ML20079F539
| ML20079F539 | |
| Person / Time | |
|---|---|
| Site: | Clinch River |
| Issue date: | 06/01/1982 |
| From: | Longenecker J ENERGY, DEPT. OF |
| To: | Check P Office of Nuclear Reactor Regulation |
| References | |
| HQ:S:82:039, HQ:S:82:39, NUDOCS 8206080034 | |
| Download: ML20079F539 (61) | |
Text
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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:039 l
JUN 01 1993 l
l Mr. Paul S. Check, Director CRBRP Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Check:
RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION - CORE PERFORMANCE J
Reference:
Letter, P. S. Check to J. R. Longenecker, "CRBRP, Request for Additional Information," dated March 25, 1982 This letter formally responds to your request for additional information contained in the reference letter.
l Enclosed are responses to Questions CS 490.1 through 10, 12, 14 through 23, 25 through 34, and 36 through 39. The responses to questions CS 490.11, i
13, 24, and 35 will be supplied by June 4,1982. These responses will also be incorporated into the PSAR Amendment 69; scheduled for submittal later in June.
- incerely, C*: PAC tn JohnR.Longenept,er, Manager Licensing & Environmental Coordination Office of Nuclear Energy Enclosures cc: Service List Standard Distribution Licensing Distribution 0I p0 8206080034 820601 PDR ADOCK 05000537 A
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Paga - 1 [8,22] #62 l
l l
Ouestion CS490.1 The plutonium concentration in the mixed oxide driver pins has been changed from 20 and 25% in the FFTF and the previous CRBR design to 33% in the current CRBR design. This gives rise to concern over whether any of the data base on Integral fuel pin perf ormance or on f ueled-cladding behavior is relevant to the current design of CRBR. Specific concerns include:
- 1) How does the change of Pu concentration ef fect f uel cladding chemical Interaction? What is the basis f or thic assessment?
- 2) The thermal conductivity and solldus and liquidus temperatures of the f uel (plus probably other phenomena that enter into i-hermal perf ormance) are af fected significantly by the change in Pu concentration.
it seems inescapable, theref ore, that the power-to-melt tests and the thermal perf ormance models based thereon do not apply to the revised CRBR fuel design.
If the applicant agrees with this assessment, how does he plan to If he replace these two key pieces of the fuel design evaluation methods?
does not agree with this assessment, he is requested to justify his position.
- 3) How does the change of Pu concentration af fect the applicability of Is it properties and models that are dependent on stoichiometry?
anticipated that hitherto unimportant or unsuspected ef fects of Pu redistribution will become significant? How is it anticipated that these changes will af fect f uel perf ormance? What is the basis f or this assessment?
- 4) How does the higher concentration of Pu af fect the fission gas retention characteristics of the f uel?
If significantly different, how 40es this af fect the applicablilty of the f uel pin evaluation models? How does the changed Pu concentration af fect both time dependent and time independent def ormation behavior of the fuel? How will this af fect f uel perf ormance?
What are the bases f or both answers?
If an assessment is not po'ssible now, how does the applicant propose to resolve the issue?
- 5) How does the change of Pu concentration af fect f uel swelling as a f unction of burnup? What is the basis f or this assessment?
s QCS490.1-1 Amend. 69 May 1982
Pzg3 - 2 f,8,22] #62 l
l 6)
Why doesn't this change Invalidate both the CDF and the Ducility Limited Strain models f or evaluating f uel perf ormance?
If it does Invalidate both models, how does the applicant plan to evaluate f uel pin perf ormance?
7)
It would seem to be a minimum requirement that some check tests of 33% Pu concentration f uel pins be perf ormed in FFTF and some of those pins be given transient tests in TREAT to confirm the predicted ef fects of the higher Pu concentration.
What plans does the applicant have for such tests? Please be specific In the response.
If there are no such plans, how does the applicant plan to justify his assessment of the ef fect of the change? Are there any data at all on the behavior of irradiated mixed oxide f uel with this high a concentration of Pu?
Resoonse The question addresses the ef fects of switching from 25% Pu in depleted Uranium to 33% Pu In depleted uranium, in general, the response to the question,Is as follows:
The increase in plutonium enrichment In CRBRP will result in a very slight
~
increase in the free oxygen due to the slightly higher ratio of plutonium-to-uranium fissions. The previcus CRBRP plutonium enrichment was ap roximately 25% Pu (or 75% U238) whlge the new enrichnent is 33% Pu (or 67%
8),
'73 Because the ratio fission rate to Pu fission rate is so low, the difference in fir 3 ducts produced between 25% Pu and 33% Pu f uels wil I not be significan,.
A slightly lower fuel melting temperature and a decrease in thermal conductivity will result from the increased plutonium content.
The question Implies concerns regarding the ef fects of extrapolating from the EBR-il data base (25% in 20-80% enriched uranium) to the CRBRP enrichment (33%
Pu in depleted uranium). The response is as follows:
The CRBRP f uel will have nearly all fissions take place in plutonium whereas the EBR-il data base had a high percentage of uranium fissions.
Since the uranium fissions produce more zirconium fission product, which acts as a getter for oxygen, and CRBRP has essentially no uranium fissions, CRBRP is expected to have increased free oxygen in the f uel pins, w
s QCS490.1-2 V
Amend. 69 May 1982
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Pcge - 3 [8,22] #62 5
The potential consequences to CRBRP of a slightly higher oxygen potential are; slightly increased f uel cladding chemical Interaction (FCCl), enhanced Pu and 09 migration, enhanced fission product migrat!on, and increased f uel-sodlum rbaction. However, based on the results of the ANL-08 and P-15 experiments which had mostly plutonium fissions, the ef fects of the fission product yleid has little or no ef fect on f uel pin perf ormance. A programmed startup is needed at the beginning of a cycle to mitigata potential thermal perf ormance degradation.
1) increasing the Pu content in the CRBRP f uel from 25% to 33% did not change the fission rate per unit volume.
Since nearly all fissions occur in plutonium in both the lower and higher Pu f uels, the fission product inventory In both cases is nearly identical. FCCI is primarly influenced by fission product inventory in this case and therefore, the af fect of the higher plutonium concentration is insignificant.
Preliminary results from the ANL-08 experiment, which included f uel rods with 30% and 40% Pu f uel, showed that FCCI to be similar or less severe than 25% Pu f uel.
2) increased Pu concentration results in slightly decreased thermal conductivliy and decreased solidus and liquidus temperatures. These properties have been well established, and f uel perf ormance analysis codes, such as LIFE, account for these variations. Although the codes were calibrated using the 25% Pu f uel data, the f undamental nature of the thermal perf ormance models, and the availability of the basic data, reduces the uncertainty In extrapolation to higher Pu content f uels.
In tddition, power-to-melt data will becomo available from the DEA-2 test which contains some 29% Pu f uel rods. Additional specific tests are ongoing f or both FFTF reloads with higher plutonium (29%)
pioonow or content, including DE-9, D9-3, and CRBRP (33% Pu) prototype tests (3 assemblies), and EBR-il Operational Reliability Testing (ORT) program.
~
The ef fect of Pu content on restructuring will be available from the ANL-08 post-f rradiation examination results. Preliminary Information Indicates that the restructuring behavior is not dissimilar to 25% Pu fuel.
In addition, Information will be available from the af orementioned tests.
QCS490.1-3 Amend. 69 May 1982
Pqge-4[8,22]#62
- 3) The CRBRP f uel will have the same as-fabricated 0/M range as in the current EBR-ll and FFTF oxide f uels.
Increased plutonium concentration will result in a slightly higher increase in 0/M with burnup compared to 25% Pu f uel. The properties and models that depend on stlochlometry are primarily f uel creep, thermal conductivity and melting temperatures, which are already included in the property correlations and models in the current analytical codes. The available preliminary Information on high Pu content f uels (ANL-08) does not Indicate any significant dif ference in the Pu redistribution from that observed in 25% Pu f uel. Additional confirmation of this will be obtained in the aforementioned test and in particular, In CRBRP prototype tests in FFTF, which are scheduled for Irradiation starting FFTF cycle 2.
