ML19351D190
| ML19351D190 | |
| Person / Time | |
|---|---|
| Issue date: | 06/27/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1741, NUDOCS 8010090180 | |
| Download: ML19351D190 (51) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _________-__- __.
/fdNS - / 7Yl n'
f b
'l DATE ISSUED: 6/27/80 t1 3[ L:2
$l27lb0 MEETING MINUTES OF THE ACRS REACTOR FUEL SUBCOMMITTEE MEETING APRIL 29, 1980 WASHINGTON, DC On April 29, 1980 the ACRS Reactor Fuel Subcommittee met in Washington, DC, to begin discussion of the NRC Research Program in the area of fuel behavior for the Com-ittee's annual reports to the Commission and Congress. The notice of the meeting appeared in the Federal Register on April 14, and April 25,1980. There were no requests for oral or written statements from members of the the public and none were made at the meeting. Attach-ment 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 for the meeting are Attachment D to these minutes. A complete set of slides and handouts is attached to the office copy of these mir:utes.
OPEN SESSION - INTRODUCTION Dr. Shewmon, Subcommittee Chairman, called the meeting to order at 8:30 a.m.
The Chairman explained the purpose of the meeting and the procedures for con-j ducting the meeting, pointing out that Mr. Paul Boehnert was the Designated Federal Employee in attendance.
Dr. She< mon introduced Dr. William Johnston, Chief, Fuel Behavior Research Branch (FBRB) to begin the day's presentations.
Dr. Johnston begin by discussing the priorities, the objectives, and summarized the FBRB programs.
He noted that the overall objectives of the fuel program are to evaluate fission product and fuel behavior un@r normal and accident conditions, develop physical models through laboratory scale separate effects tests, verify fuel codes and models through integrated tests, and utilize models and codes to assess the consequences of severe reactor accidents including core melt events and to aid in the design and evaluation of mitigation fea tures.
Dr. Okrent asked a number of questions centering on the objectives of the FBRB programs.
Following a lengthy discussion, Dr. Okrent asked that 8010090/IO
4/29/80 Reactor Fuel Meeting.
during the FBRB presentations, NRC discuss how the program results relate During e discussion of power to the defined objectives of the given program.
coolant mismatch (PCM) tests run at PBF, Dr. Okrent said he believed these tests have not addressed the issues of concern for that type of accident.
Dr. Johnston neu t tat since the TMI-2 accident, the priorities of the FBRB Program have been rearranged to the following order:
(1) core damage beyond LOCA, (2) cladding ballooning and blockage, (3) fission product release and migration, (4) operational transients (Classes I-III), and (5)
Dr. Okrent observed that the Research Program does not fuel meltdown.
factor in risk reduction and he felt that it should.
Dr. Murley replied that the NRC does not license plants based on risk analysis, rather technical judgment is used.
CLOSED SESSION (9:30 - 10:20 am), FY-81, FY-82 FBRB BUDGET - W. JOHNSTON In a closed executive session, the FY-82 FBRB budget was discussed.
Dr. Murley observed that the budget situation is very uncertain since Congress has not acted on either the FY-80 Supplemental request nor the FY-81 He also noted that a budget cut is expected, however the amount at budget.
this time is unknown.
The FY-81 and proposed FY-82 budget for the various programs under.the five priorities noted above were detailed (Figures D1-4).
In general, it was noted that the budget is increasing in the new research areas, such as core damage beyond LOCA, and decreasing in older program areas (cladding, ballooning and blockage).
In response to questions from Dr. Okrent, Dr. Murley noted that the FBRB programs are in a state of
,i transition with emphasis shifting to the priority ranking noted above.
Dr. Okrent agreed that this was the case, but questioned whether or not the pace of the shifting emphasis should be accelerated.
FUEL CODE DEVELOPMENT AND VERIFICATION - G. MARINO He Dr. G. fiarino discussed the fuel code development and evaluation programs.
noted that the objectives of these programs are to predict transient and steady
b Reactor Fuel Meeting 4/29/80 state fuel behavior under normal, off-normal, and accident conditions and to provide an integrated, easily accessible storage bank of fuel behavior information. The codes used include MATPRO - the material property corre-lations, FRAPCON - a steady-state code, used to develop fuel behavior under normal conditions, and FRAP-T - a transient code to simulate fuel rod behavior for a transient situation.
Dr. Shewmon asked when Research will feel that code development can end. He noted that he believes in many instances NRC codes are better than what the Industry has, and that the Industry is probably using NRC codes for their license evaluations.
Dr. Marino acknowledged this may be the case, but noted that code development can never stop, and that the codes will be continuously updated as new information becomes available. He noted that FRAPCON will stop at Version 11 and that the FRAP-T code will stop at the T-6 version.
From then on, the codes will be'in a maintenance status to be updated as new information becomes available.
The FBRB Code Verification Program was described.
This program involves both developmental verification, and independent verification. An example of the results of the assessment program was shown (Figures 05-6).
In response to questions from Ors. Shewmon and Okrent on the Code Assessm07t Program, Dr. Marino noted that the assessment program is showing the limits of the code's predictability, given the uncertainties in data input.
Dr. Marino reviewed the expected fuel code accomplishments for FY-80/81 as well as developments expected beyond FY-81 (Figures 07-9). Work planned beyond Fi-81 includes development of a small-break (slow transient) fuel rod damage code based on, and linkable to, FRAP-T and FRAPCON.
FBRB requested Subcommittee opinions on how far into the core melting sequence the code should go.
Dr. Shewmon expressed primary concern about coolability versus failure modes.
Dr. Okrent expressed skeptism that the above LOCA code will be able to model core coolability for severe damage cases.
i i
1
i Reactor Fuel Meeting 4/29/80 FUEL PELLET AND FUEL ROD PROPERTIES RESEARCH - G. MARINO Dr Marino noted that the objectives of the Fuel Pellet and Fuel Rod Properties research programs are to provide information on changes to fuel pellets during steady-state and transient operation, improve models for calculating gap conduc-tance in a fuel rod, and determine the extent to which fuel pellets affect the transient axial flow of gas within a fuel rod. The bulk of the program is centered on a series of instrumented fuel assembly (IFA) tests being conducted at the Halden reactor in Norway. Figures 010-14 detail the Halden tests and results to date.
Another section of the program deals with the development of a transient fission gas release code (GRASS, FASTGRASS). The GRASS and fast-- ming version (FASTGRASS) are under development at Argonne. Comparisons of the two codes predictions versus results of spent fuel pellet direct electical heating tests to determine fission gas release rates, show fairly good agree-ment (Figures D15-16).
Dr. Marino noted that work on an even faster running GRASS code version ( PARAGRASS) has been initiated. The FASTGRASS code, and when complete the PARAGRASS code, will be incorporated into the FRAPCON and FRAP-T fuel codes.
ZIRCALOY CLADDING RESEARCH - M. PICKLESIMER - NRC Dr. Picklesimer discussed the zircaloy cladding research programs now underway.
These programs include the Multi-Rod Burst Test (MRBT) program conducted at Oak j
Ridge, the study of the Mechanical Properties of Zircaloy at Argonne, a study of j
the strength and conductivity of the irradiated zircaloy using PWR spent fuel clado;ng at Battelle Columbus, and a study of zircaloy cladding creepdown in-pile via a joint program between ORNL and ECN PETON in the Netherlands.
Highlights of the above programs include the following:
'The MRBT program is designed to characterize ballooning, burst, and j
loss of flow area in bundles of LWR fuel rod simulators during the refill-reflood stage of a LOCA. A number of single-rod and 4 x 4 bundle tests have been run, an 8 x 8 rod bundle test is scheduled for early June.
Recent results (Figure D-17) indicate that single rod tests may be used to successfully model bundle behavior.
