ML19323H953
| ML19323H953 | |
| Person / Time | |
|---|---|
| Issue date: | 04/30/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1718, NUDOCS 8006170338 | |
| Download: ML19323H953 (39) | |
Text
.
PB
[ 6f.f-/ W [
J Y. '
.11 J p
f"
- . J I
6 E.
DATE ISSUED: 4/30/80 f{
5 g OIU MEETING MINUTES OF THE C0t1BINED ACRS ECCS/ REACTOR FUEL SUBC0!1f11TTEE t1EETING FEBRUARY 14, 1980 WASHINGTON, D.C.
On February 14, 1980 the Combined ACRS ECCS/ Reactor Fuel Subcommittee met in Washington, D.C., to di.scuss the effects of externally mounted thermo-couples on the LOFT fuel rods, and proposed changes in the ECCS fuel cladding rupture models for Appendix K to 10 CFR 50.46, and the effect of these changes on vendor 's evaluation models. The notice of the meeting appeamd in the Federal Register on January 30, 1980.
There were no requests for oral or written statements from members of the public and none were made at the meeting. Attachment A is a copy of the meeting agenda.
The attendees list is Attachment B.
Attachment C is a tentative schedule of presentations for the meeting. Selected slides and handouts from the meeting are Attachment D to these minutes.
A complete set of slides and handouts is attached to the office copy of these minutes.
OPE'(_S_E_SSION(8:30 am - 5:35 pm) INTRODUCTION Dr. Plesset, acting as Chairman of the Combined Subcommittee, called the meeting to order at 8:35 am. The Chairman explained the purpose of the meeting and the procedures for conducting the meeting, pointing out that Mr. Paul Boehnert was the Designated Federal Employee in attendance.
Dr. Flesset begin the meeting by commenting on the so called " fin effect" seen with the externally mounted LOFT thermocouples.
The Chairman said that tests c:nducted both in the US and ov'erseas show that externally mounted thennocouples do effect the test results.
Dr. Plesset expressed concern that NRC research did not seem to agree that the fin effect is a problem.
Dr. Catton also expressed i
concern regardina the impact of the fin effect and suggested NRC address the use of LOFT results vi:;-a-vis code development.
Dr. Plesset also expressed con-
~
cern that anomalous' data from LOFT may be used to modify such predictive codes as RELAP or TRAC.
8006170338
1 ECC5/ Reactor Fuel h.,
2/14/80
,%~
THE SURFACE-HOUNTED OR EXTERNAL THERM 0 COUPLE PROBLEM -
LOFT PROGRAM MANAGER)
Dr. McPherson said he would discuss the problem of using externally mounted thermocouples, highlight results of recent PBF tests conducted to attempt to quantify the fin effect, review out-of-pile tests conducted both in the US and In overseas, and finally state conclusions regarding what is known to date.
response to Dr. Plesset's concern noted above, Dr. McPherson said that none of the LOFT data has been used to modify such codes as TRAC or RELAP.
Dr. McPherson described the thermocouples used in LOFT and how they are mounted He also briefly reviewed results of the LOFT L2-2, and on selected fuel rods.
L2-3 tests that show that the core experienced an early quench prior to refill.
Dr. McPherson discussed the PBF-LOCA thermocouple ef fects tests (TC-1).
The cbjectives of thd tests were to determine if external thermocouples influence fuel rod behavior during a LOCA, and determine whether the thermocouples accurately The tests simulated the blowdown and rewet that measured cladding temperature.
was seen in the LOFT tests.
Figures D l-3 detail the test series, the test geometry, and the test train used. Selected test rods contained internally and/or externally mounted thermocouples. Dr. McPherson said that the tests studied such affects as delay of CHF, reduction of fuel rod temperature, and earlier than expected reflooding of the core.
Turning to the test results Dr. McPherson noted the following points:
'The tests showed a definite fin effect during film boiling (Figure D-4).
Dr. McPherson-said this result and others can be used to estimate the magnitude of the effect.
'Dr. McPherson showed a data plot for the TC-1 blowdown peak temperatures.
A best fit curve-(Figure D-5 curve 1) was drawn through the data for the rods with internal thermocouples. The slope of the curve was given as Dr. McPherson said that this meant that delaying DNB will 60 K per/sec.
raise the PCT by.60 K per/sec. Based on data results from the external thermocouple rods, Dr. McPherson suggested one could draw a parallel
~
curve (labeled McPherson curve on Figure D-5) that illustrates the fin
ECCS/ Reactor Fuel Mtg 2/14/80 effect.
Dr. Lienhard questioned the positioning of this curve given the data scatter, and the fact that there were no data points on the upper portion of the curve.
'Concerning the reflood portion of the TC-1 experiment Dr. McPherson noted that all the rod temperatures begin to decrease about the same time, whether or not they had external thermocouples (Figure D-6).
Dr. McPherson discussed the conclusions derived from the TC-1 experiments. He said that NRC believes the external thermocouples influenced the LOFT L2-2 and L2-3 experiments in the following manner:
(1) PCT decreased by approximately 30 K; (2) fin effect during blowdown is variable (0-90 K) and dependent upon the velocity of the blowdown fluid; (3) rods with and without external thermo-couples saw temperature decreases at the same time as reflood commenced; and (4) there was a faster quench during reflood.
Dr. McPherson said that another set of tests at PBF (TC-2) are scheduled to provide additional information on the fin effect, especially during the blow-down-quench period.