Based on the ANL-08 experiment fission gas release and f uel def ormation data mentioned earlier, it is not expected that gaseous swelling will be af fected significantly.
Anothe phenomenon which depends on stoichiometry is f uel-sodlum reaction in the event of a cladding breach. Since the increase in f uel 0/M with burnup is more rapid compared to 25% Pu f uel, there will be more oxygen avail able f or f uel-sodi um reaction. An assessment of the ef fects can be made based on the current RBCB test results, which included very high burnup (11 to 15%) fuel pins. The oxygen available in these high burnup pins is higher than that in a CRBR pin at goal burnup. The disneter increases due to the f uel-sodium reaction were within the acceptable range. The current ORT /RBCB program should provide additional inf ormation f or evaluating the ef fects of higher Pu content.
- 4) Currently there are no fission gas retention data on high Pu f uels.
However, based on preliminary fission gas release data from ANL-08 no significant dif ferences in retention behavior are expected compared to the 25% Pu f uel.
Data on def ormation behavior will become available when all the ANL-08 post-irradiation examinations are completed and results are evaluated.
Based on ANL-08 preliminary f uel rod profilametry information no significant dif ference in def ormation behavior is expected.
How ever, 0C5490.1-4 Amend. 69 May 1982
Pcge - 5 [8,22] #62 l
f urther evaluation of ML-08 information is planned. Additional data wilI become available f rom the FFTF DE-9, D9-3, and CRBRP prototype tests.
- 5) The change in the Pu concentration to a higher value wilI quite possibly reduce f uel swelling, since the fission product yleid of zirconium will decrease slightly. Zirconium is readily oxidized and goes into solution thereby contributing to the f uel swelling. Reduction of fuel swelling is beneficial since Fuel Cladding Mechanical Interaction (FCMI) would then be reduced. However, the ML-08 preliminary test results have not yet confirmed this point f.e., rod deformation appears very similar).
- 6) The COF and ductility limited strain models are applied to the cladding behavior. Of course, the cladding loading will be from fuel swelling and fission gas.
Having a dif ferent plutonium f uel concentration does not invalidate the models now used for evaluating cladding behavior, since there are no dif ferences that have thus f ar been noted in the deformation or swelling behavior based on the ML-08 preliminary Information (4 and 5 above). The current ML-08 data do not show any dif ferences in the total cladding loading due to fuel swelling and fission gas with 30-40% Pu fuel compared to 25% Pu f uel.
If subsequent testing did show significant deformation / swelling dif ferences, it might be necessary to alter some of the cladding loading values that input to the CDF and ductility limited strain models, but the COF and ductility limited strain models would still be used to evaluate f uel rod cladding perf ormance.
- 7) The planned tests to verify the behavior of high plutonium content In addition to the ML-08 test are:
QCS490.1-5 Amend. 69
)
Page - 6 [8,22] #62 1g11 Pu Content Reactor Remarks CRBR-3 33%
FFTF cycle 2 start l
CRBR-5 33%
FFTF cycle 5. start D9-3 29%
FFTF cycle 1 start DE-9 29% FFTF FFTF cycle 2 start DEA-2 29% Reload FFTF Test Complete PIE in progress CRBR-Transiont 33%
FFTF/ TREAT in Planning EBR-i l-Transient 33%.
EBR-II/ ORT program In Planning EBR-i l-RBW 33%
EBR-ll/ ORT program in Planning The testing mentioned abose with high plutonium content is either in place or in planning stages. The steady-state and transient program plans wil I be provided to NRC via a summary description document in September,1982 which wilI include the of feet of higher Pu content.
QCS490.1-6 Amend. 69 May 1982
e:ge ~ t Lo,LLJ 9 3 OuestIon 0C5490.2 The current data base for fuel pin response and cladding f ailure threshold under transient overpower (TOP) conditions includes no data at alI in the rap range f rom the power-to-melt tests (about 0.005 cents /s) to the W-2 test (about 5 cents /s), and very little data for ramp rates between 5 cents /s and 50 cents /sec. Please delineate the testing planned to provide data in the cited rap range.
If no testing is anticipated in the slow rap rate range, how is it planned to determine what the cladding f ailure threshold range, how is it planned to determine what the cladding-failure threshold is and what the threshold Is for molten f uel expulsion (not necessarily the same)?
Reinonic Fuel rod response and f ailure thresholds have been determined for transients overpower (TOP) conditions of 50 /see to $3/sec.
Slow re p rate tests of 0.01 to 10 /sec. are planned to the EBR-il Operational Reliability Testing (ORT) progre.
In addition, the of feet of Intermittent slow rap rate transients (side by side tests with and without) are planned, as well as the of fect of transients on tight bundles. The TOPl-1 (A, B, C, D) tests wilI explcre rep rate and transient overpower on pre-f rradiated f uel rods to define breach limits. The TOPI-2 test is a transient overpower test on a pre-f rradlated breached rod. TOP-4 (A, AA) (B, BB) are side-by-side tests (steady-state and steady-state plus periodic TOP) in a vehicle capable of reconstitution and utilize agressive rods designed to breach at mid-life.
TOP-7 Is a duty cycle test (alternating 1000 hour0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> periods at 100 and 70% power on aggressive rod designs). TOB-10 is a tight bundle test with steady state cp r.,; ; ;.,r. cr.c p cr i od i c TOP. The stesdy state and transient program plans wil l be available In a summary description document by the end of FY82 including a general description of the EBR-il ORT progrm.
The CRBRP methodology considers the response of cladding to slow rap rate transients by modelling slow strain rate, hold time and annealing of fects In the cladding.
As well-characterized slow rap rate data become available, they will be used to quellfy the analytical methodology.
QCS490.2-1 Amend. 69
P:ge - 8 [0,22] 162 Ouestion CS490.3 f
Current state-of-the-art analysis methods use stress-rupture based correlations for predicting f ueled and unf ueled cladding breach under both steady-state and transient conditions. There appears to be a lack of fundamental understanding of cladding f ailure mechanisms as evidenced by:
1)
Very large dif ference in load bearing capability and ductility of fuel cladding between steady-state and transient conditions, yet stress-rupture formulations are being esed for both classes.
- 2) Hints of fission product assisted stress cracking propagation.
- 3) The elimination of much of the damage with regard to transient capability when steady state Irradiated above 10500F.
What testing plans are there to better identify the mechanisms of cladding f ailure, to define how steady-state and translent behavior mesh together and to develop more appropriate f ailure criteria, particularly under overpower conditions.
Resoonse it is acknowledged that the use of stress-rupture type correlations for the analysis of transient events appears inappropriate. Current CRBRP methodology employs stress-rupture correlations for the analysis of steady state performance only.
On the other hand, the analyses of transient performance employ correlations which are based on short term f ailure data, e.g., tensile, fuel cladding transient testing.
Furthermore, these correlations consider the of facts of fuel-adjacency as welI as the reduced irradiation of facts at elevated steady state temperatures.
Planned testing progrens which will enhance our understanding of transient f ailure mechanisms include TREAT and EBR-il transient prograns and FCTT testing. The steady-state and transient progran plans wilI be provided to the NRC via a summary description document before the end of FY82.
QCS490.3-1 Amend. 69 May 1982 i
Pcg3 - 9 [8,22.J #62 4
Ouestion 0C5490.4 Continuing questions in predicting and understanding f uel pin response to overpower condiff ons are the ductility and load bearing capability of fueled 4'
Irradiated cladding under fuel cladding mechanical Interaction (FCMI) c::nditions, and whether cladding response to these conditions is significantly different than exhibited under gas pressure loading.
What testing is planned to define response to FCMI transient loading? Please forward whatever data cay exist in this area, along with the project evaluation of the data.
Reseense The fuel cleading mechanical Interaction (FCMI) type loading conditions are being tested on fuel rods in the TREAT and FCR-ll Operational Reliability
- ^*
Tcsting (ORT-Transient) programs.
p The fuel rod cladding performance evaluation under FCMI loading conditions rcsulting f rom short term overpower events are based on cladding tests made under gas pressure (load controlled) loading in the Fuel Cladding Transient Tcsting (FCTT) facility at HEDL.