It is hoped that if the results of the 8 x 8 rod bundle tests duplicate the 4 x 4 bundle results, a scaling factor will be able to be developed.
s f
Reactor Fuel Meeting 4/29/80
'The Mechanical Properties of Zircaloy program designed to quantitatively characterize zircaloy clad embrittlement by oxidation with steam, and relate this embrittlement to measureable mechanical properties of the embrittled material. The second phase of the program is designed to deter-mine the stress-rupture properties of spent LWR cladding under simulated 4
PCI conditions leading to clad rupture. The results of the recently
~
completed first program phase are embodied in a set of failure limits for embrittled cladding for the cases of thermal-shock and impact. For the thermal-shock case, it is proposed that at least 0.1 millimeter of 1
clad wall thickness remain with less than 0.9 weight percent oxygen. For the impact criterion, no less than 0.3 millimeter cladding wall thickness contain less than 0.7 weight percent oxygen.
'Dr. Picklesimer discussed the planned research program on PCI. Elements of the program will include the study of PCI failure by stress rupture, strain-rate ramping to PCI failure, both ex-pile and in-pile. The in-pile tests will be conducted at the Studsvik and PBF reactors (Figures D18-21).
Dr. Okrent questioned whether or not a study of PCI was more appropriate for DOE or EPRI.
Dr. R. Meyer (NRR) replied that industry does not believe that PCI is a safety concern, while the NRC does.
FBRB FUEL MELT PROGRAMS - R. SHERRY - NRC j
Mr. Sherry discussed the LWR fission product release and transport programs both underway and planned by FBRB. The overall objectives of this program are to develop fission product source terms for fuel rods under accident conditions including severe core damage and core melt, to develop models to predict the attenuation and transpor't behavior of fission products' within-the primary coolant system and containment, and to provide environmental i
release source tarms for consequence analysis and evaluation of engineered safety features and mitigation feature design requirements.
There are five existing programs and two planned programs in the fission product release and transport research.
The five ongoing programs are: (1) fission product transport analysis - TRAP code (Figure D-22), (2) separate effects
Reactor Fuel Meeting 4/29/80 tests for the TRAP code (Figure 0-23), (3) fission product vapor deposition requirements (Figure 0-24), (4) steam generator tube rupture iodine transport (Figure D-25), and (5) fission product release from LWR fuel (Figure D-26).
The two planned programs include a study of fission product release from LWR fuel at high temperature (Figures 027-28), and charcoal filter iodine retention performance studies (Figure D-29). Commenting on the last program, Dr. Lawroski said he believes such a program shculd be conducted by DOE or the Nuclear Industry, not NRC.
Mr. Sherry also described four future programs that are under consideration.
These programs are:
(1) fission product release from melting fuel (Figure D-30),
(2) fission product leaching (Figure D-31), (3) development of a fission product transport verification facility (Figure D-32), and (4) examination of TMI fission product release data (Fioure D-33).
Concerning the first program noted abovt, the study of fission product release from melting fuel, Dr. Okrent suggested FBRB try to determine if a " quick and dirty" experiment would be more productive in order to get a yard stick on the estimates for core melt rs 4 5 es described in WASH-1400.
SEVERE CORE DAMAGE STUDIES - M. PICMLESIMER - NRC Dr. Picklesimer briefly reviewed the new proposed program to study severe core damage. This program would focus on the following areas: (1) development of core damage, (2) fission product distribution, (3) modeling of severe core damage, and (4) code development for prediction of core damage. Research areas in core damage would include integral bundle effects, and separate effects and basic studies, all both in-and out-of-pile. Also to be included would be a study of fission product release and distribution in the primary system, and the modeling for prediction of core damage.
In-pile tests would be con-ducted,in the PBF or ESSOR test reactors, and the planned TMI-2 core' examinations would be relied upon to provide related information. Some of the above pro-grams will begin in FY 81 with the majority of them scheduled to begin in FY 82.
Dr. Okrent suggested that the extensive debris bed experiments conducted in the fast reactor program may provide relevant information in lieu of tests in PBF. The NRC Research noted that the diffe ences between the fast reactor
b l
Reactor Fuel Meeting 4/29/80 and LWR systeras preclude significant transfer of. knowledge from one group of experiments to the other.
In response to a question from Dr. Shewmon, Dr. Picklesimer noted that the tests planned for ESSOR would not begin until late 1982 or early 1983.
INTEGRATED CORE MELT RESEARCH PROGRAM - C. KELBER - NRC Dr. Kelber briefly discussed the program logic for the Integrated Core Melt Program recently formed in NRC Research. The prcgram objective is to determine the best-estimate of risk, analysis, and assessment of special features designed to assure containment integrity in the event of a core mel t.
To this end, Dr. Kelber said a number of significant questions must be answered (Figure D-34). ACRS review of the integrated core melt program will be conducted by the new Class 9 Subcommittee which has scheduled a meeting for May 9,1980.
TMI-2 CORE STATUS STUDIES - M. PICKLESIMER Dr. M. Picklesimer discussed the TMI-2 core status studies conducted as part of the Special Investigation Group for the Rogovin Report. He described the core status at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> into the accident when significant core darrage had taken place, and at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the accident when additonal core daca5e occurred. The estimates noted below are based on computer code and system analyses.
At the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> mark in the accident, it was estimated that the maximum core 0
temperature reached was about 4400 F in the upper 3 feet of more than two-thirds 0
of the core, and a temperature of 3600 F was reached for all of the core down 5 feet from the top of the fuel assemblies. A 2 foot thick debris bed was probably formed with a base at about 8 feet from the bottom of the core over the entire core, and its formation was aided by the thermal shock of embrittiti cladding and " liquified fuel" at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 54 minutes in the transient when the 2B reactor coolant pump was starteu.
At 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the transient, more than 60% of the zircaloy in the core had been embrittled or shattered, and the lower surface of the debris bed
4/29/80 Reactor Fuel Meeting i d fuel had dropped to about 5 feet from the bottom of the core, and liquif e It had penetrated to within 1 foot of the bottom of the core in some areas.
d had been pro-700-820 pounds of hy rogen was also estimated that a total of Dr. Picklesimer also said duced by oxidation of zircaloy at the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> mark.
ii l
that the stainless steel upper end fittings must have suffered add t ona In response damage, but the degree of damage cannot be estimated at this time.
fuel to a question from Dr. Shewmon, as to how hard it would be to pull the from the core, Dr. Picklesimer replied that the current thinking it that ible the outer fuel bundles will be able to be removed, then it would be poss l has to work from the outside of the core in to the center, if the core barre not shifted its position.
NRC_
HYDROGEN AND POST-ACCIDENT COOLANT CHEMISTR Mr. Don Hoatson discussed the hydrogen and post-accident coolant chem These programs are new, and have grown out of the concerns research programs.
The programs include a study of combustible raised by the TMI-2 accident.
gas generation in containment, a study of hydrogen, and a post-accide The Combustible Gas in Containment Program is coolant chemistry program.
ity primarily concerned with the formation of hydrogen from the large quan This program of zinc found in containments and containment-related structures.
grew out of the User's Need letter from NRR.
(1) radiolysis of reactor solutions, The Hydrogen Program will study such topics as:
(2) sampling an analysis of hydrogen in reactor emergencies, (3) fl t ccident and detonation limits under accident conditions, and (4) handling of pos -a FBRB'is developing a compendium of information on hydrogen, focu For example, it f
on the topics above and to be used in appropriate emergencies.
t ater was noted that to mitigate the effects of hydrogen in containmen, a w A draft of the compendium is scheduled to be fog system is under study.
issued in May 1980, with the final report issued in June 1980.
