ML20065B040
| ML20065B040 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/16/1983 |
| From: | Rosalyn Jones METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML20065A980 | List: |
| References | |
| ALAB-708, NUDOCS 8302220371 | |
| Download: ML20065B040 (27) | |
Text
. - _ _
February 16, 1983 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Matter of
)
)
METROPOLITAN EDISON COMPANY
)
Docket No. 50-289
)
(Restart)
(Three Mile Island Nuclear
)
Station, Unit No. 1)
)
LICENSEE'S TESTIMONY OF ROBERT C.
JONES, JR.
IN RESPONSE TO ALAB-708 ISSUE NOS. 4 THROUGH 7 (ECCS EVALUATIONS AND BOILER-CONDENSER COOLING) 8302220371 830216 PDR ADOCK T
05000289 PDR
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SUMMARY
Thic testimony responds to the Appeal Board's stated concerns with the B&W ECCS evaluations of small-break loss-of-coolant accidents and the efficacy of boiler-condenser cooling to remove decay heat at TMI-1 for those breaks for which it is predicted to occur.
The pre-TMI-2 accident analyses to demonstrate TMI-1 compliance with 10 C.F.R. 9 50.46 used the NRC approved Appendix K model and, for certain break sizes, the results of these analyses al'so exhibited the steam generator heat transfer characteristics associated with boiler-condenser cooling.
The post-TMI-2 accident analyses used the approved CRAFT 2 computer code, but modifications were made to the model to provide a more detailed examination 'of plant response under boiler-condenser conditions.
A revised B&W evaluation model, submitted to the Staff for Appendix K approval, has been used to analyze a 0.01 ft break, during which boiler-condenser cooling is predicted to occur, and an extrapolation of the results demonstrates tha; adequate core cooling is maintained.
While breaks smaller than the original spectrum (i.e., 0.04 ft2) do not need to be analyzed to demonstrate compliance with section 50.46, the response to NUREG-0737 Items II.K.3.30 and II.K.3.31 will i
provide further confirmation that the origine.1 spectrum 2
analyzed was adequate (i.e.,
that 0.07 ft is the worst case).
-i-
-The foregoing analyses demonstrate the adequacy of the boiler-condenser cooling mode to remove decay heat at TMI-1.
A heat transfer analysis of the steam generator provides yet a further. illustration of that capability.
In addition, experi-mental data is discussed which supports this conclusion from the analyses.
t
-ii-
INTRODUCTION 1
This testimony, by Robert C.
- Jones, Jr.,
Supervisory 2
ng neer, Operational Analysis Unit, Babcock & Wilcox Company, 3
is in response to Issue Nos. 4 through 7 of the Appeal Board's 4
em ran um an order of December 29, 1982 (ALAB-708).
Collec-5 ve y, those issues address the adequacy of the B&W Emergency 6
Core Cooling System (ECCS) evaluations of small-break 7
loss-of-Coolant accidents (small-break LOCAs) and the efficacy g
of boiler-condenser cooling to remove decay heat at TMI-1 for 9
10 those breaks for which it is predicted to occur.
Licensee evidence in the record which is relevant to these 11 issues, and which may provide valuable background information, 12 in ludes:
13 o
Licensee's Testimony of Robert W.
Keaten and Robert 74 C.
Jones in Response to UCS Centention Nos. 1 and 2 (Natural and Forced Circulation), ff. Tr. 4588; 15 o
Licensee's Testimony of Robert C.
Jones, Jr. and T.
g Gary Broughton in Response to UCS Contention No. 8 and ECNP Contention No. 1(e) (Additional LOCA 7
Analysis), ff. Tr. 5038; 18 o
Licensee's Testimony of Robert C.
Jones, Jr. and T.
Gary Broughton in Response to the Board Question on 19 UCS Contention 8, ff. Tr. 5039; 20 o
Licensee Exhibits 3 through 13 (small-break LOCA and other accident analyses performed before and after 23 the TMI-2 accident; small break operator guidelines).