It is assumed that gas pressure loading of cladding results In equal or lower f ailure loads and lower or equal average strains than FCMI (strain controlled) type loadings. The reason for this
'C' assumption is the capability of the FCMI loading system to bridge weak spots
-~
In the cladding, that is, local bulging does not develop in the FCMI strain controlled loading system.
In a gas loading system e developing local bulge
'6 MN continues to be loaded by the same pressure; the cladding flows until plastic I nstat,i i i s y at.J iol i ure ou.urs.
HEDL has performed exploratory tests on unirradiated cladding which was loaded in a Fuel Claading Mechanical Interaction Mandrel Loading Test (FCMI/MLT) system. As a result of these test data, strain controlled loading generates equal or larger average hoop strains at f ailure than gas loading. Because of this, minimal testing is presently planned to determine the dif f erence in cladding f ailure due to FCMI and gas loading. However, the two cladding loading systems may generate dif ferent cladding f ailure modes cladding QCS490.4-1 m '. 69 Amend. 69 r~w v.p sy ~EC2
P ge - 10 [8,22.] #62 bulges and pin holes under gas loading, and cladding rips under FCMI type loading.
t r.:
t.
The mandrel test results are not being used to support the CRBRP fuel design c
for the above reasons.
i QCS490.4-2 Amend. 69 May 1982
Pcge - 11 [8,22] #62
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Question DCS490.5 Thcre appears to be evidence that cladding ductility and. load bearing capability is much less af f ected by irradiation at temperatures above 1050 to 1100 F than by irradiation below that temperature level.
This implies that und:r transient overpower conditions the site of the cladding breach, if it occurs, is virtually certain to be below the axial level of the transition regi on. This may, depending on inlet coolant temperature er.d coolant flow reto, force the site of the f uel expulsion in an undetermined reactivity ins 2rtion accident low enough that the fuel movement causes substantial additional reactivity insertion, significantly exacerbating the accident.
Picase provide:
- 1) The data which support the cited cladding behavior:
- 2) A comparison of irradiated cladding ductility and strength above and below the transition temperature;
- 3) The best estimate of the transition temperature and 'its uncertainty, and the range of temperature involved in the transition; 4) three steam (high side) estimates of the f ractions of fuel and blanket pins where in the transition temperature occurs on the cladding below -
X/L = 0.75, 0.65 and 0.55.
What would be the impact of lowering the inlet coolant temperature by 50 to 100 F If this were necessary to avoid the transition temperature being reached at too low an axial position.
I Resoonse 0
Tho 1000-1200 F transition in the ef fects of Irradiation on the cladding's tensile properties is known to the applicant and is incorporated in the current correlations which describe these properties; relevant data are given in the Nuclear Systems Materials Handbook.
The low temperature region which is most af f ected by irradiation is chgractarivad by en irradiation induced increase in strength; as such, op: ration in this region is not necessarily detrimental.
QCS490.5-1 Amend. 69
_% mm
P2ge - 12 [8,22] #62 The location of a potential breach site is determined by the interaction of a number of f actors. These include the axial distributions in:
(a) cladding stress, (b) temperature, and (c) the phenomena which diminish the cladding's load bearing capability, e.g.,
Interstitial loss, sodium corrosion, annealing, etc. Typically, these f actors combine to yleid a potential breach site which is located at or above core midplane; indeed, this location bears little or no intrinsic dependence on the axial position of the transition in the claddings sensitivity to Irradiation.
In view of the above, and since the transition phenomenon is modeled in the OlBRP methodology, there is no basis for concern related to axial position of the transition point nor to consider lowering core operating temperature.
1 I
QCS490.S 2 Amend. 69
.WM
Pzgo - 13 [8,22] #62 OuestIon OCS490.6 Non-prototypic factors in TREAT transient overpower (TOP) tests of EBR-il Irradiated f uel pins seriously compromise translation of the results of these l
tests to the CRBR.
What plans are there to evaluate those f actors experimentally, particularly radial power depression and short vs. long pins, and now the additional f actor of 33% Pu concentration as versus the 25% Pu concentration of the tests? Other factors include non-prototypic fluence to burnup ratio and U235 to Pu fission ratio, preconditioning, and static capsule non-prototypic cladding temperature. The ratio of U235 fissioning to Pu fissioning is of concern because 305 more Zr fission product would be produced from Pu fissioning and might adversely af fact f uel cladding chemical Interaction through its of fact on oxygen potential, if there are no tests planned to evaluate the of facts of the non-prototypic factors in the data base, how does the applicant plan to account for these f actors in applying the data base to CRBR design evaluation?
Resoonse Consideration of non-prototypic f actors in TREAT transient overpower (TOP) tests is of concern to CRBRP and to the f ast breeder community.
Because of this concern, plans have been made to investigate non-prototypic factors.
Present plans call for TREAT and EBR-ll Operational Reliability Testing (ORT) programs. The of fact of radial power depression is to be Investigated by tests in the EBR-il ORT progran and in TREAT tests with neutron filters.
Shcrt ver: : !cng rod effacts wilI be investigated by utilizing prototypic rocs in TREA1. The 33% Pu concern and pre-conditioning wilI be investigated in TREAT and in the EBR-il ORT program. The U235 versus Pu fission ratio concern wilI be Investigated in TREAT transient and steady-state testing of FFTF and EBR-Il irradiated rods.
Slow ranp rate transient overpower (TOP) data will be Investigated in the EBR-il ORT progran. The steady-state and transient program plans wilI be available in a summary description document in September,1982, including a general description cf the EBR-il ORT progran.
l QCS490.6-1 Amend. 69 May 1982
'~~~~~~ ~ ~ ~ ~ ~ " ' ' ' ' "
P ge - 14 L8,hh.} Y6h" ~ ~ " ~ ~ ~ ~
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'OuestIon OCS490.7 Several of the EBR-ll pins that have been TREAT tested were Irradiated in subassemblies from which other pins have exhibited metallurgical evidence of f ar higher temperatures than can be accounted for during Irradiation by thermal hydraulic means. Please enumerate the TREAT tests that involved pins f rom such assemblies and the apparent temperature defect in each case. Please provide your evaluation of what of f act this may have on the results of the subject tests.
Response
Fuoi rods selected from the following EBR-il irradiation test essemblies were t sted in the HOP and HUT Series In TREAT.
PNL-9 in HUT PNL-10 in HUT PNL-11 in HUT P-14 in HUT P-15 in HUT PNL-17 in HOP P-23 in HOP ANL-A In HOP WSA-3 In HOP and HUT Rods used for TREAT testing, except for one (a P14 rod), were discharged f ran assemblies that prior to discharge or Interim examination did not exhibit f ailures attributed to local overheating.
Breached rods of subassemblies P14 and P23 did exhibit metallurgical evidence of hot spots.
It has been implied that these over-temperatures were the result of subassembly reconstitution or associated with special rod locations.
Since it is now known when and where the hot spots on rods occurred, they may have been present In other TREAT tested rods.
However, since hot spots are localized, the cladding exposed to both nominal and hot spot temperatures was subjected to TREAT tests with f ailure likely to occur at the weakest spot. No obvious dif ference exists between TREAT test results with rods from assemblies ehich later exhibited hot spot related f ailures and ottier TREAT test results.
l QCS490.7-1 Amend. 69 u-mgm
Page - 15 [8,22] #62 Ouestion CS490.8 The FCTT data are generated at constant load and increasing temperature.
Permanent straining occurs as the yleid strength decreased with increasing temperature. Thus straining and annealing are inextricably Intertwined in the data obtained. The data are probably relevant for loss-of-flow events; however, in overpower events, straining is more likely to l'e dependent on dif ferential expansion of the fuel against the cladding anc anly mildly dependent of cladding temperature.
What plans are there tc wrform FCTT tests in which the strain rate is Independently controlled't Pletsa supply whatever data may be available in this area.
Resoonse The FCTT test with Independently controlled strain rates.
Such tests were performed on unirradiated and some Irradiated cladding specimens by HEDL. A schedule of availability of all such topical reports in the f uel, control, and blanket assembly design area will be provided by the end of FY82.
In addition, FCTT type tests with unirradiated and def ueled cladding were perf ormed at HEDL, not at controlled strain rates, but at constant temperature and increasing pressure and at variable ratios of pressure and temperature ramps. These tests were perf ormed to simulate the pressure-temperature histories similar to those referenced In the questions.