The post-accident coolant chemistry program will address the areas of
" rom failed fuel, a study of iodine in containment, fissio. product signatu;c and a follow-on of the radiolysis work from the hydrogen program noted Mr. Hoatson noted that both the hydrogen and the post-accident coo l
Reactor Fuel Meeting 4/29/80 programs are scheduled to obtain money from the FY-80 supplement.
In response to a question from Dr. Okrent, Dr. Johnston said that had he known how uncertain the supplemental appropriation would be. -he would have requested funding of the hydrogen program be tahen directly from his normal budget.
During a Subcommittee discussion, Dr. Okrent observed that he felt that experiments scheduled for facilities such as PBF must be,iudiciously planned and carefully 4
thought out in order to obtain maximum benefit for the large amount of funds being expended.
Dr. Mark observed that Research needs to sort experiments between what is useful to NRC, and what is of more use to the Nuclear Industry.
He feels that such work as a study of degraded radioisotope filters may not be all that necessary a research topic. Dr. Johnston replied that he agrees with Dr. Okrent that any experiments run in PBF for example, encounter difficulties because of such constraints as length effects, and
- oth NRC and the PBF program representatives are cognizant of these problems and take steps to minimize their impact. Dr. Johnston also said that the work on degraded filters was based on a request from NRR for information in this area.
NRC-NRR CORE PERFORMANCE BRANCH PRESENTATIONS - R. MEYER, D. HOUSTON Dr. Ralph Meyer and Mr. Dean Houston from the Core Performance Branch of NRR discussed the CPBs' Technical Assistance Program and information pertaining to fuel failures in operating LWRs.
Dr. Meyer began by noting f
that with the reorganization of NRR, the effort in the fuel area has been reduced in man-power from 11 to 4.
He also noted that NRR is reconsidering its priority on the RIA, since recent infomation has indicated that the accident consequences may be much less severe than previously believed.
Dr. Meyer then reviewed the CPB effort in technical assistance, noting that the total budget in technical assistance is about $380,000 for FY-80 (Figure D-35). During discussion of che technical assistance program that is addressing DNB-induced fuel failure propagation, Dr. Okrent questioned how NRR would make use of the data available to answer questions in this area.
Dr. Meyer responded that he would prepare a memo to address Dr. Okrent's concern and forward it to the ACRS.
l
Reactor Fuel Meeting 4/29/89 Mr. Dean Houston reviewed the fuel failure history for operating LWRs for the past year. He noted that, excluding the TMI-2 failures, there have been 116 assemblies that have experienced some degree of fuel failure (Figure D-36).
Mr. Houston detailed the various fuel failure mechanisms (Figure D-37). Again, excluding the TMI-2 failures, failures were seen due to such mechanisms as water-side corrosion, stress corrosion cracking (in stainless steel fuel),
PCI, handling accidents, ana vibration and fretting.
.The meeting was adjourned at 5:05 p.m.
Note: Additional meeting details can be obtained from a transcript located in the NRC Public Document Room, at 1717 H St., 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.
Federal It- ' r / Vol 45. No. 73 / Monday. Apr014.19e0 / Notices c-At the conclusion of the Executive NUCLEAR REQULATveW Session, the Subcommittee willheat COMMISSION presentations by and hold discussions with representatives of b NRC StaE.
Mvteery Committee on Reactor their consultants, and other lateresed
$sfeguards. Subcommittee on Reestor persons.
~
Fus:; Meeting In addition. it may be necessary for The ACRS Subcommittee on Reactor the Subcommittee to hold one or more Ful willhold a mating on April 29.
closed susions fw the purpose of teso in Room 1046.1717 H St NW.,
exploring maners involving proprietary Washington. DC 20655 to discuss the information.! have determined,in NRC research program on reactor fuels accordance with Subsection 10(d) of the for the ACRS annual reports to NRC and Federal Advisory Committee Act(Pub.
Cocaress. Notice of this meeting was I.95-463), that. abou!d such semions be pubbshed Man h 19.1980.
required. It is necessary to close these In accordance with the procedures sessionr to protect proprietary outhned in the Federal Register on leformatica. See 5 U.S.C. 552b(c)(4).
October L 1W9. (44 FR 56408). oral or Further informstion regard!=3 topics written statements may be presented by le be discussed. whe&ar the meeting members of the public, recordings will b been cana!!ed or reacheduled, the be permitted only during those portions a ruling on ruguests for the of the meeting when a transcript is being
""=ty to t oral statements kept. and questions may be asked ordy "nd the Mme ned bmfor can be a
by members of b Subcommittee.4ts obtained by a prepaid telephone call to consultants, and Staff. Persons desiring the cognizant Desipated Federal to make oral statements should notify Employee Mr.Pam A.Boehnert the Designated Federal Employee as far (talephone 202/634-3287] between 3:15 in advance as practicable so that a.nL and SSO p.m EST.
appropriate arrangements can be made Dated:Aprt!s.isso.
to allow b necessary time during the
- )ebe C. Heyle.
meeting for such statements.
Advisory cansvase ?' _ r --s offreer, The agenda for subject meeting shall be as follows: Tuesday. April 29.1.900 pn m m.mse raa m ma,,,g au.sms ocos rune.es n 330 a.m. until the conclusion of business.
The Subcommittee may meet in Executive Session,with any ofits consultants who may be present. to axplore and exchange their prelimmary opinions regarding matters which should be consideredtiuring the meeting.
Faiwat heister / Vol 45. No. 81/ Friday. AprG 25. 1980 / Nodoes k
Advisory Committe'e'Act (Public Law Adwtoory Committee on Reactor 95-463), that, should such sessions be safeguards; Sut> committee on Reactor required. It is necessary to close Fuel; Meeting Addition portions of this meeting to prevent It may be necenary for the ACRS 6ustration of the above stated aspect of Subcommittee on Reactor Fuel to close the ACRS' statutory responsibilities. See portions of its meeting on April 29.1980 g UJ.C. 552b(c)(9)(B).
All otheritems remain the same as for the following mason:
The ACRS is required by Section 5 of announced in a notice published April the 2W8 NRC Authorfsation Act to 14.(45 FR 2S196).
review the NRC research program and Dated: Apniza,1ses.
budget and to report the results of the John C. Heyle, review to Congress. In order to perform Adrisorycamavanaaseogsame this review, the ACRS must be able to pu m so.ma raw sa es.g engage in frank discussions with mass ocas reemas members'of the NRC Star and such discusalons wod!d net be pomibleif held in public sessions. I have determined, therefore. In accordaner Attachment A with Subeection10(dl of b Federal
ACRS REAC. TOR FUEL SUBCOMMITTEE MEETING APRIL 29,1980 WASHINGTON, DC ATTENDEES LIST ACRS NRC P. Shewmon, Chairman W. V. Johnston, RES S. Lawroski, Member M. L. Picklesimer, RES J. C. Mark, Member G. P. Marino, RES D. Okrent, Member D. A. Hoatson, RES/FBRB A. Bement, Consultant R. Meyer, NRR/CPB F. Nichols, Consultant C. Kelber, RES P. Boehnert, Staff
- Designated Federal Employee R. Leyse EG&G H. I. Zeile J. E. Hanson C. R. Toole P. E. MacDough Attachment B
~.,...n-....
ARPIL 29, 1980 WASd!NGTON, DC
- Tentative Schedule of Presentations -
Presentation
- Actual Time Time 10 min 8:30 am 8.
Introduction P. Shewmon, Chairman II.
HRC Fuel Behavior Research Branch (FBRB)
Presentations 20 mi n 8:45 am A.
Priorities, Objectives, and Sumary of FBRB Program W. Johnston, Chief, FERB 20 min 9:15 am B.