In response to a question from Dr. Plesset, Dr. Tong said that it is hoped that a correction factor can be determinted for the fin effect, based on the results from the TC-1 and TC-2 tests.
There was Subcommittee discussion on the use of various types of internally mounted thermocouples and their impact on the test results.
Dr. McPherson also discussed results of the PBF LOFT Lead Rod (LLR) tests. The tests were performed to provide information on the expected response of the LOFT fuel, and to determine if any special fuel pre-conditioning was necessary. Secon-dary test objectives included determining the reliability of the LOFT thermocouples.
During the blowdown phase of the test, it was noted that the surface thermocouples delayed CHF in the upper portion of the fuel rods.
It was also noted that DNB was cccurring at lower elevations on the rods than the thermocouples, resulting in higher temperatures at lower elevations.
Figure D-7 provides details of the LLR t:st results. Dr. McPherson also said that during reflood, very little or no thermocouple effects were seen for either low, moderate, or high floooing rates.
6
1
. 2/14/80 ECCS/ Reactor Fuel Mtg Dr. McPherson concluded that he believes the LLR tests support the conclusions
~ drawn from the respits of the TC-1 tests.
The results of out-of-pile electrically heated rod tests conducted in the U.S.
(LOFT Technical Support Facility, UCLA) and overseas (Germany, England, Switzer-land, Norway, and Holland) were discussed.
Dr. McPherson said that there are important differences between electrically heated rods and nuclear heated rods, and one cannot directly apply the results of electrically heated rod tests to a in-reactor situation. Generally speaking, electrically heated rods with external thermocouples exhibit faster quench times than is seen with nuclear rr e s with external thermocouples.
shown today, Dr. McPherson concluded his presentation by noting that the results plus results from future tests, should resolve the fin-effect thermocouple pro-blem within one year.
CLADDING SWELLING AND RUPTURE MODELS FOR LOCA ANALYSIS - R. MEYER, D. POWERS, J. ROSENTHAL (NRC)
Dr. Meyer began the NRC discussion of the new proposed cladding swelling and rupture models for LOCA analysis as detailed in the draft NUREG-0630 (Cladding Swelling and Rupture Models for LOCA Analysis). Dr. Meyer noted that NRC has l
received comments from the reactor vendors, and organizations such as EPRI, as j
well as comments from representatives of the United Kingdom, Germany, and Japan
'on,the draft NUREG. Dr. Meyer also noted that Mr. A. Mann from the Springfields Laboratories in the UK would make a presentation at today's meeting.
Dr. Meyer said that the cladding behavior correlations of interest for this discussion included rupture temperature versus engineering hoop stress, burst strain versus rupture temperature, and PWR assembly flow blockage versus rupture temperature. Figure D-8 shows these three parameters in relationship t'o all the important parameters used in a LOCA calculation mandated by 10 CFR 50.46 and Appendix K.
As a result of review of the CE flow blockage model in 1977, NRC decided that review of all the vendor's models was necessary, given the disparity seen in the burst strain and flow blockage models (Figures D 9-10). NRC then decided that it should define acceptabl'e models in these areas.
In response to a
+
ECCS/ Reactor. Fuel Mtg 2/14/80 question from Mr. Etherington on the importance of accounting for the cladding strain rate, Dr. Meyer said that with the exception of the W small break model, all the other vendors do not account for the strain rate parameter. He said that the new NRC cladding models would account for strain rate.
Mr. J. Rosenthal (D0R) discussed the interim actions taken on operating reactors for this concern. He described the results of actions taken in early November 1979 when it was thought tiy the NRC that there were potential deficiencies in the vendor's ECCS evaluation models.
It was concluded that no safety problem existed, however "no safety problem" was defined to mean that:
(1) peak clad temperature (PCT) predictions were insensitive to fuel clad models, or (2) existing models were adequate over the narrow range of applicability, or (3) sufficient margin was available with existing models, or (4) off-setting credits existed for other model changes which were under NRC review. Mr. Rosenthal noted that Westinghouse found it necessary to make use of off-setting conserva-U tisms to overcome a substantive (~2-700 F) increase in FCT.
Dr. D. Powers (NRC-DSS) discussed the rupture temperature, burst strain, and flow blockage correlations developed by NRC and described in NUREG-0630. The data base used for developing these correlations was restricted to tests in dry steam and which made use of internally heated rods. The data was obtained from tests conducted at Oak Ridge, Battelle Columbus Laboratories,
the KFK Facility in Germany, and the Japanese Atomic Energy Research Institute.
Both in-and out-of-pile tests were included in the data base. Dr. Ibwers noted that the above correlations represent NRC's best estimate of the sub-ject models.
Dr. Powers discussed the rupture temperature correlation used in the NRC model (Figure D-11). He noted that~ this curve is based on the above discussed data and
)
the data shows a strong heating ramp rate dependence.
For the purposes of the j
NUREG report, the NRC has defined a slow ramp rate as 310 C/second and the fast ramp rate was defined as > 2C/second.
The slow-ramp and fast-ramp burst strain curves developed by NRC were discussed (F;9]res D-12-13). The development of the burst strain curves was based in part on ed work of Kassner and Chung conducted at ANL (Figure D-14).
I
, 2/14/80 ECCS/ Reactor Fuel Mtg Dr. Fbwers noted that the slow and fast-ramp burst strain curves have been modified from the initially developed correlation, largely at the recommenda-tion of Dr. Chapman from ORNL (Figures D-14A-15).