Figure QCS490.8-1 shows the temperatures, pressure-time histories of these tests, the number of tests and cladding fluence ranges.
QCS490.8-1 Amend. 69 May 1982
_c_
Rg.
(CS490.8.-l TEMPERATURE-TIME HISTORIES OF VARIABLE TEttPERATURE -
.~
_ PRESSURE FCTT TESTS 810 to 1030 PS1 s.
., f 10010 to 3720 PSI e
i e
f cartaw 9
' 980 to '1738*F j899 to 1533*F
.y i
TIME TIME TYPE A 20 TESTS TYPE B 10 TESTS
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..' e na.
- * * ' " ^
- p
e I
'T f790to1080*F y
f.co..,..,
'1275'F e'
L i
.-u-UME VME TYPE C 5 TESTS
. TYPE D 1 TEST
,.- eO.sf Aarf e
p r--* 8900 PSI
[
4151 PSI I
i - e""'"
1189'F
,154 9F,.
y
\\
l e-on TIME TYPE E 1 TEST TYPE F 1 TEST 6-Fluences at Failure Points 22 6.2 to 11.2 x 10 n/cm2 ( p,,0.1 Me v)
Hoop t7.17 x P (psi) o G CS4W.t - 1.
ycge - lo LDeLLJ ICL Question OCS490.9 Essantially no data exist for either steady-state or transient performance of blanket pins.
What are the current testing plans to obtain blanket pin data?
If no tests are planned f or blanket pins under either or both steady-state or transient conditionr, how Is it planned to confirm predicted cladding f ailure thrcsholds and margic s to cladding f ailure?
Resoonse Three OBR-il blanket Irradiation test have been completed in support of the CRBRP blanket design:
o WBA Steady state test with one power jump.
The data report is in preparation.
o WBA Steacy state test the post Irradiation examination is complete. The documentation is in preparation.
o WBA Cyclic transient test in EBR-Il with WBA-21 Irradiated and f resh blanket rods; the post Irradiation examination is complete.
The dccumentation is in preparation.
The f,ollowing tests are either part of the current program planning or are ongoing:
o Fear operational transient tests in EBR-il using WBA-20 and WBA-21 pre-t rradi ated rods and f resh rods.
o One operational translent test in EBR-ll to determine the run beyond cladding breach (RBCE) performance.
o One TREAT test with WBA-21 pre-f rradiated rod.
o Four tests in FFTF o
WBA A radial blanket assembly test is in the reactor.
WLA-4* - An inner blanket assembly test is in the reactor.
v o
WBA A thermocouple Instrumented blanket assembly test is in fabrication.
o WBA A power to melt test is in f abrication.
QCS490.9-1 l
Amend. 69 i
(2r WI492..
)
l P ge - 17 [8,22] #62 l
The steady-state and transient progran plans will be available in a summary description document by the end of FY82 including the blanket development program plan. A schedule of the availability of all topical reports on the f uel, control, and blanket assembly designs wilI also be provided at that time.
QCS490.9-2 Amend. 69 May 1982
Peg 2 - 18 [8,22] #62 Ouestion 0C5490.10 What plans are there to incorporate transient f uel mechanical Interaction loads into the COF f uel pin evaluation method for overpower events? Are there any plans f or incorporating FCTT test results f or fuel cladding into the CDF method? Has the method been used to analyze TOP tests (especially those in which cladding breach occurred), and if so, what were the results?
Has the criterion been calibrated to the high fluence data now a.rallable?
If so, please tabulate the additional data that have been incorporated.
Resoonse in the current procedures f or the analysis on overpower transient events, LIFE-IV-T is the principal source of information vis-a-vis cladding stresses due to FCMl; these stresses are subsequently used in the computation of the in-transient damage.
At the time of PSAR submittal, the calibrated transient version of LIFE was not operational. The trcnsient analyses were conducted with a specially developed code named POON which unployed, as Input, FORE-Il temperatures.
POON contains f uel models which result in a conservative assessment of cladding damage, tha' is, load relaxation due to f uel plasticity and f uel creep was not permitted.
Cladding models were consistent with those used in FURFAN and load relaxation due to cladding creep was not permitted.
In addition, the f uel-to-cladding gap was ansumed to be closed or the Initial c! : * r. toef * ; was assumed to be equal to the steady state FCMI If the gap was closed at the start of the transisnt.
Recently, LIFE-IV-T has been calibrated against the following TOP tests:
H5-7A, HUT 3-7A, HUT 3-5A, HOP 3-1B, HUT 12-1 and HUT 12-4.
The calibrated code was then validated against the following tests: HUT 5-7B, HUT 5-2B, HUT 3-78, HUT 3-58, HUT 3-6A, HUT 5-5B and HOP 3-20.
Subsequent to the above calibration of LIFE-IV-T, comparisons of the POON/ FORE-Il procedure with LIFE have shown that, of the two methods, the PCON/ FORE-Il procedure is the more conservative. Table QCS490.10-1 shows a nominal comparison of the average cladding and f uel temperature Increase as calculated by LIFE and FORE-II.
QCS490.10-1 Amend. 69 May 1982
P ge - 19 L8,22J #62 l
Fcr both the U2b and SSE umbrella overpower events at core midplane.
FORE-Il yloids a lower cladding temperature Increase and a higher fuel temperature Increase than LIFE which results in a higher transient FCMI loading than would bo predicted by LIFE. Table QCS490.10-2 shows the peak ' cladding stresses as predicted by both PCON and LIFE for both the U2b and SSE.
Results show that tho PCON/ FORE-ll methodology and the non-strain rate dependent tensile properties used for the PSAR yleids the peak stresses and therefore results in o conservative assessment of the FCMI due to transient overpower events.
Even when the FCMI due to overpower events was considered, the minimum margin still resulted at the top of the f uel column. Therefore, the results reported in the PSAR for damage near the top of the f uel column represent the minimum margins or worst cases results.
With respect to the utIIIzation of FCIT data frm. fusi adjacent cladding, the oppIIcant's current methodology does, indeed, incorporate formulations derived f rom the entire available data base.
QCS490.10-2 Amend. 69
_ --_-___________ ______----_ --- C h t_2 E &l - -._ -------_____
uv ssa aus rcyw su e
TABLE QCS490.10-1 NOMINAL CX)MPARISON OF FORE-Il AND LIFE-lY (TRANSIENT) TEMPERATURG Temperature increase ( F)
SSE U2b
( X/ L = 0. 5 )
(X/L = 0.5)
Average Cladding LIFE-IV 164 29 FORE-1 i 154 30 Avsrage Fuel LIFE-IV 515 182 FORE-1 I 225 181 QCS490.10-3 Amend. 69
_ Ch7E39
rage - 21 LU,ZZJ F3Z TABLE QCS490.10-2 NOMINAL COMPARISON OF PCON AND LIFE-lY (TRANSIENT) STRESSES Equivalent Stress at Cladding ID (PSI)
SSE U2b (X/L = 0.5)
(X/L = 0.5)
LIFE-IV 54,950 36,010 P(X)N with LIFE-IV 56,915 46,801 Temperatures PCUN with FORE-lI 64,168 48,938 Temperatures (PSAR Methodology)
QCS490.10-4 Amend. 69 h _ nGAQ _ _ _ _ _ _
Page - 32 L8,22J #62 Ouestion OCS490.12 What plans are there to evaluate predictions for duct dilation and f uel and blanket subassembly rof ueling loads against FFTF experience.
if it is not planned to use FFTF experience in this area, or CRBR predictions methods do not predict FFTF experience, how does the applIcatant plan to demonstrate the cdequacy of his design with regard to the ef f ect of subassembly bowing and distortion?
Resoonse CRBRP plans to utilize FFTF experience on duct dilation and subassembly rcf ueling loads. CRBRP has identified specific data needs to FFTF and has p;rformed an initial prediction of the FFTF core restraint performance using CRBRP core restraint methodology.
