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/ 7 PHILADELPHIA ELECTRIC COMPANY 1000 CHESTNUT STREET PHILADELPHIA. PA.19105 w4um u=.
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July 30,1971 r.
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Dr. 7%ter A. Morris, Director Division of Reactor Licensing
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- i.,'fr e U. S. Atomic Energy Comunission i
'g Washington, D.C.
205h5 s
Dociret t 50-352
~353 Subject Conformance of the Limerick Generating Station, thits 1 and 2 to the ABC Interim Acceptance Criteria for Emergency Core Coolian
Dear Dr. Morris:
This letter is written in response to your letter of July 19, 1971 regarding conformance of the Limerick Generating Station thits 1 & 2 to the ABC Interim Acceptance Criteria for E 4,. cy Core Cooling, issued on June 19, 1971.
In that release new criteria and ground rules for evaluation and seceptability of the emergency core cooling systems for both Boiling Water Reactors and Pressurized Water Reactors were identified.
We are pleased to report that the emergency core cooling network for the Limerick project satisfies all of.these new interim criteria. The details of the evaluation leading to this conclusion are found in the attached memorandum entitled "Conformance of Mt ECCS to AEC Interim Acceptance Criteria for Light Water Reactors,"
June 19,1971.
Although the new ground rules specf.fied by the Cormission in its June 19 policy statement appear to be overly conservative for design basis evaluations, it is reassuring to note that the emer-gency core cooling systems for the Limerick project are still f ound to be rnore than acceptable based on the new index.
Yours very truly, E. G. Bauer, Jr.
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LIMERIC7. GENERATING STATIST
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- UNITS 1 & 2 4
ECCS CONFOMACE TO NEW AEC AD0PTED INTERIM ACCEPTANCE CRITERIA l
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i July 23,1971 3472
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INTRODUCTION The following report is submitted in response to an AEC request for additional infortnationIII on the conformance of the Limerick
.f facility to the newly adopted AEC interim criteriaI2)
This docu-
,j ment (its contents and references) fully describes the basis for the acceptance of the emergency core cooling system for Limerick.
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b I--l (1) Letter, Peter A. Morris to Vicent P. McDevitt dated July,1971;
Subject:
Additional Information on LGS-ECCS.
3, (2)
Interim Policy Statement, US AEC, dated June 19, 1971;
Subject:
AEC Adopted Interim Acceptance Criteria for Perfor-i
- nance of ECCS for Light-Water Power Reactors.
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II.
REQUEST FOR ADDITIONAL INFORMATI0](
The following delineates the requested additional information).
"1.
Provide curves of peak clad temperature and percent I
clad metal-water reaction as a function of break size for the various combinations of ECC subsystems evaluated
)
by using the single failure criterion indicated in Table 2-1 of the topical report: " Loss-of-Coolant Accident and i
f Emergency Core Cooling Models for General Electric Boiling l
Water Reactors," NED0-10329. A discussion should be in-l cluded showing the justification'for the ECC subsystes f
combinations used in the evaluation.
l 2.
For several breaks that typify small, intermediate and large breaks, provide curves of (a) peak fuel clad temperature for various rod groups, (b) core flow, (c) fuel channel inlet and outlet quality, (d) heat transfer coefficients, (e) reactor vessel water level,
}
and (f) minimum critical heat flux ratio (MCHFR) as functions of time.
Indicate the time that effective core cooling is initiated, the time the fuel channel becomes wetted based upon Item 4 of Appendix A, Part 2, and the time that the temperature transient is tensinated.
3.
For the analyses performed in 1 and 2 above, discuss the range of peaking factors studied and the basis for selecting the combination that resulted in the most severe thennal transient.
Curves of peak clad tempera-4 ture vs time for the range of peaking factors studied should be included.
4.
Discuss in detail 'any deviations in the evaluation model l
used in the foregoing studies from that described in l
Appendix A, Part 2 of the Commission's Interim Policy Statement."
(1) Letter, Peter Morris to Vicent P. McDevitt dated July,1971;
Subject:
Additional information on LGS-ECCS.
s 7,
r III.
ADDITIONAL INFORMATION RESPONSE The enclosed information is supported and supplemented by the
' appropriate cited references previously submitted on these dockets.
General Following General Electric's request for a generic review of the ECCS analytical models approximately one year ago, the AEC DRL staff has been re-evaluating the models and assumptions employed in the design and performance of the Emergency Core Cooling Systems for the BWR.
No major concern was uncovered during this review, completed in r
December,1970, tha changed the confidence of the General Electric Company nor the App icant that the systems provided on BWR's unre fully adequate in e9ery respect.