FBRB Budget (Closed Session)
W. Johnston
'FY 81-82 Budget
' Status of Budget Supplement Request 10 mi n 9:45 am Break -
35 mi n 9:55 am C.
Fuel Code Development and Verification G. Marino D.
Fuel and Cladding Programs 20 min 10:40 am
' Fuel Programs - G. Marino 20 min 11:15 am
' Cladding Programs - M. Picklesimer
- Cladding Swelling and Rupture
- PCI
- Incipient Fuel Melt 30 mi n 11:45 am E.
Fuel Melt Program
' TRAP Code and Related Studies - R. Sherry
' Severe Core Damage Study -M. Picklesimer 60 min 12:30 pm Lunch -
- Additional time has been allotted for Subcommittee Questions
/)Tf/ICYN$W S April 29,1980 Aeactor Fuel Meeting
- Tentative Schedule af Presentations -
Presentation
- Actual Time Time 15 min 1:30 pm F.
Comments on Integrated Core Melt Program and Related Studies C. Kelber 45 mi n 1:50 pm G.
TMI-2 Core Status W. Johnston M. Picklesimer 30 min 2:50 pm H.
TMI-2 Research Programs D. Hoatson
' Hydrogen Programs
' Coolant Chemistry Programs 10 min 3:30 pm I.
Summa ry W. Johnston 10 min 3:40 pm
- Break -
III.
NRR Core Performance Branch Presentations 15 min 3:50 pm A.
Technical Assistance Programs R. Meyer 20 mi n 4:15 pm B.
Discussion of Recent Fuel Failures D. Houston 25 mi n 4:40 pm C.
Regulatory Position on PCI M. Tokar 15 mi n 5:15 pm IV. Discussion 5:30 pm V.
Adjourn
- Additional time has been allotted for Subcommittee Questions
DETAll BUDGET - FUEL BEHAVIOR I
EL82 EL80 EL81 EUL1 C0ilIRACIQR IIIIE CORE DAMAGE BEY 0ilILLOCA 750 (350) 500 EXAMINATION OF TMI FUEL 1695 2250 B7084 ANL CORE DEGRADATION IN ESSOR 700 (500) 800
~ '
B5702 PNL HYDR 0 GEN HANDBOOK AND DATA BASE 2810 (1900) 2135 B7100 SAN SEVERE CORE DAMAGE - PBF 300 300 UND.
EGgG INCIPIENT FUEL-CLAD MELTING 200 400 B7281 POST-ACCIDENT COOLANT CHEMISTRY (E00)
B2372 DEBRIS C00 LABILITY STUDIES 350-MODELING 0F SEVERE CORE DAMAGE
_400-
__U1001
_a00 REACTOR CHEMISTRY 6530 7960 B7200 500 1000 INLET FLOW BLOCKAGE TESTS UND.
3050 CLADD1HGJALLODilulG_AND_ BLOCKAGE 3015 1875 LOCA BUNDLE REFLOOD IN NRU 505 960+(250) 900 B2277 $
PNL MULTIR0D BURST TEST 160 100 155 B0120 ORNL RESIDENT ENGINEER-CADARACH, FRANCE 1150 11Z5_
_93Z_
- UND, LOCA TEST IN PBF 4G52 g
EG8G AG041 4105 s
j
(CONTINUED) f EIN_f CONTRACTOR TITLE ELBil FY 81 ELB2 FISS10lLERODUCT RELEASE _AllD MIGRATION B6747 Ill PROCUREMENT FISS10ll PRODUCT TRANSPORT ANALYSIS 75 3-5MY E05*
UND.
TMI FISSION PRODUCTI0il IN CONTAIN-(175) 85 200 MENT B0127 ORNL FISSION PRODUCT RELEASE AT ll1G11 (365) 400 (575)
TEMPERATURES A2016 ANL TRANSIENT FISSION GAS RELEASE AND MODELING 150 105 225 NRL 10 DINE FILTER AFFECTIVENESS TESTING (110) 115 (122)
FISSION. PRODUCT TRANSPORT VERI-FICATION FACILITY 1000 FISSION PRODUCT RELEASE FROM MOLTEN FUEL
'700)
A1227 SAN SEPARATE EFFECTS STUDIES FOR TRAP 150 210 LEACHitlG 0F FISS10il PRODUCTS FROM FUEL 100 120 -
MITIGATION OF LIQUID PATHWAYS 1201.
M RELEASES (CONTAINMENT BYPASS) 2280 ',
347
y DETAIL BUDGET - FUEL BEllAVIOR (CONTINUED)
ElfL1 CONTRACTOR TITLE FY 80 FY 81 EllZ OPERATIONAL TRANSIENIS_AND INITIAL C0!lDITIO3S A6050 EG8G FRAP AND FRAPCON CODE DEVELOPMEilT 690 730 700 A6046 EG&G CODE ANALYSIS AND ASSESSMENT 245 260 270 A6041 EG8G OPERATIONAL TRANSIENTS - PBF 3000 3060 2810 B2043 PNL EXPERIMENTAL SUPPORT AND DEVT. OF SINGLE R0D FUEL CODES 430 570 650 A2017 ANL STRESS RUPTURE OF IRRAD. CLADDING 450 370 505 B5531 NRC H0
!!ALDEN PROJECT MEMBERSHIP 477 490 535 A6041 EG8G PCM, RIA TESTS IN PBF 2761 2880 1873
- UND, MODELING 0F OPERATIONAL DAMAGE 500 TO ZlRCALOY 500 B7202 UND.
LONG BUNDLE TESTS (150)
(150)
UND.
RESIDENT ENGlilEER - NSRR JAPAN VARIOUS. EG&G PBF OPERATION AND SUPPORT 6012 5721 5170 GENERATION IN C0ilTAINMENT 100 -
_1%
410 H2 B674G '
15230 14723 i t>
i
. ' ~
W
I DETAIL BUDGET - FUEL BEllAVIOR (CONTINUED)
EllLi C0flIRACIDIL TITLE FY 80 FY 81 FY 82 EUELfELTD0lill*
A1030 SAN STEAM EXPLOSIONS 500+140 915 620 A1019 SAN MOLTEN CORE / CONCRETE INTERACTIONS 194+(56) 210 155 500+
UND.
FUEL MELTDOWN SYSTEMS CODES (ggg-UND.
FUEL MELT MITIGATION FEATURES (300)
(G00)
EVALUATION UND.
RESIDENT ENGINEER - KARLESRullE 100 130 140 INCLUDED IN INTEGRATED FUEL MELTDOWN PROGRAM.