22 23 24 25 26 ISSUE NO. 4:
Whether the modified B&W ECCS evaluation model 1
for small breaks that predicts the boiler-condenser process is an NRC spproved code under 2
Appendix K to 10 CFR Part 50 (from the staff).
3
RESPONSE
4 5
6 NRC regulations provide the definition of an ECCS eval-7 uation model.
8 An evaluation model is the calculational framework for evaluating the behavior of 9
the reactor system during a postulated loss-of-coolant accident (LOCA).
It 10 includes one cr more computer programs and all ther information necessary for 11 application of the calculational framework to a specific LOCA, such as mathematical 12 models used, assumptions included in the programs, procedure for treating the 13 program input and output information, specification of those portions of analysis y4 not included in computer programs, values of parameters, and all other information 15 necessary to specify the calculational procedure.
6 10 C.F.R.
$ 50.46(c)(2).
7 Analyses performed prior to the TMI-2 accident to demon-8 strate the conformance of TMI-1 to 10 C.F.R.
S 50.46 used the g
NRC-approved B&W ECCS evaluation model and, for certain break 2
21 sizes (e.g.,
the 0.04 ft break), the results of these 22 analyses also exhibited the steam generator heat transfer 23 characteristics associated with boiler-condenser cooling.
24 The model used for the additional small-break LOCA 25 analyses performed after the TMI-2 accident that predict the 26 boiler-condenser process technically was not the B&W ECCS evaluation model approved by the NRC pursuant to Appendix K to 1
The model used for those analyses was the approved B&W evaluation model modified only by the addition of 3
two control volumes (or nodes) tu provide a more detailed 4
examination of plant response under boiler-condenser condi-5 tions.
No changes were made, however, to the CRAFT 2 computer b
de, which is the approved Appendix K code used to predict 7
cystem response for these breaks.
8 The additional control volumes, one in each Reactor g
Coolant System loop, were included in order to explicitly 10 represent the upper head, or plenum, region of each steam g
generator.
The analytical impact of the addition of the g
ntrol volumes was to allow for a more accurate representation 13 of the formation of a steam bubble between the steam generator g
emergen y feedwater injection point and the 180* U-bend in the 15 top of eaCh RCS hot leg.
See Licensee Ex.
5, S 6.2.4.2.
It should also be noted, as I discuss more fully below in response to Issue No.
7, that the-B&W ECCS evaluation model for 8
small-break LOCAs has been further revised, in response to Item g
II.K.3.30 of NUREG-0737.
The changes made to the model include the addition of a steam generator upper head region, as discussed above, and others developed in consonance with the NRC Staff.
The revised model has been formally submitted to the NRC (see Licensee Ex.
) for review by the Staff for compliance with Appendix K to 10 C.F.R. Part 50.
26 P
ISSUE NO. 5:
1 Whether the staff has reviewed the B&W Appendix K model to determine the ability of the code to alculate the effects of small breaks, including 2
reliance upon boiler-condenser circulation (from
}'
3
RESPONSE
4 5
BY WITNESS JONES:
6 While I obviously cannot describe the scope of the Staff's 7
e re e
ey n a
se as rep r e as I indi-8 cated above the results of the analyses performed prior to the 9
TMI-2 cccident to demonstrate the conformance of TMI-1 to 10 10 C.F.R.
S 50.46, with the approved B&W Appendix K model, 11 exhibited, for certain break sizes, the steam generator heat 12 transfer characteristics associated with boiler-condenser 13 cooling.
74 The documentation of a revised B&W ECCS evaluation model, 15 submitted to the Staff in November, 1982 under NUREG-0737 Item 16 II.K.3.30 for review against Appendix K, includes the results 7
2 18 of an analysis of the 0.01 ft break, during which boiler-19 condenser cooling is predicted to occur.
See Licensee Ex.
20 at Appendix E.
21 22 23 24 25 26 ISSUE NO. 6:
Whether only breaks slightly smaller than 0.07 y
2 ft must be analyzed (from the staff).