Following the review, the G.E.
Design Basis Assumptions 'and models used were documented in detail in G.E. Topical Report NEDO 10329 and submitted on April 9,1971.
Following submission of that report and further study by a senior
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DRL task force and the ACRS, an additional series of sensitivity studies were orally requested by the AEC on April 16, 1971 froni G.E.
In response to these specific questions, Supplement fl to the cited Topical Report was submitted on April 29,1971.
These sensitivity studies encompassed a number of key assumptions including most of those later included in the AEC " Interim Acceptance Criteria."
The ECCS were again found to be fully acceptable.
On May 24,1971, the AEC requested an additional set of calculations employing all s the assumptions included in the AEC " Interim Acceptance Criteria for ECCS" of June 19, 1971, and the ECCS were again found to be fully adequate. Oral responses to this request were transmitted to the AEC task force on May 28, 1971.
The G.E. and Applicant position is that the prior design bases em-ployed by General Electric as fully documented in Topical Report NEDO 10329 sunnarized in all FSAR/PSARs are still appropriate and do
.f indeed provide a conservative evaluation of the ECCS performance.
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The new AEC " Interim Acceptance Criteria for ECCS" issued on June 19, 1971, are viewed by the General Electric Company and the Applicant as
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f providing a more conservative index for further assurance of ECCS acceptabili ty.
In order to be fully responsive to all aspects of the new AEC criteria, conformance to this newsindex of acceptability is being documented here within to complete the safety anyyses for this project.
To place all the results in proper context, the previously submitted General Electric and Applicant Design Basis documentation is compared j
in Table I with all the temperatures and models ' considered. TMs comparison provides an appropriate indication of the magnitude of margin which exists in the G.E./ Applicant Design Basis, as weH as the extensive margin which exists in this new AEC index to appropriate limits.
Included below in this document is'a discussion of the i
application of all the AEC Criteria, assumptions as well as the re-sults demonstrating acceptability and the results of the requested analyses. The responses to the cited questions provides further details which gives inkight into the analytical models and provides an understanding of the effect of key parameters in regards to the calculated peak clad temperatures.
Discussion of AEC General Criteria The general acceptance criteria which must be met by all light water reactors are listed in Section IV A, page 7 of the AEC Interim Acceptance Criteria and are discussed below.
A.
AEC Criteria (for all light water power reactors)
These general requirements have been the basis of AEC safety review for some time.
On the basis of today's knowledge, the perfomance of the emergency core cooling system is judged to be acceptable if the cafculated course of the loss-of-coolant accidentU) is limited as follows:
1.
The calculated maximum fuel element cladding temperature does not exceed 2300*F.
This limit has been chosen on the basis of available data embrittlement and possible subse-quent shattering of the cladding. The results of further -
-. ~ - ~,.-
detailed experiments could td the basis for future revision of this limit."
(1)
"A loss-of-coolant accident is a postulated accident that results from the loss of reactor coolant ct a rate in excess of the capability '.,f the reactor cool-ant makeup system from breaks in the reactor coolant pressure boundary, up to and including a break equiva-j lent in size of the double ended rupture of the I
g largest pipe of the reactor coolant systmus."
2.
"The amount of fuel element cladding that reacts chemically
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with water or steam does not exceed 11 of the total amount of cladding in the reactor."
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1 3.
"The clad temperature transient. is terminated at a time
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when the core geometry is still amendable to cooling, and before the cladding is so embrittled as to fail during or after quenching."
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4.
"The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core."
8.
Gr/f,rplicant Position on AEC General Acceptance Criteria 1.
The maximum temperature limit of 2300*F for any size break anywhere in the'reacto' systems is met, even when using the r
new conservative AEC assumptions. This index of accept-ability is based upon limiting the oxidation of the zircaloy cladding to avoid embrittlement and possible fragmentation upon cooldown. General Electric believes there is con-siderable margin in this number which should be recognized as being present.
As discussed in detail in Pilgrim Nuclear Power Station, Unit 1, Amendment #14. (AEC Docket No. 50-293))
i a limit of 2700*F appears reascrable for the times of interest to BWR ECCSs.
Even this latter value is conservative since l
the test data show embrittlement only for material which was
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taken to considerably higher te :.eratures than 2700'F.
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2.
The metal water reaction limit of 1% of the active fuel cladding in the core is met, since the maximum metal water reaction for this plant is less than 0.12% of the active fuel rod cladding using all the AEC assumptions.
This is conservative by at least a factor of 2 to 3 when compared to data. Therefore it is well.below the AEC limit of 11.