l ts
(
4
FRAP-T5 STANDARD MODEL ERRORS Outout carameter Sa cle (rods /ots)
Standard error
, 0. 5
, n E (P - M )2/n-1 j
.i=1 CHF power at known flow 30/87 0.04 kW/CC channel 2
CHF flow at known power 30/97 390 kg/s-m Initial fuel centerline 21/32 250 K temperature at scram Fuel thermal decay constant 21/32 5.7 s during scram Equilibrium fuel centerline 21/32 57 K temperature during scram MATPRO FRAIL Cladding burst temperature at (155/155) 160 K 94 K known pressure Cladding burst pressure at (61/61) 16 MPa 23 MPa known temperature Cladding permanent hoop strain (327/327) 32% cladding 00 33% cladding 0.0 FRAP-T4 STANDARD MODEL ERRORS Outout carameter Samole (rods /ots)
Standard error 0.5 in I (P - M )2/n-1 g
g 1=1 CHF power at known flow 18/97 0.06 kW/CC channel 2
CHF flow at known power 18/37 400 kg/s-m Initial fuel centerline 21/32 280 K temperature at scram Fuel thermal decay constant 21/32 5.4 s during scram Equilibrium fuel centerline 21/32 54 K temperature during scram MATPRO FRAQ Cladding burst temperature at (153/158) 290 K Not Analyzed known pressure Cladding burst pressure at (64/64) 34 MPa Not Analyzed known temperature-Cladding permanent hoop strain (370/370) 57% cladding OD Not Analyzed p-s-
RY OF STANDARD DEVIATIONS BETWE SUMMA EMENTS AND PREDICTIONS Standard Deviation Sample Size FRAPCON-1
(# of Rods /# of Points)
Output Parameter 294K
-32/274 (Pressurized Rods) 170K Fuel Centerline Temperature 61/472 (Unpressurized Rods) 15.9 %
145/145 1.38 MPa Released Fission Gas 20/330 (Unpressurized Rods) 1.93 MPa Rod internal Pressure 28/285 (Pressurized Rods) 11.4 KW/M 88/88 0.37 %
Gap Closure Heat Rating 18/160 Axial FuelThermal Expansion 0.45 %
Permanent Fuel Axial 97/354 Deformation 0.47 %
Permanent Cladding Hoop 154/358 0.15 %
Strain Permanent Cladding Axial 96/119 Strain 5.8 micron -
Cladding Surf ace Corrosion 40/69 37.2 ppm Layer 33/46 10821 W/m K 2
Cladding Hydrogen 17/112 (Unpressurized Rods) 21200 W/m K 2
Concentration Gap Conductance 20/115 (Pressmized Rods) 208K Fuel Off-Centerline
'20/111 Temperature l
E i
h
EXPECTED FUEL CODE ACCOMPLISHMENTS IN FY 80/81 A.
ASSESSMENT OF FRAP-T5 COMPLETED B.
COMPLETION AND ASSESSMENT OF FRAPCQE2 - LAST VERSION OF CODE - MODEL UPDATING AS A RESULT OF ASSESSMENT AND NEW DATA WILL CONTINUE.
HOWEVER, A NEW VERSION I.E., FRAPCON-2 MOD 1 WILL NOT BE MADE UNTIL SUFFICIENT CilANGES TO THE MODELS WhRRENTIT.
C.
COMPLETION AND_ASSESSBENT OF FRAP-T6 - LAST VERSION OF CODE.
D.
MaIER0-11 REVISION-1 COMPLETER O
l 4
, ss
EXPECTED FUEL CODE ACCOMPLISilMENTS IN FY 80/81 (CONT.)
E.
MAJOR IMPROVEMENTS EXPECTED:
FRAP-T6:
LINK WITil FASTGRASS GAS RELEASE MODEL FROM ANL, A NEW BALLOONING MODEL BASED ON MRBT RESULTS, COMPLETE DYNAMIC STORAGE ALLOCATION, AN UPDATED FAILURE SUBCODE (FRAIL 6) COMPATIBLE WITil BALLOON-2, IMPROVED USER INPUT AND OUTPUT, 0-VARYING HTC MODEL, AND MANY OTHER SMALLER IMPROVEMENTS.
COMPLETION DATE JANUARY 26, 1981.
FRAPCON-2:
LINK WITH FASTGRASS, COMPLETE DYNAMIC STORAGE ALLOCATION, PELET MECllANICAL PACKAGE FROM GAPCON-3, IMPROVED INEL MECilANICAL PACKAGE, IMPROVED RELOCATION MODELS FOR BOTli MECHANICAL PACKAGES, ANS 5.4 GAS RELEASE OPTION, NRR-APPROVED EM MODEL OPTIONS, AND MANY OTilERS.
COMPLETION DATE AUGUST 15, 1980.
MATPRO-11 REVISIO(1-2:
INC BCL ANNEALING PROPERTIES, TRUE STRESS / STRAIN U.F.
DATA, REVISED CLAD CREEP AND THERMAL EXPANSION MODELS, UPDATED 110T PhESSING MODEL.
COMPLETION MID 1981.
l t
A
WORK PLANNED FOR FY 81 AND BEYOND BEGIN DEVELOPMENT OF A SMALL BREAK (SLOW TRANSIENT)
A.
CODE BASED Oli AND LINKABLE TO FRAP-T AllD FRAPCON.
CONTINUE TO IMPROVE THE HOST CRITICAL fl0DELS IN FRAP-T AND FRAPCON; B.
HAMELY, FUEL RELOCATION AND CRACKED FUEL THERMAL AND MECHANICA CLAD BALLOONING, PCI FAILURE ANALYSIS, AND LINKS WITil T/H CODES SUCH AS TRAC AND COBRA.
C0 ORDINATE WITil NRR PERS0ilNEL TO PLAN AND AClllEVE FUEL C.
STUDIES PERTINENT TO LICEHSING STUDIES USING Tile AB0VE CO O
\\
4
PROGRAMS TO STUDY FUEL ROD PROPERTIES Halden Tests (EG&G?
IFA-429 In-Reactor Measurement of Helium Absorption, Steady State and Transient Fission Gas Release, and Fuel Centerline Temperature as a Function of Burnup, Power, Gas Pressure, and Pellet Cladding Gap.
18 PWR - Type Rods-Pressurized to 375 psi - 25 cm Long.
l IFA-430 - In-Reactor Measurement of Transient Axial Gas Flow l
and Centerline Temperature as a Function of Gap Size, Power, and Gas Flow Rates Plus Two Rods Unpressurized instrumented for Fuel Temperature Measurements.
i a
t u
l C
1
ACCOMPLISilMENTS TO DATE FOR IFA'S f129 AND 1130 AMOUNT OF HELIUM ABSORBED RESULTS:
IFA l129 - IIELIUM ABSORPTION REPORT ISSUED.
PERIODIC POWER INCREASES REDUCES PRESSURE BY AN lilSIGNIFICANT AMOUNT (1.5%).
BURNUP IS NOW AT 9000-211000 f
(UP TO 50%) DID NOT DRIVE DUT THE ABSORBED HELIUM.
A PIE REPORT IRRADIATION WILL CONTINUE THROUGH 1980 AND IN WILL BE ISSUED ON TWO RODS REMOVED AFTER 8000 MWD /MIM.
IN 1981 OR 1982 Tile TEST TRAIN WILL BE REMOVED AF PRELIMINARY RESULTS INDICATE THAT AT MWD /HTM.
IFA fl30 - BEGAN IRRADIAT'ON 11/26/78.
POWERS WilERE REDUCED GAS FLOW WAS EXPECTED, AN TilAN ONE-IIALF Tile INITIAL WAS PRESENT IN PRESENT BURNUP IS 3000 MWD /MTM.
TIGilTLY CLOSED.
DATA USING DIFFERING GAP GAS COMPOSIT AT VERIFIED Tile MODELS IN FRAP-T FOR GAP CO PRESSURES > 1.0 MPA AND XE CONCENTRATIONS OF LOWER GAP CONDUCTANCE THAN OBSERVED.
DATA WILL BE CONTINUALLY IRRADIATION WILL CONTINUE THROUGli 1980 AND 1981.
C5 COLLECTED AND REPORTED.
t I
1
PROGRAMS TO STUDY FUEL ROD
~
PROPERTIES (CONT'DD Halden Tests (PNL?
IFA-431/432/527 In-Reactor Measurement of Centerline Temperatures (Both Ends) as a Function of Burnup, Power, Gap Width, and Gas Composition. Six 50 cm Unpressurized BWR Type Rods Each Assembly.
i IFA-431:
PIE
- Complete, Peak Burnup 5000 MWD /MTM.
Reports issued: NUREG/CR-0318, NUREG/CR-0332, I
NilREG/CR -0749, - 0797.
5 i
IFA-432:
Presently in-Reactor,.