2
RESPONSE
3 4
BY WITNESS JONES:
5 The smallest break analyzed in the demonstration, prior to 6
the TMI-2 accident, of TMI-1 conformance to 10 C.F.R. 5 50.46 7
2 was f the size 0.04 ft See Jones and Broughton, ff. Tr.
8 5038, at 12 (Table 1); Licensee Exs. 3 and 4.
Breaks smaller 9
2 10 than 0.04 ft do not need to be analyzed to demonstrate the 11 conformance of TMI-1 to section 50.46.
12 Section 50.46 establishes the criteria for an acceptable 13 emergency core cooling system.
Appendix K to 10 C.F.R. Part 50 14 sets forth the required and acceptable features of an eval-15 uation model used to show compliance with 10 C.F.R. 9 50.46.
16 ECCS cooling performance is to ".
.be calculated for a number 17 of postulated loss-of-coolant accidents of different sizes, 18 locations, and other properties sufficient to provide assurance 19 that the entire spectrum of postulated loss-of-coolant acci-20 dents is covered."
See 10 C.F.R.
$ 50.46(a)(1).
21 B&W's selection of the spectrum of small breaks to be 22 evaluated pursuant to 10 C.F.R.
S 50.46 was based on the 23 following considerations:
29 1.
A Core Flood Tank (CFT) line break, by its location, 25 severciy limits the Emergency Core Cooling Systems 26 l------
available for accident mitigation.
Considerations of y
break location and single active failure dictate that core 2
Cooling must be provided by one high pressure injection (HPI) train and one core flood tank, until the active low 4
pressure injection (LPI) train can be switched from its assumed injection into the broken CFT line and balanced 6
between the two CFT lines.
This break.is analyzed, then, 7
because it would appear to represent a limiting condition.
8 2.
A series of break sizes are evaluated wherein the conse-g quen es f the rupture are mitigated by various combina-10 tions of the three ECCS systems:
11 A.
A break is considered for which mitigation is g
Provided by the LPI, CFT and HPI systems.
13 B.
A break is considereci for which mitigation is g
supplied by only the OFT and the HPI systems.
15 C.
A break is considered for which mitigation is 6
provided solely by the HPI system.
a s ar u
m a
with the exception of the 18 Core Flood line break, between the high pressure injection g
it in the cold leg (reactor coolant pump discharge r
piping) and the inlet to the reactor vessel.
This location minimizes the amount of high pressure injection available for core cooling since a significant portion of the HPI flow can be discharged directly out the break.
In addition, breaks at low elevations within the Reactor 26 Coolant System drain the Reactor Coolant System of y
signifi antly more water than breaks at higher elevations.
2 Thus, for accidents in which the HPI or other ECCS systems 3
cannot instantaneously provide core cooling and cooling 4
must be sustained for some period of time via the initial 5
RCS inventory, that inventory is reduced in the most rapid 6
way possible.
7 3.
-Additional breaks are considered to confirm that the above 8
spectrum has indeed boundad the worst case.
That is, as 9
necessary, break sizes smaller and larger than the 10 calculated worst case are considered in order to confirm 11
~
that the most adverse core cooling situation has been 12 identified.
13 Very small breaks, i.e.,
those smaller than the smallest 14 15 break considered in the spectrum (0.04 ft2),
are not 16 evaluated because they are bounded by larger breaks for the 17 following reasons:
18 1.
Because of the internal vent valves, condensation within 19 the steam generator must occur prior to uncovering of the 20 reactor core.
At TMI-1, this occurs because the injection 21 location for emergency feedwater is near the top of the 22 steam generator.
Ultimately, the steam generator is 23 filled to 95 percent on the operating range, which assures 24 a condensing surface above the top of the core continu-25 ously.
26 __
2.
If steam condensation is occurring in the primary side of 1
the steam generator, then the RCS pressure will be reduced 2
to near the pressure of the secondary side of the steam 3
generators (approximately 1000 psi) or at a higher 4
pressur wherein the HPI flow matches the leak flow.
5 3.
The breaks evaluated in the spectrum, those with HPI 6
mitigation oniv, drain the RCS loops faster and establish 7
steam Condensation earlier than do smaller breaks.