For breaks smaller thaa the Design Basis Accident (DBA),
the core metal-water reaction is even less. The AEC limit of 11 metal water appears reasonable provided it is used consistently, and it is recognized that it is nearly a decade above the calculated value. However, metal water reaction is a surface phenomena and a better criterion would be to base the metal water reaction on cladding sun face area rather than the total weight zircaloy in the cladding.
3.
Clad embrittlement is avoided for.all breaks up to and fe-cluding the Design Basis Accident, since the temperatures are well below 2300*F. Locally and throughout the core, the degree of metal-water reaction is very low compared to that which causes embrittlement upon cooldown.
Even at a postulated peak of 2250'F and for times well in excess of those which occur in the reactor, the maximum local metal-water reaction is only 5.1%, determined experimentally and reported in a General Electric Report (GEAP 13112, p. 49).
This compares to about 17% required to reach fragmentation.
It should be recognized, however, that even with clad fragmentation, the ECCS's have a high probability of success in terminating the temperature transient and this is built-9.
in margin tha't is present.
8 4.
For any single failure as outlined in NEDO 10329, there will be at least one core spray.
Also, within a short time the core refloods to the 2/3 elevation. Thus, once the core is I
flooded, the temperature remains at saturation indefinitely.
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t However, even with no sprays, the core when flooded to tl 2/3 elevation will limit the peak temperature to below 9C as documented in detail in Quad Cities, Units 1 & 2 Amer ment #26 (AEC Docket Nos. 50-254/265). This is applicabl to all BWR's.
Thus, the long-tenn core cooling is assure redundantly in the BWR. Whenever there is either a core spray in operation or flooding action occurring, long-ten cooling occurs. This situation is outlined in NEDO 10329 l
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and is documented in both Quad Cities, Units 1 & 2 Amendr'
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- 26(AECDocketNos. 50-254 & 50-265) and Zinner Nuclear
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Station, Unit 1. Amendment #12 (AEC Docket No. 50-358). -
i Discussion of AEC Assumptions 1;
The evaluation was performed using the new conservative AEC assumptions
<1 no deviations. The AEC assumptions to be used in conjunction with the i.
in Topical Report NEDO 10329 are given in Appendix A, Part 2 of the AEC policy statement of June 19,1971, " Interim Criteria for Emergency Core Systems for Light Water Reactors," Page 17, in which it statcs:
" Analyses should be performed for the entire break spectrum, up to and i ciuding a double-ended severance of the largest pipe of the reactor cool pressure boundary. The combinations of. systems used for analysis should derived from a failure mode and effects analysis, using the single failui criterion as indicated in Table 2-1 of the topical report " Loss of Coolar Accident and Emergency Core Cooling Models for General Electric. Boiling i Reactors," NEDO 10329. The analytical techniques described in NEDO 10325 f ts supplement should be used with the following exceptions:
(AEC Exception 1:)
"During the period of flow coastdown af ter the minimum critical heat flux ratio at the hot spot is less than one and until the top of the jet pumps cover, the heat transfer coefficient shculd be calculated using the D.C.
Groeneveld correlation (AECL-3281, egaation 5.7)."
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i th2 (GF/ Applicant Comment) 900*F In most BWR plants, with few exceptions, the LOCA leads to uncovery os iend-jet pumps prior to the occurrence of the critical heet flux condition <
ble Therefore, this exception is not necessary.
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(AECException2:)
"During the period of lower plenum flashing untf1 the care becomes un3
,29 the heat transfer coefficient should be calculated using Groeneveld's n h nt lation as in 1 above."
r (GE/ Applicant Comment:)
The Groeneveld Correlation was applied during this period as directed.
it h grossly conservative for two reasons. First, den the hot spot {
ins w m the critical heat flux, it has been experimentally detanrfned that ac@
'e mod 21s out and film boiling is delayed for several seconds; thus removing an (
0 amount of stored heat during blowdown. Secondly, preliminary data indi re Cooling that rewdting will occur and nucleate boiling become re-established da lower plenum flashing when a surface temperature drops below a given vc The G.E. design basis heat transfer coefficient is an attempt to incluj I I"~
analytically simple manner the combined phenomenological effects abovec
'olant hoped that additional future experimental work in this area will demone sid b with sufficient assurance to the AEC that the GE Desige Basis is, indeG I"#8 sufficiently conservative.
lant g Water (AEC Exception 3:)
329 and "The heat transfer coefficients associated with rated core spray flow st co-respond to those derived from experimental data, assaing the claddit channel box emissivity is equal to 0.9."