Average Burnup 2f,,000 l
MWD /MTM.16 of 26 Instruments Still Working. To be Discharged From Reactor in CY 1981. Reports I
Issued:
NUREG/ CR- 0110,-osso,- 1: 39 IFA-527:
Xenon Filled Rods to Determine Pellet Relocation Effects. To go In-Reactor M AY 1980 i
1.
9
PROGRAMS TO STUDY FUEL ROD
~
PROPERTIES (CONT'DD Halden Project Sponsorship IFA-513 Same as IFA-431 Except: He-Xe Gas Mixtures; Longer Length; Continuously Recording Pressure Transducers; intermediate Power; and One Rod Pressurized to 45 psi Helium.
Began irradiation 11/78. Rods Will be Used Later in PBF for Rf A and LOCA Tests.
Decision Regarding Removal / Continued Irradiation to be Made CY 1980.
j Reports issued: NUREG/CR-0862, NUREG/CR.1077.
l9
!m
ACCOMPLISilME!!TS TO DATE FOR IFA'S 431, 432 AND 513 BOL MEASUREMENTS OF TEMPERATURE, POWER, AND CLADDING A.
N0 lilGil BURNUP ENHANCED FISSION GAS RELEASE NOTED TO
- 1.
NO ADVERSE EFFECTS NOTED IN TWO RODS CONTAININ 2.
Tile DEVELOPMENT OF A NEW MODEL FOR FUEL RELOCATIO 3.
AND CRACKED FUEL ELASTIC MODULI WillCil WILL BE T CRACKED FUEL CONDUCTIVITY WAS REDUCED BY 4.
FOR 80% FUEL RELOCATION AT 30 KW/M.
1/40 0F SOLID U02 Tile RESULTING FUEL / CLAD GAPS IIAVE REMAINED 5.
IFA-432 WILL BE REMOVED IN, SUMMER OF 1981-B.
PIE COMPLETE ON 431.
9 4
m-i
ae 3I-O MEASURED GAS RELEASE (%)
~
~
w w
s a
w w
m en O
w O
w O
V O
w O
O O
w O
V i
l I
i w
De e
cn 2 o
a
- o o
m e
y
>< P Z
w
-4 o
-i l
en G
Oo m
I Z cm (n
en e
G
-I _
N cn d
O o
1 o
i a
M N
i m
s2 l
o o
u I
s g
0 0
m u
O t.11
-i w
- 23 A
p o
>z c-
'A m
y m
B; w
2j m
w
-N O
O o-l.11
~
m
. C.11 m
E 8
>en O
CD w
W O
4
i I
i l
1 DEH TEST 60 55
- H.B. ROBINSON 50 "l
i 2 45 e
w E
A0 b
M 35 W
e 30 Sg 25 W
to 20 1
1 5-10 5 -+
0 5
10 15 20 25 30 35 40 45 50 55 60 55 70 0-PREDICTED GAS RELEASE (%)
FASTGRASS-PREDICTED TRANSIENT GAS RE O
D-!C
N I
4 S!:;CLE ROD HEAID SilF3L'] TESTS TYP! CAL OF 57% !.1 035ERVD Stu!. I:
1 FOR COTP.WELE CD'31710'G !!! EL*. LE TESTS 8
m I
0 j
t\\
- R
~m r%
y s
h ar _asgamLfirs / A r1 wie*--
E Ug 83'4 se.1tl Tps t P ess. Te me. s t=m.*
Teer 44.k. Sie j
3 Svest aut5r nie. (47.8
('et
( *. 7 Oce)
( %s
)
b d
- iA 71.
9+if 779 47 42.7a 4J7 4-h, g 83 6 48tt 770 st 4( 49 e s.9 +
Q Berk tests.woth Jame Ast hme/sto#
)
d sai same ss e msss tres i
y I
E
(
l e
/
\\
l
[
=
s' (sv\\
j %
(
u, y~
D
/
arta t,,,
so go so a
70 so so too to ao
-so o
DISTRNCE FRCa BOTTCM Cr HERT C CCNC (CM) i 1
4 STMIN IN $!1GLE RCD HEAID SUP0','3 TE$iS TYPICAL OF STull CESERvD FOR CCf?WSLE CD'131TIOiS IN !!,*.0LE TESTS 4
s --
rc 1: it te n it te it 1
1 I i r_
m a
g w -- - _
/
x,,,,.<
.y i
Carmennten ol 8 err l A.fu ne.e$e**f
=
N strnne r.e.e 1.o. w. s u-,
.?
rest rets1 re.mo
~
x A ctral c cy cv.s ese<>
< *?.s
"' - se.t s ssst Tss ss
- .ro 421 n-
~
Es 9313 774 JJ 4J 11 47.s f b - G3 sos.in rases
- same not on tai.-
0 h A
=
and some 2/ene mest fles dS s
l l
uun I k O
~-
ma b
f u
,.g m /a V
Q
-4ri0
~
y 4ttp 23 30 to 13 W
To ao so 100 e
~
- to 3
to DISTP.NCC FRCM COTTCM CT HERTED ICNC (CMI
~g9o]g
- ]g'S' nAk RL
PCI FAILURE BY STRESS-RUPTURE STUDY BY KASSNER, ANL, BEGUN IN FY 80 o
EXAMINATION OF PCI FAILURE IN SPENT FUEL CLADDING BY o
LOADING SPECIMEN WITH EXTERNAL PRESSURE AND INTERN EXPANDING MANDREL IN AUT0CLAVES USING HIGH-TEMPERATURE STRAIN GAGES ON EXTERIOR o
OVER " CRACK" 0F EXPANDING MANDREL TO SENSE Il00P INITIATION AND GROWTH RATE OF GROWING PCI CRACK TESTS IN FY 80 AND FY 81 SHOULD HAVE ESTABLISHED ST l
o RUPTURE FAILURE CURVES WITil00T STRESS CORR 0 DANT TESTS IN FY 82 TO EXAMINE FAILURE WITil STRESS CORR o
PRESENT BWR SPEllT FUEL CLADDING WILL BE EXAMINED AS WE i
o PWR MATERIAL (H. B. ROBINSON, MAINE YANKEE, OCONEE) 4.
9 STRAIN-RATE RAMPIl1G TO PCI FAILURE EX-PIL 4
STUDY BEGUN IN FY 81 BY P. PANKASKIE, BNWL.
OFIT o
0BJECTIVE IS TO OBTAIN DATA FOR ESTA MODEL 0F PCI FAILURE, DEVELOPED FOR CPB/flRR IN FY o
SPECIMEN CONSISTS OF TUNGSTEN WIRE CENTERLIN 2
ANNULAR PELLETS, ZIRCALOY FUEL CLADDING, E o
HEATER WILL BE RAMPED IN POWER TO VARIOUS RATES, POWER INCREMENT BETWEEN HA IN A LOOP.
CAN ALSO BE USED FOR TIME TO FAILUR FAILURE DETERMINED.
PRESELECTED LOADING PAST HARD CONTACT.
E INITIAL STUDIES TO BE WITH UNIRRADIATED CLA CLADDING, AND WITH AND WITHOUT STRESS-CORR 0 DAN o
4 h
STRAIN-RATE RAMPING TO PCI FAILURE IN-PILE:
DEM0-RAMP PROGRAM NRC PARTICIPATING IN DEM0-RAMP PROGRAM AT STUDSVIK ON H o
FUEL.
SELECTED PP.E-IRRADIATED FUEL RODS TO BE POWER-RAMPED IN THE o
REACTOR AT STUDSVIK TO DETERMINE POWER INCREMENT OR TI IN FUEL RODS AT ABOUT 25 MWDh U BURNUP.
DATA WILL BE COMPARED WITH THAT FOR LOWER BURNUP RODS OF o
AND INTERRAMP PROGRAMS.