At the g
start of the steam condensation mode, the decay heat rate 9
"#9*#
10 break.
The larger break will al.3o be losing initial RCS 11 inventory faster than the smaller break.
Thus the g
potential for core uncovery is greater for the larger 13 breaks.
g 4.
Because it has been shown by evaluation that the HPI 15 provides successful mitigation of a transient at a higher 6
decay heat rate at an earlier time, the HPI will provide successful mitigation of the transient at a lower, later decay heat rate.
Therefore, smaller breaks cannot have g
Consequences in the core region more severe than the smallest break considered in the spectrum evaluation.
Therefore, while breaks smaller than the spectrum analyzed to demonstrate compliance with 10 C.F.R. 5 50.46 may involve different system behavior (i.e.,
the repressurization cycle which is caused by the interruption of natural circulation),
26.
i y
core cooling is dependent upon maintaining core coolant inventory.
Regardless of the specific sequence of events 2
during a very-small-break LOCA, before core uncovery can occur, 3
reactor coolant pressure will decrease to a point (approxi-4 mately 1000 psig) where high pressure injection has been 5
demonstrated to provide adequate core cooling for the maximum 6
r decay heat level.
7
^ """ Y**
E"
^
~
8 the TMI-2 accident provided further confirmation of the g
validity of the above described methodology.
While these 10 evaluations were for the purpose of providing an improved 11 analytical basis for emergency operating procedures, rather 12 than to demonstrate compliance with 10 C.F.R.
$ 50.46, several 13 2
14 breaks smaller than the previously analyzed 0.04 ft break 15 2
were addressed.
Specifically, breaks of 0.005 ft and 0.01 2
ft were evaluated.
See Jones and Broughton, ff. Tr. 5038, 17 at 6-7 and 17 (Table 6).
In my opinion, the analyses for the 18 2
2 0.005 ft and 0.02 ft breaks are sufficient to g
demonstrate conformance to 10 C.F.R.
S 50.46 pursuant to Appendix K.
The results indeed showed that, compared to the l
larger break sizes, an increased margin relative to core i
uncovery existed.
The effort now underway, pursuant to 23 l
NUREG-0737 Items II.K.3.30 and II.K.3.31, to analyze small breaks with an improved Appendix K model, is aimed at providing yet further confirmation that the original spectrum of breaks l
26
_9_
analyzed was adequate to demonstrate conformance to 10 C.F.R.
y 5 50.46.
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 l
19 20 21 22 23 l
24 l
25 26 e-ISSUE NO. 7:
Confirmation (such as by means of detailed y
computational analysis or experimental testing) that boiler-condenser circulation flow will 2
transport sufficient core decay heat to the steam generators to prevent core damage (from 3
the licensee and the staff).
4
RESPONSE
5 BY WITNESS JONES:
7 For certain sized small-break LOCAs, the steam generators 8
are r.ecessary to remove a portion of the decay heat added to 9
the primary system.lf The Appeal Board has questioned the 10 adequacy of energy removal via the steam generators while 11 operating in the boiler-condenser mode of cooling.
Additional 12 analyses are presented in this testi:r.ony to demonstrate that 13 boiler-condenser heat removal at TM5-1 is sufficient to remove 14 core decay heat following a LOCA.
I have also provided a 15 discussion of the experimental data which supports this 16 conclusion from the analyses.
17 Before discussing the boiler-condenser mode of cooling, 18 however, it is necessary to discuss the relationship between 19 energy removal from the fuel rods (core cooling) and energy 20 removal from the reactor coolant system (RCS).
To ensure 21 adequate core cooling during a small-break LOCA, it is 22 23 1/
The discussion that follows assumes the availability of emergency feedwater and one HPI train.
Steam generator heat 24 removal is not necessary if two HPI pumps are available.
See Jones and Broughton, ff. Tr. 5038.
26 1
necessary to maintain a two-phase level within the reactor vessel which is at or near the top of the core.
In this 2
manner, the core decay heat which is being generated can be 3
removed from the fuel rods by pool boiling or, if the core is 4
slightly uncovered, by forced convection to superheated steam.