Results from the FFTF predictions will be used to qualify the CRBRP methodology prior to FS/R submittal.
l QCS490.12-1 Amend. 69 May 1962
PIge - 33 [8,223 #62 s
Ouestion GCS490.14 Substantial power jumps are anticipated in some f uel and blanket pins on starting up af ter ref ueling. The Impact of these power jumps is to cause a sharp increase In f uel cladding mechanical Interaction (FCMI) which then decays of f as f uel and cladding def orm under stress. The phenomenon was evaluated for the PSAR using a version of the LIFE code which had not been calibrated as to its prediction of FCMI or the manner in which FCMI decays.
How is it planned to calibrate the method f or predicting FCMI and the ef fect thereof on f uel damage?
Resoonse The LIFE code (which provides FCMI loadings) has been calibrated to EBR-ll f uel rods which have expe'rlenced small power jumps during Irradiation.
In addition, the LIFE code utIIIzes blanket rods in the calibration which underwent power jumps in EBR-Il varying between 5 and 25%. The LIFE code has been checked against f uel and blanket rods in EBR-II which experienced 54 cyclic transients and a final power jump of from 25 to 50%. Future plans are to utilize the data that is to be obtained from the EBR-Il Operational Reliability Testing (ORT) program to quellfy the code. The ORT program will examine the ef fect of slow transients on f uel and blanket rods that occur when FCMI is at or near a maximum. The steady-state and transient program plans will be aval;able f or review via a summary description document by the end of FY82, including a description of the EBR-il ORT program.
In addition, it should be noted that CRBRP Intends to mitigate the ef fects of power jumps at ref ueling on FCMI vie a prograr.med startup (see PSAR Figure 4.2-16).
QCS490.14-1 Amend. 69 May 1982
PIge - 34 [8,22] #62 Ouestion GCS490.15 Experience with the FFTF f uel system will be essential to resolving several licensing issues f or the CRBR.
Please detall the plans f or surveillance of FFTF driver f uel pins (including both non-destructive and destructive exaninations) and for transient tests of FFTF Irradiated f uel pins. To what extent will the project have inf f ence over the base technology program?
Resoonse The FFTF has a Driver Fuel Evaluation plan that includes Post Irradiation Exar.Instion (PIE) surveillance of f uel assemblies and of f uel pins contained within the assemblies. Present program plans (which are periodically updated) include approximately 20 FFTF driver assembiles. Presently, two assemblies are to be removed at each of the first four cycles, in addition, two Instrumented assemblies, one power-to-melt assembly (PIE) ongoing), six Run to Cladding Breach (RTC8) assemblies, and several special test assembiles are to be exam ined. The PIE Includes both non-destructive and destructive exanination of fuel rods and also duct exanination. Transient testing plans of FFTF driver f uel rods have not as yet been f inalized.
CRBRP Input to the transient testing plan has been presented, and a preliminary TREAT testing program is being finalized. The steady state and transient program plans will be provided in a summary description document to be available by the end of FY82.
QCS490.15-1 Amend. 69 May 1982 l
PJge - 35 [8,22] #62 Ouestion OCS490.16 Several transient tests have been conducted since the most recent CRBR licinsing activity. Documentation (data report, final report) available now or which becomes available in the f uture is requested for the following tests:
lied.L 13.
HOP PTO 1-2A J1 HOP PTO 3-2E P2 HUC PTO 2-2A P3 HOP 1-6A P3A HUT 3-5B P4 HUT 5-5B H6 HUT 3-6A HUT 3-6B W-2 W-1 All HEDL tests above bear the new test designation.
Resoonse A schedule for the availability of all such topical reports supporting the CRBRP f uel, control, and blanket assembly design wilI be provided by the end of FY82.
QCS490.16-1 Amend. 69
_ _ _ _ _ May _1982
Page - 36 Le,22.1 #62 l
_Duestion OCS490.17 Tho behavior or irradiated core I steel for the FFTF duct and f uel pin cicdding was found to be significantly dif ferent in swelling and steady-state stress rupture strength from N-lot and other test steels upon which much of How does the applicant plan to demonstrate the CRBR design was based.
c:nservatism of CRBR design in view of the unexpected additional uncertainty in material behavior that this Implies? Does the applicant plan to initiate Irradiation tests early-on to discern whether a similar deviation will occur for CRBR duct and f uel pin cladding material?
If no such testing is planned, how does the applicant plan to discern such a deviation before it has become a problem in CRBR7 Resoonse OERP Intends to procure 316 stainless steel to FFTF core 4 type specifications.
FFTF core 4 material will be Irradiated in FFTF to discern any major deviations f rom the data base, it is expected, however, that only heat-to-heat variations will exist between FFTF core 4 steel and CRBRP production material. The ef f ect of heat-to-heat variations have been f ound to bo small as evidenced by FFTF core 1 data.
QCS490.17-1 Amend. 69 May 1982
(
P ge - 37 [8,22] #62 Oucst f on OCS490.18 The discussion of operation with defected fuel pins in the PSAR was apparently current as of September,1979. Please update this discussion as may be warrented by new Inf ormation.
What additional def ected pin test results have become available since 1979, and how do they apply to the CRBR7 Have any translent tests been performed (or are any planned) on defected pins?
If no transient tests on defected pins are planned, how does the project plan to demonstrate that continued operation with f ailed pins would be safe, since this entails exposure to anticipated and unlikely events?
How is the change to 33% Pu concentration expected to af fact continued operation of f ailed f uel pins?
Resoonse Q
A comprehensive test program to address the Run Beyond Cladding Breach (RBG) capability has been undertaken and is ongoing. Two pre-def ected irradiated f uel rod assembly steady-state tests have been completed in EBR-Il and reported (RB G-6, 7, Reference 1 and 2).
Three additional irradiated fuel rod assembly steady-state EBR-ll tests have also been completed in which breaches occurred naturally and operation continued af ter breach (RBC8-1, 2, 3).
In addition, another pre-f rradiated f uel rod bundle steady-state EBR-il test In the Breached Fuel Test Facility (BFTF) has also been completed (XY-2) in which a breach also ocurred naturally. United Kingdom data have been reported (Reference 3) on the of fects on Dounreay Fast Reactor (DFR) frradition of large number of pre-defected fuel rods. They conclude that even in high burnup rods, continued operation af ter f ailure can be allowed for at least 60 days without unacceptable deterioration of the rod or significant loss of fuel. The U.S.-obtained steady-state information also Indicates benign operation.
Preliminary conclusions are that there was little f uel loss; enhanced detectable delayed neutron detector signals were observed and no rapid rod-to-rod propagation occurred.
Steady-state operation can saf ely continue for at least (approximately) 22 days. The defects appear to grow mostly due to reactor shutdown-startup operation.
QCS490.18-1 Amend. 69 Amra______ _ ____ __ _ ____ ___ _j
PIge - 38 [8,22] #62 As part of the EBR-il Operation Reliability Testing (ORT) program, additional R8W testing is either planned or underway. Fuel steady-staie kinetics and contamination f uel rod bundle tests are planned (RBW-KI, K2, K2A, K28, K2C).
The objectives of thee tests are to utilize the EBR-ll BFTF to confirm the kinetics of fuel-sodium reaction in breached mixed-oxide rods, to determine the delayed-neutron (DN) signals from them as a function of time and breach condition, and to characterize the type of contamination to be expected from RB W operation. Three detection and Instrumentation f uel rod bundle tests are planned (RBG-DI, D2, D3). The objectives of these tests are to determine delayed neutron release characteristics from breached rods as a f unction of reactor power and sodium temperature and will utilize both open oore EBR-il positions and a specially Instrumented Fuels Performance Test Facility (FPTF)
In the EBR-il.
Five additional Fuel and Irradiation Variables tests are planned (RBW-V8, V4, V5, V6, V7) with the objectives of thee tests to invettgate a range of fuel and irradiation variables expected to influence the swelling and degradation of fuel and blanket rods in open core EBR-il positions. The V2 test will explore plenum defects, the V4 test involves larger diameter f uel rods, the V5 test will use blanket rods, the V6 test will be an unreconstituted test to explore mid-life f ailures, and the V7 test will provide Information on storage of facts in sodium.
There have been no recent transient tests run on defecied f uel rods.