Tills PilASE OF STUDY COMPLETED IN JUNE 1981.
o i
4 6
b O
vc
STRAIN-RATE RAMPING TO PCI FAILURE IN-PILE:
PBF-0PTRAN TESTS PBF-0PTRAN TESTS DESIGNED TO EXAMINE PELLET-CLADDING INTERACTION DURING RAMPING CAUSED BY VARIOUS SCENARIOS OF OPERATIONAL TRANSIENTS LIKELY LWR POWER PLANTS DATA DETERMINED WILL INCLUDE ONE OR MORE OF THE FOLLOWING:
INCREMENT OF POWER FR0f1 BASE TO HARD CONTACT BETWEEN PELLET AND CLADDIN INCREMENT OF POWER FROM HARD CONTACT TO CLADDING FAILURE CLADDING FAILURE AS FUNCTION OF RATE OF RAMPING l
TIME TO FAILURE AS FUNCTION OF POWER INCREMENT AFTER HARD CONTACT EFFECT OF BURNUP EFFECT OF REPEATED CYCLING BELOW RAMP FAILURE LIMIT.
SCHEDULE OF TESTS:
FIRST OPTRAN TEST PLANNED FOR 1930 FOUR OPTRAN TESTS PLANNED FOR FY 1981 i
OPTRAN TEST MATRIX COMPLETED IN FY 1982 TOTAL OF SEVEN OPTRAN TESTS PLANNED, SIX flX, ONE 9-ROJ BUNDLE I
O 1
i
FISSION PRODUCT TRANSPORT ANALYSIS - TRAP CODE -
TO DEVELOP A MECilANISTIC COMPUTER CODE TO OBJECTIVE:
PRODUCT TRANSPORT BEllAVIOR WITilIN THE PRIM t
SYSTEM AHD CONTAINMENT.
PRIMARY SYSTEM MODEL ESSENTIALLY COMPLETE.
STATUS:
RFP ISSUED FOR ADVANCED CODE.
DEPOSITION OF FISSION PRODUCTS WITlilN REACTOR COOL ACCOMPLISilMENTS: SYSTEM UNDER CORE MELT ACCIDENT CONDIT UNIMPORTANT.
GROWTH OF AEROSOLS WITHIN RCS IS IMPORTANT.
IMPROVE TRAP CODE MODELS, EXTEND TRAP CODE TO MODEL CONTAINMENT FISSION FUTURE PLANS:
SOURCE TERM MODELLING, SENSITIVITY ANALYSIS, DEFINE VERIFICATION TEST FACILITY FUNCT FY 80-83 -- 10-12 MAN-YEARS FUNDING:
g
$s u
SEPARATE EFFECTS TESTS FOR TRAP CODE - SANDIA OBJECTIVE:
TO PROVIDE BASIC DATA ON FISSION PRODUCT COMP 0UND VAPOR PRESSURES AND CilEMICAL INTERACTIONS IN A HIGH TEMPERATURE STEAM ENVIRONMENT TO SUPPORT DEVELOPMENT OF Tile TRAP CODE.
STATUS:
VAo0R PRESSURE EXPERIMENTS IN PROGRESS AT SANDIA LABORATORIES AND AT THE NEW MEXICO INSTITUTE FOR MINING AND TECllN0 LOGY -
COMPOUNDS OF CESIUM AND IODINE BEING INVESTIGATED.
FISSION PRODUCT REACTION SYSTEM (FPRS) APPR0XIMATELY 60%
COMPLETE.
FUTURE PLANS:
VAPOR PRESSURE TESTS ON OTHER FISSION PRODUCT COMPOUNDS WILL BE CONDUCTED AS NECESSARY.
FPRS WILL BE COMPLETED AND TESTING INITIATED.
i NON-INTRUSIVE REAL TIME FISSION PRODUCT COMP 0UND IDENTIFICATION BY LASER RAMAN SPECTROSCOPY WILL BEGIN IN Tile FPRS APPARATUS.
FUNDING:
FY 80 - 150K FY 81 - 210K q
\\1 bd
FISSION PRODUCT VAPOR DEPOSITION EXPERIMENTS - BCL TO PROVIDE EXPERIMENTALLY DERIVED FISSION PRODUCT DEPOS OBJECTIVE:
AT HIGH TEMPERATURE ON PRIMARY SYSTEM SURFACES TO AID Tile TRAP CODE.
TO DETERMINE THE NATURE OF Tile INTERACTION BETWEEtt VARIOUS FISSION PRODUCT COMP 0UNDS AND PROT 0TY 4
CONSTRUCTION OF FISSION PRODUCT VAPOR DEPOSITION APPARATU STATUS:
COMPLETE.
STAINLESS STEEL AND INCONEL DEPOSITION COUPONS HAVE BEEN TO SIMULATED PRIMARY SYSTEM AGING.
10 DINE VAPOR DEPOSITION EXPERIMENTS HAVE BEEN INITIATED AN BE COMPLETE IN APPR0XIMATELY 4 MONTHS.
FUTURE PLANS:
PROGRAM TO BE COMPLETED FY 80.
CESIUM AND TELLURIUM VAPOR DEPOSITION EXPERIMENTS WILL B APPR0XIMATELY 4 MONTilS.
DATA WILL BE ANALYZED AND MODELS DEVELOPED FOR INCORPORAT INTO TRAP CODE.
FullDillG:
FY 79 (PART OF TRAP DEVELOPMENT PROGRAfD FY 80 - 95K FY 81 <
- ~~~~17
STEAM GENERATOR TUBE RUPTURE 10 DINE TRANSPORT - BCL
~
OBJECTIVEi TO DEVELOP MECHANISTIC COMPUTER MODELS FOR 10 DINE TRANSPOR WITillN THE STEAM GENERATOR AND SECONDARY SYSTEM UNDER SGTR ACCIDENT C0llDIT10NS.
TO EXPERIMENTALLY DETERMINE Tile AMOUNT OF ATOMIZATION OF Tile PRIMARY C00lANT DURING BLOWDOWN INTO Tile SECONDARY SYSTEM.
PROJECT STATUS:
DESIGN.0F THE EXPERIMENTAL FACILITY TO MEASURE PRIMARY COOLANT ATOMIZATION IS COMPLETE AND CONSTRUCTION IS UNDER THE 10 DINE TRANSPORT MODELS HAVE BEEN DEVELOPED AND ARE ASSEMBLED INTO A COMPUTER CODE.
FUTURE WORK:
PROJECT WILL BE COMPLETED Ill FY 80.
Tile AMOUNT OF ATOMIZATION ATID DROP SIZE DISTRIBUTION WILL BE MEASURED AS A FUNCTION OF PRESSURE DIFFERENTIAL (100 - 1300 Tile SGTR 10 DINE TRANSPORT COMPUTER CODE WilL BE COMPLETED A DELIVERED-TO NRC/llRR.
g FUNDING:
FY 79 - 70K FY 80 - 63K
FISSION PRODUCT RELEASE FROM LWR FUEL - ORill OBJECTIVE:
TO DETERMINE THE QUANTITY, SPECIES AND CHEMICAL FORM 0F FISSION PRODUCTS RELEASED FROM DEFECTED FUEL RODS UNDER ACCIDENT CONDITIONS.
STATUS:
PROGRAM COMPLETE.
i e
b
.N
^
FISSION PRODUCT RELEASE FROM LWR FUEL - HIGH TEMPERATUR TO EXPAND THE INVESTIGATION OF FISSION PRODUCT RELE OBJECTIVE:
DEFECTED LWR RODS WITHIN THE TEMPERATURE RANGE OF 1750*C.
189 RECEIVED - PROGRAM START AWAITING SUPPLEMENTAL STATUS:
AUTHORIZATION.