There was considerable discussion over the applicability of the data used to develop the high temperature portion of the fast-ramp burst strain curve (Figure 0-15-arrow).
These data were taken from the Oak Ridge tests; however, in response to a question from Dr. Shewmon,' Dr. Chapman of OfNL expressed doubt that this data should be characterized as fast-ramp data, due to problems encountered during the experiment.
Dr. Shewmon observed that there appears to be a doubious basis for the development of that portion of the curve.
The derivation of the flow blockage model was described by Dr. Fbwers (Figure D-16 ). He said the NRC blockage model is expressed as a function of the cladding rupture temperature. Figures D-17 and D-18 show the flow blockage curves derived by NRC along with the appl; cable data. Figures D-19 and D-20 show the differences between the draft and final flow 51ockage correlations.
There was extensive Subcommittee discussion centering on the use and interpretation of the data as well as the assumptions that went into the development of the above curves.
Dr. Meyer discussed the proposed schedule for implementing the revised fuel cladding models (Figures D-20A-21).
He stated that NRC would like a'n ACRS letter on the NUREG report at the March 1980 meeting. NRC would than issue a final version of the NUREG report around April 1,1980, along with requirements for vendor reanalysis of their ECCS models.
Dr. Shewmon expressed concern over what he felt was excess,i.ve conservatisms on the Staff's part in developing the above curves, and the rel'ation of this infonnation to what happens on a realistic basis given a LOCA in a power reactor. He also said He asked if he felt uneasy with the NRC interpretation of the test data.
tests planned by NRC research at the NRU and LOFT facilities, as well as overseas tests would relate to the infonnation being considered today.
Dr. Meyers replied that he believes that the work documented in NUREG-0630 represents a significant improvement in the cladding models over the 9
-~-
+-
ECCS/ Reactor Fuel Mtg 2/14/80 situation that has existed since 1974.
Dr. Rosztoczy noted that NRC hopes to avoid changing the curves for a long time.
Dr. Rosztoczy also estimated that the changes proposed in 0630 will cost the vendors about $10 million to implement.
COMMENTS BY M. L.
PICKLESIMER - FUEL BEHAVIOR RESEARCH BRN4CH - NRC Dr. Picklesimer gave a brie.f presentation commenting on the NUREG report.
He said he objects to the use of burst strain curves to determine flow blockage, since it leads to an overally conservative situation. He also noted however that at this time he has no alternative to the method used by the Staff, principally because he has not had time to work on this prcblem due to the TMI-2 accident.
Dr. Meyer said NRC has only used brust strain to develop an average strain which in turn is converted to a value for flow blockage.
COMMENTS BY A. MNi'i - SPRINGFIELDS NUCLEAR POWER DEVELOPMENT LABORATORY Mr. A. Mann from the Springfields Nuclear Power Development Laboratories of the UKAEA di. cussed the position of the UKAEA concerning clad deformation following a LOCA and the future work needed to clarify present uncertainties in this area. He noted that the amount of cladding strain is determined by temperature, the time the clad is at temperature, and circumferential temperature variation of the clad, the last probably being the most important parameter.
Addressing the potential problem of co-planar blockage, Mr. Mann noted that tne location of the defornation depends primarily on the temperature distribution of the cladding. The temperature distribution in turn depends on the axial varia-tion in power of decay heat iq the rod and the heat transfer at the cladding surface.
If these two parameters are similar in adjacent rods, deformation is likely to be co-planar.
l The UKAEA believes 'that a predictive code is needed that can successfully model the interaction of the thermal-hydraulics and associated clad deformation para-meters.
Further experiments should focus on such parameters as thermal-hydraulics (dryout, rewet, and heat transfer during reflood), clad deformation, and compari-b
ECCS/ Reactor Fuel Mtg 2/14/80 son of in-pile and out-pile experiments.
In response to a question from Dr. Shewmon, Mr. Mann stated that he believes the NRC may have underestimated the degree of blockage, given a worst-case LOCA, i.e. cladding swell could be co-planar.
However, Dr. Mann went on to say that on a judgment basis, he feels the draft report is probably conservative but the problem, as he sees it, is proving it is conservative.
COMMENTS BY T. KASSNER - ANL Dr. Tom Kassner from ArgonYie National Laboratory provided a presentation that discussed the relationships among various parameters used to develop cladding embrittlement criteria.
The central theme of Kassner's r esentation was that the new NRC flow blockage curves should be evaluated vis-a-vis the cladding oxidation requirements of Appendix K to determine the overall effect of clad
~
swe'.1 and rupture in a LOCA situation. He indicated that substantial wall thinning could le'ad to violating the 17% oxidation limit specified in Appendix K 0
at PCTs well below the 2200 F limit. NUREG-0630 has not considered this.
WESTINGHOUSE COMMENTS ON THE NRC FUEL R0D MODELS - D. BURMAN - W Mr. Dennis Burman provided Westinghouse comments on the NRC fuel rod models; he commented on the burst temperature, burst strain, and flow blockage correla-tions proposed by the NRC Staff.