In the ORT program, it is presently planned to have a transient overpower test (TOP 1-2) on a breached rod. The steady-state and transient progran plans wil I be available in a summary description document by the end of FY82, including the ORT /RBT progran plan.
The RBW-V6 test mentioned earlier is presently planned to study the 33%
plutonium rods, it is expecteG that there may be a slight enhancement of fuel-sodium reaction due to increased oxygen availability.
QCS490.18-2 Amend. 69 Y }M- - -
Pcg3 - 39 L8,22) #62 R:forneces:
1)
HEDL-TE-79-23, "Results f rom the RBG Irradiation of a Pre-Def ected Pin (RBW-6)", D. C. Langstaff, et.al., March, 19/9.
2)
HEDL-TE-79-62, "Results f rom the RBT Irradiation of a Pre-Def ected Pi n (RBG-7)", D. C. Langstaff, et.at.,
February, 1980.
3)
"Def ect Pin Behavior in the DFR", W. M. Sloss, K. Q. Bagley, E. Edmonds, and P. E. Potter.
International Conf erence on Fast Breeder Reactor fuelg Perf ormance, Monterey, Cal i f orni a, March 5-8,1979 (Library of Congress Catalog Card No, t
79-50775).
\\
\\
QCS490.18-3 Amend. 69
_ ~ nm _
Question DCS490.19 There are e number of thermal and mechanical fuel performade codes, both steady-state and transient, that more or less satisf actorrTy'prodlet data available on thermal performance, cladding breach, and cladding Inelastic ctrain.
Yet these codes can very wildly as to f uel-cladding Interf ace pressure, gap conductance, fission gas release, etc., particularly In cxtrapolations outsida the calibration data base.
This situation arises b:cause the only data available for calibration are Integral-pin, post-test data including thermal data relatively remote f rom the region of Interet.
it is virtually impossible to qualify Individual phenomena models; this coupled eith the number and complexity of models and the uncertainty of material properties allows an unlimited number of solutions to the problems of predicting extrapolated perf ormance.
It is of particular concern therefore, that all temperatures and performance predictions be perf ormed in a consistent f ashion for the purpose of reviewing fusi performance.
Correlations and models that use calculated input parameters (for instance, the Failure Potential Correlation, or any of the SIEX code correlations) should be used only in keeping with the manner in which they wre developed.
Otherwise their uso may yield totally involld results.
Alternatively, when assessing compilance with a criterion based on Independent data, use of dif ferent models may give very dif ferent assessments.
What steps has the applicant taken, or does he plan to take, to ensure that all evaluations will be done in a consistent f ashion, and that inputs to all empirical correlations will be determined in a f ashion consistent with their dsvol ep.r :'- ?
Reseense CRBRP agrees with ite stated concern and takes considerable care to insure consistency in analytical methods through implementation of the following project proceaures:
QCS490.19-1 Amend. 69 CD7 M l
Page - 41 [8,22] #62 o Analysis checking procedures.
o Analysis code verification and quellfication procedures o Requirements for approval of material properties to be used by the project.
o Configuration control of all baselined documentation (drawings, discussion, hot channel f actors, etc.)
The checking, verification, and code qualification procedures will preclude the possibrilty that Individual models developed and calibrated using one code are taken out of context and used in another code.
The CRBRP final analysis verification and qualification are subject to Project audit.
QCS490.19-2 Amend. 69 May 1982
]
Psga 1 (82-0321) [8,22] #63 Ouestion CS490.20 Cladding breach for undercooling condition: !s generally considered to occur when the current burst pressure declines below the plenum pressure due to increasing cladding temperature. There are few data, If any, available to conservatively confirm just when breach would be expected.
Virtually all FCTT data were obtained for either very high gas pressure or for low fluence or nonf ueled cladding, and the loss of coolant tests conducted by ANL were all for low burnup pins.
Please supply any additional data that may now be available (either FCTT or Integral pin data) that are more relevant than the data quoted above to end-of-life, undercooling f ailure threshold conditions -
that is, at plenum pressures In the 1000 to 1500 ps! range, and at cladding Irradiation damage levels approaching end-of-life conditions.
If no more relevant data than that quoted above are available, are there plans to obtain such data?
If so, please describe those plans.
If not, how does the applicant plan to demonstrate the conservatism of the f uel and blanket design f or undercooling conditions?
Easponse Additional FC1T tests were perf ormed on def ueled cladding with fluences of 6.3 to 11.2 - 1022 n/cm2 (E > 0.1 MeV) in the 1000 to 1500 pst internal gas pressure range. These tests are included in the test series referenced in Question CS490.8, as Type A tests, see Figure QCS490.8-1. A decision on the release of the above Information including report format and a schedule f or the release of the data will be submitted by July 31, 1982. A schedule of availability of all such topical reports supporting the fuel, control, and blanket assembly design will be provided by the end of FY '82.
QCS490.20-1 Amend. 69 May 1982
P;ge 2 (82-0321) [8,223 #63 l
OuestIon CS490.21 in PSAR Section 4.2, the reader is referred frequently to PSAR Section 15.1.2 for details of the CDF f uel evaluation model and its development and qualif! cation. However, the portions of Section 15.1.2 dealing with those subjects appear to have been deleted. Does this presage a decision to abandon the CDF f uel evaluation method for preparation of the PSAR and the operating licenses appIIcation?
If so, what does the applicant plan to use In its place to evaluate fuel perfccmance?
Resoonse With respect to the description of the CDF technique, in PSAR Section 15.1.2 the reader is referred to Reference 58 of Section 4.2 as per PSAR Amendment 61.
The CDF remains an integral part of the applicant's current performance analysis methodology and wilI continue as such.
QCS490.21-1 Amend. 69 May 1982
P$ge'3 (82-0321) [8,22] #63 Ouestion CS490.22 The CDF method f or f uel perf ormance evaluation provides a model for determining the accumulated cladding damage due to steady state operation and all anticipated transient events plus one unlikely event at the end of lif e, and includes auxillary models to account f or cladding wastage, corrosion, and Irradiation damage.
All of these models, however, appear to depend on input f rom other sources as to plenum pressure, fuel cladding mechanical interacti on loeds, time-temperature history, etc.
The manner in which this input is generated is also important to the validity of the method.
Was the generation of input data f or determination of the transient limit curves (TLC's) accomplished in a manner consistent with the generation of data f or Individual l
events that were compared with the TLC's.
Resoonse The input Itans and methods used to generate the Transient Limit Curves are required to be consistent with the items and methods used to describe the Individual events.
QCS490.22-1 Amend. 69 May 1982
page 4 (82-0321) [8,22] #63 Ouestion CS490.23 The criterion f or preservation of coolable gemetry for all reactor design basis accidents in extremely unlikely events is no sodium boiling. The coolant safuratIon tamperature at tha top of the CRBR core is about 18000F f
(1255K). Presumably, therefore, no phenomenon has been identified up to 18000F that could af fect coolable gemetry.
However, there may be a mechanism to comprmIse coolabie gemetry short of coolant boliIng under ioss-of-fIow conditions.
In the space between 1600 and 18000F, significant numbers of end-of-lif e f uel pins could breach, releasing large amounts of fission gas.
Studies have shown that at f ull flow, release of all of the gas in the f uel pins in one subassembly at the end-of-life could uncover the portion of the subassembly, or the top part of it, for approximately 0.1 to 0.2 seconds.
Presumably, the cladding that was uncovcred would be without cooling during this time. The encovered t,ime would be much less than the approximate 0.8 seconds without cooling required at f ull power to reach the cladding solidus temperature. However, under loss-of-flow conditions (Iow flow), the time some portion of the core would be uncovered would undoubtedly be much longer, probably more than long enough to melt cladding at f ull power.
At the other extreme, if the power were instantaneously reduced to zero f ra f ull power, the f uel and cladding with no cooling would equilibrate to a temperature above the cladding solidus f or all powers above about 20 KW/m.
It therefore, follows that indefinite loss of cooling is not necessarily tolerable even with an Instantaneous scram.
In short, the existence of some combination of flow coastdown, residual heat generation rate, and residual stored energy that would culminate in cladding melting cannot be ruled out now for a loss-of-coolant event that penetrates the temperature space between 0
1600 F and coolant saturation.
Therefore, a no-bolling limit does not necessarily preclude loss of coolable geometry under loss-of-flow conditions.