5 The HPI system has been designed to provide the necessary fluid 6
makeup to the RCS to ensure adequate core heat removal.
7 Decay heat removal from the RCS can be accomplished in 8
oi several ways, e.g.,
by break flow, steam generator heat
-t rem val, r c mbinations thereof.
During a small-break LOCA, 10 the decay heat removal is important in that it determines the yy 12 system pressure and, hence, the HPI flow being provided.
Therefore, to demonstrate core cooling, it is only necessary to g
g show that sufficient decay heat removal is provided, prior to re un very, to allow the HPI system to replace the inventury 15 being boiled by core decay heat removal.
In this manner, level 6
in the core can be maintained above the top of active fuel g
18 19 For break sizes smaller than 0.02 ft decay heat 20 removal from the RCS is accomplished by a combination of the 21 break flow and the steam generators.
See Keaten and Jones, ff.
22 Tr. 4588, at 7.
If the break sizes are smaller than 0.005 2
23 ft the HPI system can compensate for the break flow and 24 maintain the primary coolant loops essentially full of liquid 25 such that natural circulation is not interrupted.
26 -
2 1
Assuming a break size between 0.005 and 0.02 ft the 2
HPI flow is unable to compensate for che leak flow and the RCS 3
will saturate.
Steam pockets will eventually form and grow to 4
a volume sufficient to fill the 180 inverted U-bends at the 5
top of both hot legs.
This will result in an interruption of 6
natural circulation.
The loss of natural circulation leads to 7
a loss of heat removal via the steam generators and the system 8
will pressurize.
See Jones and Broughton, ff. Tr. 5038, at 9
6-7; Keaten and Jones, ff. Tr. 4588, at 7.
10 An the RCS continues to lose inventory, a condensing 11 surface will be exposed in the steam generators.
This vill 12 estaclish the boiler-condenser mode of heat removal.
This mode 13 of heat removal will terminate the pressure increase and 14 control RCS pressure at a value sufficient to assure adequate 15 HPI flow for core cooling.
See Jones and Broughton, ff. Tr.
16 5038, at 6-7.
17 Small-break LOCA analyses have been performed which 18 demonstrate the adequacy of this cooling mode.
These are 19 documented in Licensee's Exhibit 5.
Those analyses were 20 performed utilizing the presently approved CRAFT 2 code.
21 Comparison of the steam generator heat removal rates calculated 22 in those analyses to that which would be obtained by using the 23 theoretical formulations in the new model show reasonable 24 agreement.
That is, an approximate three-foot adjustment in l
25 the condensing length would yield the same heat transfer.
This l
26 l
1 small loss of inventory, approximately ten percent of the 2
available inventory above the top of the core, would not affect 3
core cooling.
4 Since the analyses in Licensee's Exhibit 5, the B&W ECCS 5
evaluation model and the CRAFT 2 code have undergone modifica-6 tion in response to II.K.3.30 of NUREG-0737.
The revised 7
evaluation model and CRAFT 2 code have been submitted to the NRC 8
for review.
9 Within the modified CRAFT 2 code, an upgraded steam 10 generator mcdel has been incorporated which inc).udes heat 11 transfer correlatione specifically oriented to the boi'.er-12 condenser mode of cooling.
A new O.01 ft break analysis 13 has been performed using the revised ccde and is documented in 14 BAW-10154.
See Licensee Ex.
Appendix E.
Extrapolation 15 of the results demonstrate that adequate core cooling is 16 maintained for breaks of the size for which boiler-condenser 17 cooling is predicted to occur.
18 The capability of'the steam generator to remove sufficient 19 core decay heat to assure adequate core cooling via the HPI 20 system during a small break LOCA is further illustrated by the 21 analysis described below.
As stated previously, adequate core 22 cooling is assured if the core is continuously covered by a 23 two-phase mixture.
Maintenance of this condition is assured if 24 the HPI flow provided to the system is sufficient to match or 25 exceed the inventory b. oiled off from core decay heat removal.