Regarding the burst temperature correlation, Mr. Bruman said that Westinghouse agrees with the NRC that there is a heat-up rate dependence for zircaloy cladding, but that the Westinghouse model accounts for known biases in burst temperature measurements. Commenting on the burst strain correlation, Mr. Burman showed slides that he said indicated the Westinghouse data envelopes pertinent data from other sources. He said the NRC burst strain curves are upper bound curves, not best estimate as stated in NUREG-0630.
In describing the flow blockage correlations, Mr. Burman referenced some Japanese multi-rod burst tests that resulted in a large degree of co-planar bicckage (Figures D-22-23). Mr. Burman said he believed this co-planar block-age was the result of the tungsten wire electrical heating element used (Figure D-24). - (Note: Dr. Kawasaki was contacted by NRC concerning Mr. Burman state-ment regarding cause of the co-planar blockage.
Dr. Kawasaki said that he does not believe the heater design was the cause of the co-planar blockage.
i t
L
?
ECCS/ Reactor Fuel Mtg 2/M/80 Rather it was probably due to effect of the steam flow and/or the grid spacers used.in the experiment.) Westinghouse believes their flow blockage model is sufficiently conservative.
In conclusion, Mr. Burman stated that the preliminary nature of current data should preclude development of a new cladding model at this time, and noted that there are several tests scheduled for the near future which would provide data for development of more definitive models. He also said that tests in Germany show that high aegrees of blockage (90%) do not adversely impact PCT, thus there is no apparent safety issue, and therefore there is no need for new models at this time.
COMMDITS BY R. CHAPMAN - ORNL Dr. Chapman from ORNL~ briefly discussed the use of data from his single-rod ant multi-rod burst test. He detailed information he had noted earlier in the day; that is, some of the high temperature (beta-range) burst-strain data should not be ch'aracterized as fast-ramp data. He noted that other beta-range tests may suffer from similar problems. Mr. Chapman also said that his tests j
show a clear heat-up rate effect.
(W stated that the heat-up rate effect was minor.)
Dr. Meyer made some summarizing remarks. He believes that NRC and Westinghouse are in agreement concerning the use of the strain data base and that the W flow blockage model is in fairly good agreement with the NRC model.
Dr. Meyer also noted however that he believes Westinghouse mischaracterized the NRC blockage model concerning the consideration of average versus maximum flow blockage in a bundle. Dr. Meyer also requested explicit ACRS comments Cn the adoption of the new NRC models, in particular whether or not the Committee finds the models acceptable and what should be done if they do not.
. Dr. Shewmon comented that the NRC should present infonnation on the NUREG report to the full Comittee in March, but noted that one of the centr'al questions in'his inind was whether it.is better to proceed now, as the Staff
.is proposing, or wait for a year until additional test data has been generated.
The Chairman also expressed concern over how well the Staff has adequately
LECCS/ Reactor Fuel Mtg - 2/14/80 7
allowed for what might take place in the core of a power reactor. Dr. Shewmon suggested that the $taff address the question of how serious is the impact on
~
plant safety of waiting an additional year for the new test data noted above.
The meeting was adjourned at 5:30 p.m.
NOTE: Additional meeting details can be obtained from a transcript located in the NRC Public-Document Room, at 1717 H Street, N.W., Washington, D.C., or can be obtained from International Verbatim Reporters, Inc.,
499 South Capitol Street, S.W., Suite 107, Washington, D.C. 20002.
6 9
-r e -
r
==
-5 Federal Registee / WL 4S. N3. 6 / Wedn:sday. J:nuiry 30, 1980 / Notice?
HUCLEAh REGUt.ATORY Westin'ghouse. and other interested COMMISSION persons regarding: (1) proposed changes la the fuel clad rupture modela for Advisory Committee on Reactor Appendix K to 10 CFR 50.54 and the
$sfeguarets..%bcommdtees on eHect of these changes on vendor Emergency Core Cooling Sys: ems and evaluation models. (:) the effects of Reactor Fasels; Meeting exteruslly mounted thermocouples on De ACRS Subcommittees on LOFT fuel. (3) the results of the L3.-1 knergency Core Cooling Systems and Test, and (4) the analysis of:=a:1 break Reactor Fuela will hold a joint meetmg LOCAs in Westingbouse UHI reactors.
h addjtion. It may be necessary for en February 14.1980 in Room 1048.17:7 the Subcorhmittee to hold one or more H St NW., Washington. DC 20555.
Notice of this meeting was pnMinheid closed sessions.for the purpose of,'
explanng matters involvir g proprietary ~
Jencary 22.1380.
Inidrmation.f have determined. in As accordance with the procedures setlined in the Federal Register on accordance with Subsection 1Cfd) of the v
October 1.1979. (44 FR 55408). oral or Federal Advisory Committee Act(Pub.
. written statements may be presented by -
L 9:-483). that. should such sessions be members of the public, recordings will required. it is necessary to close these l
)
be permitted only dunng those portions sessions to protect preprietary of the meeting when a transcript is being information. See 5 U.S.C. 55;h(c)(4).
kept. and questions may be asked only earther infonnation regarding topics by members of the Subcommittee.its in be discussed. whether the meeting consuhants, and Staff. Persons desiring h=, been cancz!!ed or rescheduled, the to make oral statements should nonfy Chcirman's ruling on requests for the the Designated Federal Employee as far
- apportunity to present oral statements b' advance as practicable so that and the time allotted therefor can be appropriate arrangements can be made obtained by a prepaid telephor.e call to to allow the necessary time duruu; the the coSnizant Designated Federal meeting for su:h statements. -
Employee.Dr. Andrew L Bates
- The agenda for subject meeting shall (telephone 2tc/s34-3:s7) between a:15 be as fol1ows: Thursdcy, Februar7 u.
s.m. and $20 p.m EST.