Rather, protection against loss of coolable geometry is ensured by scrmming the reactor rapidly enough to avoid breach of any fuel pins. With these considerations, explain the adequacy of the no-bolling limit for undercooling events. The f act that there are no identified protected loss-of-flow events in which cladding (let alone coolant) exceeds 16000F f or end-of-life (high plenum pressure) fuel pir.s does not answer this question.
Response
Protection against loss of coolable gemetry is ensured by scrmming the reactor rapidly enough to prevent melting of cladding. The no-bolling events.(1)is adequate to ensure no cladding melting for all design basis guideline The question specifically addresses a postulated, extremely unlikely undercooling event in which cladding breach occurs leading to fission gas rolease and a consequent degradetion of cladding cooling. The concern expressed is that this degradation of cooling might lead to cladding molting and thus Ioss of core coolable gecznetry. An enveloping event involving cladding breach has been evaluated and maintenance of core coolable geometry QCS490.23-1 Amend. 69 May 1982
Psge 5 (82-0321) [8,22] #63 demonstrated.
A summary of this evaluation is provided in the f ollcwing paragraphs.
A more detailed discussion of the evaluation is being prepared f or inclusion in the PS AR (Section 15.1.4) in the near term.
Accident Assurrotions Although the question only addressed undercooling events, it was assumed that a large increase in reactor power (due to a 60c step reactivity insertion) occurs concurrent with a flow coastdown in all 3 PHTS loops, it was postulated that the primary reactor shutdown system f alls completely so that reactor shutdown depends on the secondary shutdown system.
It was f urther postuisied that the secondary control assembly (SCA) with the highest worth f ailed to insert and that the SCAs which did Insert did so with a speed reduced to account f or conservatively determined ef fects of a Saf e Shutdown Earthquake.
Analvsts An analysis was perf ormed assuming f ailure of all fuel rods in the hot subassembiles at various points in the reactor cycle, with ruptures assumed at the top-of-core location.
it was conservatively assumed that while f Ission gas blankets a pin, there is no cooling of the af fected cladding.(2) The flow transient due to the fission gas release iests only approx'mately 0.2 seconds bef ore the coolant flow velocity is restored.
There is a sufficient margin to coolant saturation temperature that the coolant re-enters all coolant channels.
Conclusions The ef f ect of temporary less-of-cooling due to f ission gas release would not result In ciadding met ting.
Summarv Assurance of core coolable geonetry requires reactor shutdown suf ficiently rapidly to avoid cladding melting.
The no-sodium bolling criterion assures that the cladding does not reach its melting temperature.
A postulated event wbfe" a""m +ket fission gas release degrades cladding cooling by t;,... * :,..~. ;r.g the sodium coolant does not resul t in cl adding melti ng and thus does not result in loss of core coolable geonetry.
(Il l n f act, prevention of sodium boiling may be an overly conservative guideline because there is considerable experimental data which demonstrates that at low power levels, no cladding melting occurs even in the event of sodium bolling (Ref erences QCS490.23-1 and 2).
(2) Experimental and analytical data show that introduction of vapor into a channel does not renove all cooling f ran the cladding (References QCS450.23-3 and 4). The mitigating ef fects which occur in two-phase flow were not considered in this analysis.
QCS490.23-2 hnend. 69 f"w M9
P$g~T (82-0321) [8,22] #63 e
Ref erences:
(QCS490.23) 1.
J. L. Wantl and, P. W. Garri son, W. R. Nel son, et al., "Sodi um Bol l ing incoherence in a 19-P!n Wire-Wrapped Bundle," in Proceedings of the International Meeting on Fast Reactor Safety Technology, Vol. 4, pp. 1678-1685, American Nuclear Society, LaGrange Park, ll (1979).
2.
J. F.
Dearing and S. D. Rose,
"Two-Dimensional Modeling of Sodium Boll Ing in the W-1 Sodium Loop Saf ety Facil Ity Experiment," Trans. Amer. Nucl.
Soc.
39, pp. 1067-1069 (1981).
3.
J. M. Henderson, et al., "W-1 SLSF Experiment Final Report," TC-2050-R1, September 1981.
4.
G. Hoeppner, F. E. Dunn, and T. J. Heames, "The SAS3A Sedium Boll ing Model and its Experimental Basis," Trans. Amer. Nucl. Soc.
20, pp. 519-521 (1975).
QCS490.23-3 Arnend. 69 May 1982
l Page 7 (02-0321) [8,22] #63 Ouestion CS490.25 Plenum pressures f or the FFTF were determined f or design purposes assuming 100% release of fission gas.
However, the CRBR takes credit f or retained fission gas, predicting the f raction of release using a correlation to data (PSAR page 4.4-40).
In developing the correlation, were peak or average values of linear heat rating and burnup used to represent the overall pin?
If peak values were used, will not the correlation underpredict the f ractional release? This would not necessarily be detected by comparing predicted with observed values in Table 4.4-13 unless the pins in Table 4.4-13 had axial power distributions that were signf icantly dif ferent f rom those of the cal ibration pins l isted in Tabl e 4.4-12.
Were the predictions shown in Table 4.4-13 made f or nominal parameters or 2 signa parameters?'
Resoonse Fission gas release analytical models and experimental data were not available in the late 60's at the time of FFTF design, thus the only alternative was to conservatively adopt 100% release.
In CRBRP design, empirical models derived f rom experimental data are used to calculate the f ractional release f rom unrestructured f uel. One hundred percent release is still assumed f rom restructured f uel.
In evaluating the release f rom the unrestructured f uel, the entire fuel column, exially and radially, is considered; the f uel temperature distribution is calculated, restructuring isotherms are determined and the local fractional release is calcu!cted f rem empirical correlations (which are f unctions of linear power rating ana burnup). The local release is then Integrated over the entire f uel to obtain the total release.
Thus, the actual distribution Qant a peak or an average value) of the linear. power and burnup is used.
Similarly, the experimental pins used f or calibration and verification were characterized in detail, using their actual power /burnup distribution.
Release calculations were done using exactly the same procedure as f or CRBRP design pins and compared with the measured release, as reported in PSAR Sectioin 4.4.2.8.16.
The NICER code, which adopts simplified models f rom the LIFE code, was used to perf orm these calcul ations.
Predictions of data were theref ore made using nominal paraneters.
Uncertainties in fission gas release predictions were obtained through comparative regression analysis of predictions versus ru coro-ente Tna nacertainty on f ractional fission gas release is only one of the uncertainties considered in evaluating the plenum pressure.
Uncertainties on plenum temperature, plenum volume, fission gas production (lineer power and burnup) and fission gas yleid arc also considered, in addition to the ef f ect of residual gases, as discussed in Section 4.4.3.2.4 The 2o l evel of conf idence is used in design predictions of the f ission gas plenum pressure.
1 QCS490.25-1 Amend. 69 May 1982
.Page 8 (t.2-0321) [8,22] #63 Final ly, it must be pointed out that f or the highest power / temper ~ature pins, nominal +2o f Ission gas release is equivalent to 100% release (see Figure 4.4-30).
Only for the " cold" pins is the calculated 2o release less than 100%. Use of a 100% release on these pins would have resulted in an overcooling requirement (see orificing philosophy discussion in Section 4.4.2.5).
The advantage of the CRBRP method is that by no "overdesigning" the cold pins, flow can be saved and this " saved" flow is allocated where most needed, i. e., to th e hot pi ns.
The hot pins thus receive more flow than they would if a 100% release assumption were used throughout the core and, therefore, the CRBRP approach is not only more realistic than the FFTF, but also more conservative.
t QCS490.25-2 Amend. 69
%L 1_982__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Page 11 (82-0321) [8,22] #63
_0uestion CS490.26 What is the predicted pienum pressure f or the absorber rods? How do predicted pressures compare with test data on absorber rods?
Resoonse New B C test data (see Question CS490.28) is currently under evaluation.
4 Preliminary correlations f rom these data predict PCA pin pressures of 3600 ps!
at 275 FPD f or the peak power pin.
