, Washington 99352, July 1987 9.10 "Hanford 2 - 251 BWR-5 Stress Report For Feedwater Nozzle", Section E4, Contract 72-2647, Chicago Bridge and Iron Nuclear Company,1973 9.11 J.
B.
Edgar, letter to Washington Public Power Supply Systen,
" Impact of Final Feedwater Temperature Reduction on t.0CA",
ANFWP-87-0098, Advanced Nuclear Fuels Corporation,
- Richland, Washington 99352, July 22,1987 I
l _-_ _ _ _