2stti, 2J0 a.mi. wrtil ths conclusica cf Background information concerning items to be discussed at this meeting Aus!.,ess each day..
can be found in docments on file and De Submmmittee may meetin tsecutive Session, with any ofits evallable for public inspection at the NRC Pub!fe Docurnent Room.1717 H ususultants w' o may be present, to Street NW, Washington,DC20555.
n
.xplore and exchange their preliminary opinions regarding matters which should Deted:!anmary se.inen. "
joka C. Hoyle. -
be considered during the meeting.
us,,,7 cooiyijdy@g og;,',
.-...v.-
At the conclusion of the Execunve Session, the Subcommittee will hear-
,,u w o,,
presentations by a:d hold discussions
. n e,eg -
with representatives of the NRC Staff.
- e a.
ATTACHMENT A
i
~
MEETING OF THE C0 MINED ACRS ECCS/ REACTOR FUEL SUBCOMMITTEE MEETING FEBRUARY 14, 1980 WASHINGTON, D.C.
ATTENDEES LIST _
us NRC ACRS H. Sullivan M. Plesset, Chairman, EECS G. McPherson P. Shewmon, Chairman, Reactor Fuel R. Landry H. Etherington, Member L. S. Tong A. Acosta, Consultant M. L. Picklesimer
- 1. Catton, Consultant P. S. Anderson, Consultant J. Lienhard, Consultant F. Nichols, Consultant ARGONNE NATIONAL LAB, Y. Chen, Special Consultant P. Boehnert, Staff
- T. F. Kassner cDesignated Federal Employee DUKE POWER CO EXXON NUCLEAR S. T. ROS E G. Owsley WESTINGHOUSE COMBUSTION ENGINEERING D. L. Burman S. D. Kopelic G. Menzel V. J. Esposito E. F. Jageler J. M. Cicerchia BABCOCK & WILCOX VEPC0 H. A. Bailey R., P. Wolfhope UK ATOMIC ENERGY AUTHORITY C. A. Mann UK NUCLEAR INSTALLATIONS INSPECTORATE
~
L. G. Willians YMiKEE ATOMIC ELEC K. E. St. John 0AK RIDGE NATIONAL LAB l
R. H. Chapman ATTACHMD1T B
- U Sch:dulo far ACRS ECCS/Recetar Tuels Subcommittso Meeting F;bruary 14, 1980 8:30-8:d5a.m.
Opening Comments - Executive Session -
M.-Flesset/P. Shewman S:45 - 9:30 a.m.
Raview of PET Tests TC-1 and the Effects of Fin Thermocouple an IDFT Fuel Thermal Performance -
- [,,-
D. McPhearson i
~
9:30 - 10:30 a.m.
EI. Eeview of LOTT, L3-1 Test
- 1. Landry 10:30 - 10:40 a.m.
Breait L
Enview of 10ET L3-2 Test
- 1. Landry 10:40,11:30 a.m.
.g
.y Baview of Proposed NRC Fuel Clad
- g.,.
- C..
?
Swelling and Rupture Models 21:30 - 11:40 a.m.
c) Background and Status - 1. Meyer.
11:40 - 11:50 a.m.
~ In'terin Actions on Operating Reactors -
b)
- .?
. J. Rosenthal 11:50 - 12:30 p.m.
s c)' Description of Clad Mcdels - D. Powers
' ~
12:30 - 1:30 p.m. '
.r-Lunch
- t..
1:30 - 1:45 p.m..
- 6..
d) Flan for final Resolution - 1. Meyer
? -
1:45 - 1:50 p.m.
.M. Picklesimer Copments on Fuel Model Discussion 1:50 - 2:00 p.m.
' '.d.,. i
.
- G)
Comments on NRC Fuel Model s,.
~ 2:00 -~ 2:20 p.m.
- 1) United Kingdom - A. Mann 2:20 - 2:40 p.m.
Discussion r..
2:40 - 3:00 p.m.
2)
T. Kassner 3:00 - 3:20 p.m.
Discussion
.,3:20.-
3:40 p.m.
- 3) Westinghouse - D. Burman 3:40 - 4:00 p.m.
General Discussion 4:00 p.m.
~
Adjourn 9
Il i!
- 7. ~
t.-
piracHeter c
?
,..-,.,,.,.n._,__,e-v.,.,.n--,,
,-.--..-c.,,
.%l a
q y
'M TC-1 Test Series I
Reflood Rate
_r-A _
Blowdown System I
cm/s Test Operation 4
1 2 second slug l
4 I
2 4 second slug i
l~
3 6 second slug
~
4 i
4 4
4
~6 second slug INEl. S 22 787 i
l
-E ri 4
/
/
s N
m Flow tube-
~
j TC-1 Geome'try-
~
f hx$
61.09 mm
[
z 34.93 mm
(
Fluted
N/
shroud
.,/
270*f I 0" 9
INEL-S 22 758 U i l l '
180*
o Cladding thermocouples
,,),
. Internal fuel thermocouples (not. welded) l e internal thermocouple (welded)
PBF Inpile Tube and LOCA
~
Test Train IPT
- upper head d
II5
?g-gsE
.5 gg --Inpile tube
' Filler piece} ' k.e; S.4 ML
-r t Upper plenum Outlet d
.'! 2
{
C k'
Metered flow bypass I'
5
,InletM El W
SiE Upper particle screen 7
P.