FY 80S - 365K -- FY 81 1100K -- FY 82 - (575K)
FU!lDING:
N Q
W
FISSION PRODUCT RELEASE FROM LWR FUEL - HIGil TEMPERATURE SCOPE:
MEASURE RELEASE FROM 12 BWR AND PWR RODS UP TO ~1750*C.
DETERMINE Cs, KR, Ru, As, SB, AND Eu BY GAMMA-SPECTROSCOPY.
DETERMINE I BY NAA.
RATIONALE:
CAN USE. EXISTING APPARATUS WITil MINOR CilANGES TO REACH ~1750*C.
~
LIMITATION:
MAXIMUM TEMPERATURE IS ~1750*C.
t vA
CIIARC0AL FILTER 10DIllE RETEllTION PERFORMANCE - IIRL OBJECTIVE:
T0 INVESTIGATE Tile PERFORMANCE OF ACTIVATED CHARC0ALS IN RLM0VING AIRBOR'IE RAD 1010 DINE UNDER LWR ACCIDENT i
CONDITIONS.
TO ASSESS Tile EFFECTS OF IN-SERVICE WEATHERING AND EXPOSURE TO CONTAMINANTS ON Tile REMOVAL AND RETENTION OF RADIOI0 DINE.
189 RECEIVED - PROGRAM INITIATION AWAITING SUPPLEMENTAL STATUS:
FUNDING AUTHORIZATION.
PROGRAM ELEMENTS:
EXPOSE SAMPLES OF COMMERCIALLY AVAILABLE ACTIVATED AND OTilER WIDELY CilARC0ALS (IMPREGilATED WITH TEDA, Klx USED IMPREGilANTS) TO WEATHERING AND TO KNOWN ATM0SPilERIC CONTAMINANTS.
TEST TilESE CilARC0ALS FOR RAD 1010 DINE RETENTION UNDER Tile RANGE OF SEVERE ACCIDENT CONDITI0ilS INCLUDING:
A.
EXPECTED RADIOI0 DINE LOADINGS, B.
TOTAL RADIATION LOADlHG, AND C.
EXPECTED TEMPERATURE AND llUMIDITY ERVIRONMENT.
FUNDING:
FY 80S - 110K -- FY 81 - 115K -- FY 82 - 120K p
4
~
~
FISS10ll PRODUCT RELEASE - MELTING FUEL TO EXPERIMENTALLY DETERMINE THE RELEASE OBJECTIVE:
PRODUCTS FROM IRRADIATED LWR FUEL IN Til RAllGE '1800*C TO 2800*C.
PROGRAM UNDER EVALUATION.
STATUS:
CONSTRUCT A FACILITY CAPABLE OF TRANSIEN PROGRAfl ELEMENTS:
COMMERCIALLY IRRADIATED FUEL R0D SEGMENT ENVIRONMEllT.
IN A STEAM OR STEAM /ll2 CONDUCT EXPERIMENTS TO K ASURE Tile RATE, QUANTITY, SPECIES AND CllEMICAL FORM 0F RELEASED FIS UNDER llIGH TEMPERATURE INCIPIENT FUEL ME FUNDING IDENTIFIED FOR FY 83 AND BEYOND IF FUNDIllG:
PROGRAM IS ESTABLISilED.
D, 1
O
FISSION PRODUCT LEACHING OBJECTIVE:
TO EXPERIMENTALLY INVESTIGATE THE LONG TERM RELEASE OF FISSION PRODUCTS FROM SEVERELY DAMAGED FUEL RODS UNDER THE PHYSICAL AND CilEMICAL CONDITIONS EXPECTED WITHIN THE PEACTOR VESSEL FOLLOWING A SEVERE ACCIDENT.
STATUS:
PROPOSED PROGRAM UNDER REVIEW.
PRELIMINARY JUDGEMENT IS TO NONSUPPORT.
4 b
s
PISS 10N PRODUCT TRANSPORT VERIFICATION FACILITY
'0BJECTIVE:
TO CONSTRUCT OR MODIFY AN EXISTillG FACILITY FOR THE PURPOSE OF TESTING THE VALIDITY OF CURRENT LWR FISSION PRODUCT TRANSPORT CODES SUCH AS TRAP-MELT, CORRAL, NAUA, ETC.
EMPHASIS WILL BE ON CONTAINMENT FISSION PRODUCT BEHAVIOR.
HOWEVER PRIMARY SYSTEf1S EFFECTS WILL ALSO BE IKCLUDED.
STATUS:
PROGRAM UNDER EVALUATION, PROGRAM ELEMENTS:
DEFINE FUNCTIONAL DESIGN REQUIREMENTS FOR A FISSION PRODUCT TRANSPORT CODE VERIFICATION TEST FACILITY.
INVESTIGATE THE CAPABILITY OF EXISTING FACILITIES, SUCH AS STCF, NSPP, ETC., IN MEETING THESE REQUIREMENTS.
CONSTRUCT (0R MODIFY) A FACILITY TO PERFORM TESTS.
CONDUCT FISSION PRODUCT BEllAVIOR TESTS IN PROTOTYPIC ACCIDENT ENVIRONMENTS.
D d
FUNDING:
FY 82 - 1000K 9
l TMI FISSION PRODUCT RELEASE DATA EXAMINATIO;l
~
l OBJECTIVE:
TO PROVIDE FullDING FOR FISSION PRODUCT RELEASE AND TRANS DATA GATilERING ACTIVITIES AND ANALYTICAL SUPPORT DURING T RECOVERY.
JOINT NRC, DOE, EPRI, GPU DATA GATHERING ACTIVITY UNDERWAY.
STATUS:
DOE ilAS COMMITTED TO PROVIDE GOVERNMENT SHARE OF FUNDING FUNDING INDICATED BELOW REPRESENTS A CONTINGENCY FOR DATA ACQUISITION AND ANALYSIS NOT AGREED TO BY JOINT COMMITTEE (AND NOT FUNDED BY DOE).
FUNDING:
FY 80S - 175K -- FY 81 - 85K -- FY 82 - 200K i
m W
l
CLASS 9 ACCIDENT RESEARCH:
PROGRAM LOGIC CLASS 9 ACCIDENTS CHALLENGE CONTAINMENT. PROGRAM OBJECTIVE:
DETERMINE BEST ESTIMATE OF RISK; 2
ANALYSIS AND ASSESSMENT OF SPECIAL FEATURES.
NATURE OF CHALENGES IDENTIFIED IN WASH-1400:
1.
CAN PRESSURES IN PRIMARY SYSTEM BREACH THE SECONDARY?(EVENT V; SG. TUBE RUPTURE) 2.
CAN A MELTED DOWN CORE BREACH THE PV AND DVER-LOAD THE CONTAINMENT?
(DEBRIS BED C00LABILIT(; STEAM SPIKE) i 3.
CAN A HYDR 0 GEN EXPLOSION BREAO{ THE CONTAINMENT?
(HYDR 0 GEN LOADS; HYDROGEN CONTROL; CONTAINMENT RESPONSE) 4.
CAN A STEAM EXPLOSION BREACH THE CONTAINMENT?(EXPLOSION EFFICIENCY; PV LOADING) 5.
CAN A HOT CORE MELT THE BASEMAT?
(CORE CONCRETE INTERACTIONS; CORE CATCHERS) 6.
CAN THE CONTAINMENT SLOWLY HEAT UP AND BE OVER PRESSURIZED?
(AUXILIARY SPRAYS; FVCS) 7.
CAN MAINTENANCE OF VITAL FUNCTIONS BYPASS CONTAINMENT OR THREATEN ITS INTEGRITY?
y 8.
CAN FAILURES IN I & C C0WROMISE SAFETY SYSTEMS?
ll
. - -