.ux (GE/ Applicant Comment:)
ps un-As indicated in Supplement 1 to NEDO 10329, the best ag*eement with the for core spray convection coefficients is obtained with an assumed emisg of 0.6 or 0.7.
Actual measurements of the emissivity of the heaters use the tests also indicate a value of 0.6 to 0.7 should be : sed.
- However, shown in Supplement #1 to NEDO 10329, the assumed value cf emissivity h@
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i a small effect on the resulting coefficients and only a minor effect on the predicted cladding temperature responso, if applied in consistent manner.
The analyses using AEC assumptions employ a test bundle emissivity of 0.9.
(AEC Exceptio
- i "It should be a..,umed that channel wetting does not occur until 60 seconds following the wetting time calculated using the Yamanouchi analysis."
(GE/ Applicant Coment:)
For those cases in which core spray is the ECCS cooling mechanism, the time
)
at which the channel box is cooled by rewetting is an important variable on which extensive experimental data now exists.
The Yamanouchi analysis is one n
of the more rigorous treatments of 'this effect.
The G.E. design basis in-volves a conservative fitting of the Yamanou, chi derived data encompassing two-thirds of all the points., As required by the AEC exception, the fitting has been made more conservative by adding 60 seconds to the best fit of the-Yamanouchi data and thus encompassing over 96 percent of data. This is obviously more conservative; however, it has been employed in the analyses as instructed.
(AEC Exception 5:)
"A range of conservatively calculated peaking factors should be studied and the combination selected which results in the most severe thermal transient for the break spectrum and combinations of systests analyzed."
s (GE/4Mic nt Coment)
A generalized study was performed to determine the worst set of peaking fectors which are applicable to the Limerick Generating Station and other cores of this class. The range of peaking factors studied included variations of the outside rod peaking factors from 1.21 to 0.94 ; central rod peaking factors varied from 0.80 to 1.05.
The combination that resulted in the most severe thermal transient was selected by performing the core heatup analyses with the various possible distributions of peaking factors which occur as a function of bundle exposure.
t Due to the nature of the heat transfer during core, spray cooling, the "hig:eest" local peaking is not the worst peaking factor combination in the bundle which
,.results in the highest temperatures.
The worst peaking combination is the flatter, shape in Ehich the central rods have higher relative power. Thus, the peaking factor distributions calculated as a func, tion of exposure have been examined and 'the one which results in the highest fuel cladding temperatures following a LOCA is the one chosen as the " worst" local peaking. The maximum linear power generation used in conjunction with this range of peaking factors corresponds to the maximum attainable within Technical Specification limitJs for the particular fuel exposure / peaking factor combination studied. The same approach of using the worst peaking was used to determine the maximum tva-f tures for the small break analyses.
t Figures 13 to18 are curves of peak cladding temperature vs. time showing the j
results of this generalized study.
Figure 18a shows that for this class of cores, th'e maximum cladding temperatures are not significantly affected by the
)
change in local peaking factors with exposure. The exposure, resulting in the l
worst case peaking factors for this class of cores is shown in Figure 18a.
I I
(AECException6:)
'The decay heat curve described in the proposed ANS Standard,' with a 20%
allowance for uncertainty, should be used.
The fraction of decay heat generated in the hot rod should be considered to be 100% of this value unless a smaller value is justified. The effect of voids on reactivity during the blowdown may be taken into account."
(GE/ Applicant Comment:)
The design basis for power generation within the fuel rods during the loss of coolant accident has been a standard for a considerable period of time.
This standa'rd, which is reported in NEDO 10329 involves the use of the ANS standard for decay heat with a conservative treatment of the heavy isotopes, as well as no credit for voiding or flow decay.
This was felt to be sufficiently con-servative to become an appropriate standard for all BWR analyses.
10
/
s By AEC Exception 6, this standard was modified as instructed.
Rather than using the ANS standard, a value which is 20l greater is being used.
A conservative interpretation of the voiding which occurs during the loss of coolant accident is now being included in the AEC requested index analyses, as well as in all future G.E. design basis calculations. A plot of this new I
.po, er genera'tfon function is included in this report as Figure 19.
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Although it is felt that the G.E. design basis was conservative, this AEC Exception 6 makes the power generation assumption even more conservative.
Discussion of Requested Analyses Results The following summarizes the results of the analyses requiested.
The para-
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graph numbers correspond to the specific questions asked.
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)
1.