Filler piece Center hanger rod-i.-
L de
..L Check valve _
~
Flow tube
.m I@
C" i Eh
' Fuel rod and flow s.hroud Bypass Y lume Filler piece
- e s
5 L
3:
-Lower support plate
+(i i ri-4 T
Lower particle sicreen L[,
k*/#
Catch basket Lower M
plenum e
e'
.4.
P
.3
-. ~.........
M
. _g
's
\\.
s l
s i
s 1
s v
\\
\\
- s. m.
g s
44.,
ys s
1 A
Z A
o N
\\
W 3\\
\\
Z g
a.
\\
CQ 5
1 I
1 E'*
i
\\
M
\\'Y.
4 Ed K'
f,.fu..,
1 3) 32 O
V.
.!i s
E h.l.-
G z
\\
O Q
\\
Q
/(.-
)
(
s's 3
y ~.'......... 8 u
cn
,.' g k g
g S*
s888 E
1
- (
I t
- : 1 J
=
j a
!\\
A d
(usus) tueusassi sig 1
1 l
j i
l
l 5
- =
W
'I.
id"
~
f
.) 8, n%'
TC-1 BLOWDOWN PEAK TEMPERATURES
~
/
1 D
CURVE /
O D E
3 aans, f
,~.
yJ e
zg, x
E.
'\\
+
+
8 f
I I
.. ms.-
CVRYS 0
y
.+
E l
$ 8889-l k
wE o
I l
r l
0 Rod 01
\\
1
~
0 Rod 02 1C-AA a Rod 03
+ Rod 04 I
W S
3 e
I TlIDS(8) i
i hM i
. - s+...
.- -..:+ -
e l
a e
i
-a i
e i
s A
0:
r r.Q b
~
2:
2-
=
/
w b
o O
/,I
/
/',,==,===,s'
,s g
/
l Q
l
\\
e g%
3
=
4 g-l
_A, -
=
N
{gjrYf i, q u
/& g#
IlI
// ')
//
I $1 Ie IW l
15 fa
<x) ====.4 2
.-%4
I"
,e;
~~~
e
+m
{!
I Blowdown Thermocouple Effects p
! l-(continued)
^
1
~
,ij'
- LLR-3 (41 KWim) showed no apparent effectw ~4*)
p
'; L.y
- f'*/
- LLR-5 (47 kWim:>
- Time to CHF ranged from f I
- 1.8 - 2.3 s for the TCs
- 0.4 - 0.5 s for the LVDTs liL i
i 1
- LLR-4 and -4A (57 kWim:i g
i.jj
- Time to CHF ranged from l
- 1.6 - 2.0 s for the TCs s.i
l.
l O.25.s on all the rods
)!. f l.
INEL424 097 pl 15,p l'
Lt/
_IMPORTANT' PARAMETERS
" CLADDING TEMPERATURE l
FUEL TEMPERATURE BURNUP (FISSION GAS) rPOWER DIMENSIONS HEAT TRANSFER;'
PLENUM TEMPERATURE L
4 PLASTIC STRAIN a
TE [
1
[ RUPTURE TEMPERATURE 4 ~
y 7 BLOCKAGE
]
3 4
1 STRAIN Jjl0WAREA CllDDINGD1i!SIONS e
RADIAT10l!
rFLO'[ DIVERS 10s 2
CONVECTION r
e u
GAP HEAT TRANSFER TRANSFEP[
METAL-WATER REACTION p
,, uy, CLADDING TEMPERATURE
,0XIDATION e
1 i
l
,.ee
=r-
-=--== g a _= = = :_ a p =_ _ _
_ _._ _ _ _=a_.__
CIRCUMFERENTIAL STRAIN (7.)
o
?
?
?
?
?
?
I
\\
\\
- \\ ' w
?
m 1
a 4
c:
\\.$
y/
8-m
- ;\\
\\
s a
.I s
/
4 m=
s m
g 8~
/
m i4
~
^
m 2
- s
). 2 '>
>$a 3:
/',
\\
^
m m
I-4 p
as h
I*
O
~
m o
o-
\\.. l\\
' l-M 9
\\
.1 N.
3 a
\\
a
\\
w gi s/
8 i
e
\\ '.1 8-I
- M l
\\'4 s
IL l
i\\
- j
?
6 e
+
l A
fi
....., FLOW BLOCKAGE MODEL COMPARISONS:..
j
~
jt B&W 3
M ENC lA s
- y i,
s 42-m fs
=
--=*/
t g
/
1
/
\\
/
/
\\
/
/
\\
p 7
')
\\
s i!
f t
/
O z
/
1
/
,/
1 p'.
Es4 g
/
i
- E YR
\\
g
,- o
.*.8 t
e-4
^\\
t
! 9
{
..- x a s e
/
8 gg 1
O.
y o
p
........,,,,.....N.
7 Va i
~ i O
y pg O
i
- ,I N
C-E V
M l! )
=4
.,g go s
ne'co atm 12 "
i TEMPERATURE (DEG. C)
,l?:j % x r
h o
y i
i.