26 1
Because the HPI flow varies with system pressure, the time 2
at which the injected flow and core boiling match will be a 3
function of the system pressure.
The pressure / time relation-4 ship for this matchup is illustrated on Figure 1.
Thus, the 5
significant question is whether the boiler-condenser mode will 6
assure a pressure / time relationship, before the core becomes 7
uncovered, to yield adequate HPI to keep the core covered.
8 A heat transfer analysis of the steam generator, while 9
perating in the boiler-condenser nede, was performed to 10 develop the pressure / time relationship.
Prior to any possible 11 uncovering of the core, the full condensing surface of the 12 steam generator will be exposed.
Using this surface area, an 13 analysis was performed to determine the RCS temperature, and 14 hence pressure, necessary to condense all the steam being 15 generated as a result of core decay heat removal as a function f time.
It should be noted that since none of the generated 16 17 steam is assumed to be removed via the break, this analysis w uld 18 verpredict the RCS pressure that could exist just prior 19 to possible core uncovery.
Figures 2 and 3 show the results of 20 the steam generator heat removal analysis for cooling on the 21 steam generator level (at 95 percent on the operating range) and the emergency feedwater spray, respectively.
22
-[ Combining the results of the HPI cooling and steam 23 generator heat removal analyses, as illustrated in Figure 4, it g
is seen that boiler-condenser heat removal will provide 25 26 i
l sufficient pressure control to result in HPI flows necessary to 2
assure adequate core cooling after 1650 seconds.
The next 3
subject of the analysis, then, is to determine whether the core 4
is predicted to become uncovered prior to this time.
5 Several small break LOCA analyses have been performed 6
which indicate that the core could not become uncovered prior 7
to 1650 seconds for the break sizes of interest.
In Licensee's 8
Exhibits 3 and 4, which are the section 50.46/ Appendix K lanalysesforTMI-1, 2
9 it can be seen that the 0.04 ft break 10 ! ranches its minivum system inventory at 3000 seconds.
No 11 uncovering of the core is calculated for this break.
Since 12 smaller breaks would lose inventory at a slower rate, the 0.04 13 9
, ft" break would bound the results.
14 2
In addition, the analyses of the 0.01 ft break 15 (documented in Licensee's Exhibit (BAW-10154), show that 16 the boiler-condenser mode of ecoling is calculated to occur at 17 approximately 1500 seconds.
At this time, there is a substan-18 tial quantity of liquid (105,600 lb or 2440 ft3) remaining 19 above the top of the core.
This inventory would have to be 20 lost through the break prior to the core uncovering.
Even if 21 an RCS pressure of 2500 psi was assumed, which is well above 22 the 1800 psi pressure calculated for this time, this inventory 23 could not be lost prior to 1650 seconds.
24 Based on this analysis, it is clear that uncovering of the 25 core would not occur prior to 1650 seconds for the break sice 26 l
i 1
range'for which boiler-condenser heat removal is necessary.
2 Since the boiler-condenser' cooling mode assures adequate 3
pressure control after this time to enable the HPI to match or 4
exceed the core boil-off, adequate core cooling is assured.
5 Turning to the Appeal Board's interest in experimental 6
testing of the boiler-condenser ~ mode of heat removal, it should 7
be recognized that the actual heat transfer-mechanisms are well 8
understcod.
Within the primary system steam is condensed on 9
the inside wall of the cooled steam generator.
The heat then 10 flows through the tubes, via con.Netion, and is tranaferred to 11 the secondary side fluid.
Two possible mechanisme exist for 12 the secondary side heat transfer.
These are by scol boiling on 13 the immersed steam generator tubes and/or cooling-by the 14 emergency feedwater which is sprayed directly on the steam 15 generator tubes.
16 There are several data sources available, or planned, 17 which demonstrate the capability of the steam generator to rem ve heat in a boiler-condenser mode.
First, there is the 18 TMI-2 accident'itself.
After all of the reactor cool' ant pumps 19 had been tripped at 100 minutes, filling of the steam generator 20 by emergency feedwater commenced.