Fjgure 1 is a break area spectrum analysis of the peak clad tempera-
)
ture for the worst single failure, i.e., failure of the LPCI injection l
valve. Also shown on the, Figure is th5 DBA(X) and worst' intermediate break (dashed line) temperature for' the worst single failure of a diesel generator.
Failure of the LPCI injection valve will disable the entire LPCI subsystem, leaving two complete core spray sys'tems to cool the core.
Failure of the most critical diesel generator will result in having 3 LPCI pumps and one and one-half core spray systems.
. This assumes failure of one of the diesels which drives one LPCI pump l
and one (one-half capacity) core spray pump.
2a.
Figures 2, 3, and 4 show the peak clad temperature for four rod groups for a small intermediate and large break, respectively.
The inter-mediate break size is the specific one which results in the highest peak cladding t.emperature in the smaller break size range.
2b.
Figure 5 shows the core flow versus time for the Design Basis Acci-dent.
The core flow' transient for small and intermediate breaks is much less severe than that for the DBA and is similar to a pump trip transient.
This was discussed in detail in NEDO 10329.
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l 2c. Figure 20..shows core inlet and outlet quality versus time for.the DBA. The curve is shown for only the DBA because quality affects the film boiling heat transfer coefficient.
For small and inter-mediate size breaks, nucleate boiling is assured as long as the core is covered and nucleate boiling hea,t transfer coefficients are in-dependent of fluid quality.
When the core uncovers, the heat transfer coefficient is assumed to be zero, even though a significant steam cooling coefficient would actually exist.
2d. Figures 6, 7, and 8 show the heat transfer coefficient versu.s time for the small, intermediate and large break, respectively. The intermediate break size is the one which results in the highest peak cladding temperature in the smaller break size range.
2e. Figures 9,10, and 11 show the RPV water level versus time for the 9
small, intermediate and large breaks.
2f. Figure 12 shows the minimum critical heat flux ratio (MCHFR) versus time for the design basis accident. Because the flow transient for the small and intennediate break sizes is very mild compared to the DBA, it is now shown.
The MCHFR for the smaller breaks is always greater than unity as long a's the core is covered and nucleate boiling 1s always assured.
3.
Figures 13,14, and 15 show the peak cladding temperature response typical of this class of cores for three fuel exposures throughout the lifetime of the fuel for an intermediate break event. Figures
' 16,17, and 18 show similar results for the design basis accident.
Although these calculations were not performed specifically for
~
this power station, the controlling core parameters are the same as the generalized study.
Therefore, the worst set of peaking factors I
from this study will also be the worst set for the Limerick Generating Station.
Figure 18 (a) clearly shows that the maximum cladding temperature for the intermediate break event occurs at 30,000 MWD / ton 6.
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and at 1,000 MWD / ton for the DBA. These peaking factors sere then 4
used in conjunction with the actual flow and other ECCS parame'ters to determine the maximum temperatures for this power station. !
4.
There are no deviations from the evaluation model described in Ap-pendix A. Part 2 of the AEC Interim Policy Statement in the above analyses.
]
Conclusions It is, concluded that.....
(a) Peak clad temperatures are well below the 2300'F accep+ ability
- limit, I
i (b) The amount of fuel cladding reacting with steam is nearly an order below the 1% acceptability limit, (c) The clad temperature transient is. terminated while core geometry is still amenable to cooling, (d) The core temperature is reduced and the decay' heat can be removed for an extended period of time.
IV StfMARY The ECCS for the Limerick Generating Station, Units 1 & 2, are in conformance with the AEC criteria of acceptance using all the assumptions requested.
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,a LIMERICK PROJECT TABLE 1 PEAK CLAD TEMPERATURES Single Lgrge Failure (A)
Break.
Temp *F
' Temp *F Area ft Assumed
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AEC Index of Acceptability 2300 2300 Case I 1.
AEC Assumptions LPCI Inj. (B) 1950 1605 0.05 Valve 2.
AEC Assumptions Diesel Gen. (C) 1760 1650 0 05 3.
,1835 1530 0.05 Valve 970 0.5
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4.
GE Design Basis None 1610 1
l 5."
GE "Best Estimate" LPCI Inj. (B) 1400-1600
,1100-1300 0.05 Valve Calculated metal water reaction is less than 0.12% of cladding for all cases abovd AEC acceptability index is 1%.
A.
- In addition, HPCI neglected except for Case f4.
With HPCI included, intermedi:
break temperature would be approximately 100'F less and worst intermediate bre' size would be several times larger.
B.
Two core spray systems remaining.
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C.
Three' flooding pumps and three core spray pumps (150% of C.S. system) remainini 1
D.
No distinct peak.
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