I A,
I I
o RUPTURE TEMPERATURE CURVES & ALL DATA l
4 I
g, l
Gd
.0 i-4.1 e-
[
\\
N n
e m.
a o
w D5 i
n
~
B 0
Q 4
e M
6 0
M h' h
g o E 28 C/S g*
y
{
", 8 14 C/S a
g x
B il 8
i i
0 C/S 1-25 ENGINEERI G HOOP RESS (KPS )
i
,&w.4,a m-a----a_w.a.-
--a.--e,-
-w-r u.aw-m
.+ -...,.. _ g; g,p e q-v. -~ %..a.,,.,wm?<my&A %_M'=_. w.-*. s --t>.e,.h ( +
- m..~ t sim. w
~
~
r/
CIRCUMFERENTIAL STRAIN (%)
0 20 40 60 80 100 120 l
l I
I I
i 1
Cl2 C#O4 w
8-
=
l e
2>
a
+
=
D E ",
M S 6 q
4
++
++
- g.,
5 4
m K
W
=.
1 5
=
C 8
m r
a m 2
M CQ a
a 9
4 C
5 2
o trj
=
^
i 2
P $.
n n
C v
4tu
=
8-Q>4>
m 8
t
-4..m.--m
.a s..sA am u,
hw_-as---aA---a
4------m--
-aa -
-+-------.h.hH+-
d-Ae a-
. *==* I e
S.
e e
i CIRCUMFERENTIAL STRAIN (7.)
o ao 40 m
ao 9
9 g__
p
~
a g-6 l
3:
l
.a 1
GM
- g 1
4 W
g, h-O 3
x EmB 1
S g 'M :
Q m
+ o W
+
k x*
+
q
+
C k
s
%M 2x 3 m g
w i
e 3
25 l
D 5
m 3
O bg_
Og n
4 v
3 N
x
'2 3
x xx x
l 3
8 c
>h m
=>*
H 9
9 e
~
a.
l l.8 T
CLADDING CONSTRAINED WITH MANDREL AXIAL GAP 2.5 mm l.6
- HEATING RATE 5* C/s 1
21.4 o HEATING RATE 55' C/s y
o HEATING RATE 115'C/s ;
.2
$l i
s a
t 0
i
~
1.0 el I
BURST IN STEAM E O.8 9'
g k
48 D
o 'o
.e4 o 0.6 o
t h
/*
,e o
@,/o 0.4 R
f 4,
\\
0'2 o
y%
0 600 700 800 900 1000 l100 1200 1300 TEMPERATURE (DEG. C)
~
~
I e
J
- ?_3 Q- }ss.j g _-
~
..m,_3 ;~:
,, _ e.g
.g
_ i --- -
8 I
s g
l EE E
F i
.8
>E
=
.g, l
i D
Wm O
28 1
G 2:
as f
-g em
,e
-e i
W B'
/
Q l
V go M
-l>4
/
g M
/c' b
e m
/
a CG a
CL.
A
~k E-g l
(
O s
-8 4
s M
N N
N l
l i
i e
i i
OET 00T OB 00 09 02 0
i
(%) NIYHIS 'IVIIN2HMNnOHIO 1
.\\
A
_.=
N e
9 FAST-RAMP BURST STRAIN CURVES DASHED = DRAFT CORRELATION
~
SOLID = FINAL CORRELATION M"
2: 8-4 m
8m 8-
\\
a
\\
4 em
/
/
98-
\\
/
/
\\
/
y
\\
/
m
/
\\
/
t
/
m S-
/
\\
N
/
y
/
.\\
j l
p g g_m-O j
TEMPERATURE (DEG. C)
~
.~
BLOCKAGE MODEL I
BURST STRAIN t
x 0.57 SMALL EUNDLE BASED ON BUNDLE IESTS l
x 0.47 LARGE BUNDLE u
i
' UNIFORM COPLANAR
{
STRAIN I
FIs. 12 0F NUREG-0630 GEOMETRIC CONVERSION 1 r LOCAL BLOCKAGE 5
COMPARE WITH BUNDLE TESTS
-5%
GEOMETRIC REDUCTION 1 r ASSEMBLY BLOCKAGE
= PWR CORRELATIONS l
//
T W'
' ~^ ~
H
- d SLOW-RAMP LOCAL FLOW BLOCKAGE &
l 8,
~
l 1
48-a
+
g r
M 4
5 8-0 a
=
N 8
Z M g_
i Zo p
w g
D S-o m
--d Ai 12m itn 4
ofa 10m a
]
TEMPERATURE (DEG. C) 7#
em i
=b 1 't
a-d--w_e--44A.a h---m4mw-Am--w-----AA=w--44---'=-.a-Cs-e A-na---e.
.,--;a-. -
m ma
.u-b f M
l l_ _
}~_$Q*LW
~Y$bL
?$Y~ W?
-~
bm I
REDUCTION IN FLOW AREA (%)
0 20 40 80 80 100 i
i i
i w>
I 8-W>K T
4 M
ta g ~
9 iiC e o
m m
t-i
- c s
8
+
l 4
m c,e F
l z d-O
\\
N m
D en C#Q C =g.
Q i
O R
S v
o tu R*
8-O>
a N
8 9
e
SLOW-RAMP PWR ASSEM. FLOW BLOCK. CURVES 8
DASHED = DP. AFT CORRELATION SOLID = FINAL CORRELATION m6 48-il g
7 y
c A
\\
$ 8-
/
\\
A
/
n
\\
g
/
\\
" S-
/
z
\\
O.