During the fill period, heat 21 rem val fr m the RCS occurred which controlled the primary 22 system pressure within 100 psi of the secondary side pressure.
23 The only explanation for the pressure curves. tracking together 24 is the effect of boiler-condenser cooling in removing decay 25 26 heat.
See UCS Ex. 1 (minutes 100 to 125).
If the HPI system 1
2 had been actuated and maintained at this time, adequate 3
inventory would have been maintained to prevent core damage.
4 Thus, the TMI-2 accident did not demonstrate an inadequacy of 5
RCS heat removal
(_i.e.,
an inadequacy of Soiler-condenser c
ling), but rather showed the importance of maintaining 6
7 adequate core inventory via the HPI.
8 Tests have also been run at the Alliance Research Center 9
( ARC) which examined condensation phenomena in a high pressure 10 facility.
In these tests, a single steam generator tube was 11 tested by exposing a condensing surfac.e by adjusting water level n the inside surface of.the tube.
Then, by varying 12 13 steam flow to the test section, temperature measuraments were 14 taken in order to determine the heat transfer coefficient.
The calculated coefficients for these tests have confirmed the 15 16 conservatism of the heat transfer model employed in the 17 upgraded CRAFT 2 code.
18 In the future, additional experimental data on the boiler-19 e ndenser mode of cooling and small break LOCA response will be devel ped at ARC.
At present, an integrated systems test 20 fa ility at ARC (GERDA) is being tested.
It is a scaled 21 single-loop, full height, full pressure test facility of a B&W 22 NSS and is of similar size to Semiscale.
This facility was 23 developed for the BBR company in Germany in order to examine 24 small break LOCA phenomena.
The data from this facility is 25 expe ted to be available in mid-1983.
26 1
The B&W Owner's Group, in conjunction with the NRC, is 2
presently exploring a two-loop facility to further examine 3
plant response to small break LOCA and other transients.
This 4
data will be used to confirm the adequacy of the computer models.
Through the computer codes, this data will then 5
6 enhance the understanding of plant response for improved 7
perator training and procedures.
Data from this facility is 8
projected to be available in mid-1985.
9 In summary, the boiler-condenser mode of cooling is relied 10 upon for heat removal during certain sized small break LOCAs.
e 11 The basic heat removal prccesses are well understood and have 12 been successfully applied in other engineering applications.
13 The ability of the TMI-1 steam generator to remove core decay 14 heat has been demonstrated as cufficient to provide adequate 15 core cooling.
Thus, while there are presently plans to obtain 16 additional experimental data for the purposes of improved 17 understanding of plant response and for code benchmarking, 18 peration of TMI-1 prior to receipt of this data will not 19 endanger the public health and safety.
20 21 22 23 24 25 26 i
Figure 1 HPI MATCHUP WITH CORE DECAY HEAT 1
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ROBERT C. JONES, JR.
Business Address:
Babcock & Wilcox Company Nuclear Power 3eneration Division Post Office Box 1260 Lynchburg, Virginia 24505 Education:
B.S.,
Nuclear Engineering, Pennsylvania State University, 1971.
Post Graduate Courses in Physics, Lynchburg College.
Experience:
July 1982 to present:
Supervisory Engineer, Operational Analysis Unit, B&W. Responsible for the performance of plant transient analyses and analyses used in the development of operator guidelines.
During this period, has continued as Project Engineer for B&W analyses performed in response to NUREG-0737 Item II.K.3.30.
June 1975 to July 1982:
Acting Supervisory Engineer and Supervisory Engineer, ECCS Analysis Unit, B&W.
Responsible for calculation of large and small break ECCS evaluations, evaluations of mass and energy releases to the containment during a LOCA, and performance of best estimate pretest predictions of LOCA experiments as part of the NRC Standard Problem Program.
Involved in the pre-paration of operator guidelines for small-break LOCA's and inadequate core cooling mitigation.
June 1971 to June 1975:
Engineer, ECCS Analysis Unit, B&W.
Performed both large and small break ECCS analyses under both the Interim Acceptance Criteria and the present Acceptance Criteria of 10 CFR 50.46 and Appendix K.
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