/
i
)
l
\\
H j
U s
i.'
D 8-
/
o N
g 600 700 800 900 1000 1100 1200 o
TEMPERATURE (DEG. C) l I
)
a
e l
FAST-RAMP PWR ASSEM. FLOW BLOCK. CURVES DASHED = DRAFT CORRELATION SOLID = FINAL CORRELATION
< s-41 i
M
/
l
/
in 8-
/
o
/
g a
g
/
/
1
/
Z
/
g
/
)\\
/
/
f g
s o
f l
E
\\
/
f O
't
/
p g-Q i
Ng o
800 700 800 000 1000 1100 1200 TEMPERATURE (DEG. C) qo
/
PROPOSED SCHEDULE PF/ISED FUEL CL@ DING t10DELS FOR LOCAL ANALYSIS 2-ll:-80 DISCUSS NUREG-0630 WITH ACRS SUBCOMMITTEES.
3-7-80 PRESENT NUREG-0630 TO ACRS FULL COMMITTEE.
3-15-80 (APPR0x.)
RECEIVE LETTER FROM ACRS ON NUREG-0630 AND PROPOSED Il1PLEMENTATION.
L1-1-80 1.
ISSUE FINAL NUREG-0630.
2.
REQUEST UPDATE FROM LICENSEES OF ASSURRANCE THAT OPERATING REACTORS WILL MEET 2200 F LIMIT WITH NUREG-0630 CLADDING MODELS.
3.
INFORM OPERATING REACT,0RS THAT:
(A)
CURRENT CLADDING MODELS DO NOT MEET APPENDIX K.
(B)
USE CLADDING MODELS IN NUREG-0630.
(C)
SusMIT REVISED ECCS ANALYSES.
fl.
SEND LETTERS TO VENDORS REQUIRING
- REVISION OF ECCS MODELS TO INCORPORATE CLADDING MODELS IN NUREG-0630.
CONTINUED ON NEXT PAGE bkO-Af 5_9
~
- b'$
. =. -
l 1
.i PROPOSED SCHEDULE (CONT'D) 4-15-80 RESPONSES DUE FROM LICENSEES WITH INTERIM ASSURRANCE THAT OPERATING REACTORS WILL M'EET 2200*F LIMIT (RESPONSE TO ITEM 2 AB0VE).
1-1-80 REVISED ECCS MODELS (SMALL AND LARGE BREAK)
DUE FROM VENDORS (SEE ITEM 4 ABOVE).
10-1-80 COMPLETE MRC REVIEW OF REVISED SMALL-BREAK ECCS MODELS (PREVIOUS Es0 SCHEDULE).
1-1-81 1.
COMPLETE NRC REVIEW OF LARGE-BREAK ECCS MODELS.
2.
SMALL-BREAK PLANT ANALiSES DUE FROM LICENSEES (PREVIOUS B&O SCHEDULE).
j i
7-1-81 LARGE-BREAK PLANT ANALYSES DUE FROM LICENSEES.
i I
O l
/
Ass.No7807 i
Ipper grid 1
soo lTest I
Exterior rods j
Interior rods l:
~
a 250 -
h l
ev ic 1
~ 2*
l
}
]* 150 3
n I
? '
- )i-A100 i
8 8
d d E
O
'I:
yt J.
,~
._'l_.E 2 yfI a,o
'l,
v b
k
' " a c
.I T
i ;
E
-f
']
f0 D*.
M._
a
~
l.
o E
S 50 1
[ p\\
i}
._i_Tc 3 o
!100 l.
u l
D4 A1 t
,' I '
8 150 -
c G 2M
'n C
b
.2 l
250 -
.5 M: Denotes position of 9 l',
y l
9: Nonmeasured l
l
- Lower grid
,,,,,,,,,,,,,,,tr3,,,,,,,,,,,,,,,,o, E-,,LL.
---fE S2, g,g,y a c -- c u-ll ]p. Fuel asserably.
700 goo goo ( C) 12 3 4 5 6 712 3 4 5 6
- 1 t i l l 77777 23456 234562345623456 A
-Ao-
. a cot t a co I
4 Temperature N
at burst time I
l!
l l
l
~
As3.N9 7808 LNpur grid I
I*' ~?
'"i" ' **
=
=
'" ' "i" "
66 m-i 200 o
a
!c 12 150
=
4 a
A 100
- -k-Tc-2
'~
o 9 Ji s
A n
a 50 g
9 k
+
4%
--o-g v
y,
.1
. 'y
- 4*
t,-
+
(,
4+
g so re-s e*
4
./
\\
s Al 04 l100 g
e E iso
~
s a 2w g
_~_ 2so
~
.e R
t
= = Denotes nesitlen of be L
W2 k N.'i......,'.
No ' Eo ' E 'c.'c) l' ' ' ' ' } [' ' ' ' ' I I ' ' ' I I ' ' ' l l 's ' l 18 8 i l I s I t.tr e l l 12 8 45 s 7 2 3 4s s t s iij$! jyy y j,4 3j j,4 3, g,4 S g g3 4 5It s Temp a r
l t
4 9
g e & s*
6
' 4
--Tungsten nitres (()
e es e se e as e e e*
- e W
. m l
(
A1 0 Pellet
~
~
23 6
I e
e Zr Burst Specimen
+
~
,)
e e-n I
JAERI MRBT Heating Element
- - - -..