Verification of B C helium release 4
correlations will be obtained from PCA f rradiation Test in FFTF.
QCS490.26-1 Amand. 69
_ _ h _ 51CO2__ __
Pagelb (82-0321) [8,22] #63 Ouestion CS490.27 lt appears that the CDF method f or fuel pin performance evaluation is primarily oriented toward prediction of design Iif e, is the method used f or evaluating the extent of damage shert of design iIf e, or is It used strietiy as a f all, no-f all Indi cator?
In applying the CDF method, is the 1.0 for design lif e partitioned into separate allocations f or steady-state and/or anticipated transients?
Resoonse The CDF method, as employed in PSAR Section 4.2, provides both an Indication of cladding damage prior to achieving design goals as well as an Indicator of adequate perf ormance vis-a-vis the preclusion of cladding f ailure.
The CDF is never partitioned into separate allocations f or steady-stals and/or anticipated transients.
I QCS490.27-1 Amend. 69
_____ ____ ____. _ _ _ h MG@
Page 11 (82-0321) [8,22] #63
_OuestIon CS400.28 Please pro tide the available BaC test data and documentation f or the tests identif ier. In Tabl e 4.2-46A, PSAR page 4.2-413, as supporting the CRBR control assembly resign.
What revelant experience has been gained thus f ar in FFTF startup testing and operation with regard to CRBR control assembly design?
Resoonse The 84C tests, ilsted in Tcble 4.2-46A have been completed and the data is currently under revlew.
Final correlations on B4C pellet swellIng and hellum release wili be incorporated in the Nuclear System Materials Handbook.
During FFTF startup and operation, the f ollowing results wer e obtained:
1.
Measured scram insertion times are within requirements, and are in good agreement with predictions.
2.
Measured control assembly worths were welI predicted and meet or exceed requi ranents.
Because of the similarity between FFTF and CRBR designs and analysis methods, these results provide conf idence that similar results will be obtained f or CRB R.
QCS490.28-1 Amend. 69
__h nas
PTge 14 (82-0321) [8,22] #63
_Ouestion CS490 21 Although not necessary for review of the PSAR, all codes used in design and l
evaluation of the f uel and blanket rods will need to be reviewed. Based on our current understanding, the codes to be reviewed will include.
FURFAN LIFE-Ill any of the LIFE-IV series anticipated to be used for the PSAR FORE-2M FRST Resoonse The applicant acknowledges that codes used in design and evaluation of the f uel and blenket rods may need to be reviewed by NRC during the PSAR review.'
Appendix A of the PSAR provides existent information for the various oodes cited. Relevant inf ormation f or the codes will be made available at the time of PSAR submittal to f acilitate NRC review.
QCS490.29-1 Amend. 69 May 1982
Page 15 (82-0321) [8,22] #63 l
l Ouestion CS490.30 The coolan' flow rate is automatically cut back upon the initiation of a reactor scram to minimize thermal shock. Are there any conceivable circumstances under which a scram could be called for, the rods f all to be Inserted, and the flow cutback is still executed? What is the outcome of such an event?
Resoonse in view of the complete redundancy and diversity of the design of the CRBRP shutdown systems, no such circumstances are conceivable.
- However, hypothesized events similar to the one discussed in the question are analyzed by the CRBRP in the beyond-design-base evaluations and are documented in Ref erences 10a and 10b of Section 1.6 of this PSAR.
l t
QCS490.30-1 Amend. 69 L
May 1982
Prge 16 (82-0321) [8,22] #63 Question CS490.31 l
On page 11 of Reference 58 to PSAR Section 4.2 It states that "...alI possible emergency events are divided into two broad categories according to the f
physical processes involved, viz., undercooling and rapid reactivity insertions. " The definitions of the two categories appear to exclude any consideration of transient fuel cladding mechanical Interaction on a slow time scale, that is, on a time scale much greater than one second. Yet, the possibility of such occurring clearly cannot be ruled out.
In fact, ell reactivity insertion (or overpower) events are f undamentally dif ferent from loss-of-flow events regardless of speed. How, the, does the CDF model evaluate " slow" reactivity insertion events?
Resoonse The quertion refe.'s to the observation that unterminated transient tests on f uel rods in TREAT resulted in a decrease of the power at the f ailure threshold with 6ecreasing rap rates. CRBRP is protected against reactivity insertions at very slow rap rates. The automatic control features preclude slow reactivity ramps transients exceeding 103% of full power. This band is a combination of a i 2% dead band on power oscillation plus a i 1% bar.d due to calorimetric power measurement uncertainty.
In the manual control mode, the slowest reactivity ra p fr. 0.05%/sec. The primary trip is activated af ter 300 seconds into the transients when 15% overpower is being exceeded.
In this time-power envelope of 115% power and 300 sec. time alI combinations of power rep rates are possible.
The CDF model documented in the PSAR Section 4.2 does not employ a strain rate sensitive transient (DF model. The worst possible slow reactivity insertion event, the U2b transient, is enveloped by a rapid insertion transient followed by a 300 sec. hold of power.
Such a transient accumulates larger transient damage than the postulated slow reactivity insertion transient. This conclusion is based on a preliminary comparison of the #enstents calculation and assumptions presented in the PSAR with a transient damage model which
!ncludes a strain rate sensitive cladding damage model. Preliminary results Indicate that a f ast power ramp results In higher fuel-cladding mechanical interaction stresses than slow reactivity stresses.
Dur!rg the 300 sacond hold of power, the maximum stress relaxes and the strain rate decreases, but not suf ficiently to result in lower damage than accumulated during a slow power increase as a result of slower cladding strain rate. An updated transient CDF model which incorporates cladding strain rate of fects is currently under development and wilI be discussed as part of the PSAR.
Unterminated transient tests were performed with the objective of determining f ailure threshold as a function of power rap rate. Preliminary results have been published in Ref erence QCS490.31-1. Additional slow overpower transient tests are planned in the Operational Transient Test (ORT) progra In EBR-ll.
A preliminary description of the progra objectives is presented in Reference QCS490.31-2. The ORT progre will include a simulation of the CRBRP U2b event with f ast power rap rates and a subsequent power hold of 300 sec., in addition to the slow rap rate tests described in Reference QCS490.31-2.
QCS490.31-1 Amend. 69 May 1982
]
Page 17 (82-0321) [8,22] #63 The steady-state and transient program plans will be available in a summary description document before the end of FY '82 including the EBR-il ORT progran.
References:
QCS490.31-1, " Oxide Fuel Perf ormance During Normal Operation and Of f-Naninal Events", E. T. Weber, et al, Reactor Saf ety Aspects of Fuel Behavior, ANS Topical Meeting, August 2-6,1982, Sun Valley, Idaho.
QCS490.31-2, "A Progran for Operational Transient Testing of Breeder Reactor Fuel", A. Boltax and J.1. Sackett, Reactor S:f ety Aspects of Fuel Behavior, ANS Topical Meeting, August 2-6, 1982, Sun Val ley, Idaho.
QCS490.31-2 Amend. 69 May 1982
Page 18 (82-0321) [8,22] #63 Ouestion CS490.32 If for whatever reason, one or more absorber rods breached, and B4C were washed or eroded out of the breached rods, how would this be detected? What is the maximum reduction in shutdown capability In either the primary or secondary control system that could occur through either burnout or washout at the detection threshold?
Is any survelllance planned to ensure that the f unctional capability of neither the primary nor secondary control systems has degraded unacceptably?
Resconse There is no system in CRBR designed to detect the rupture of a control assembly pin or the subsequent erosion or washout of the B4 This position C
is supported by the foilowing Information:
1.
As detailed in Section 15.4.2, the f ailure of an absorber pin is considered unlikely due to the conservative loads and design criteria used in the design analysis. The only identified mechanism which could cause pin cladding f ailure is excessive Internal pressure, and pin rupture tests have shown that this mechanism produces only pin hole or very fine crack defects.
Since pin pressure increases with lifetime, large pin pressures would only occur late in life when the required shutdown reactivity is a m ini mum.
Recent test data from FFTF on irradiated B C pellets have shown that for a 2.
very large cladding defect (1.0" x 0.1"), 4a maximum of only 0.16 grams of B4C were eroded af ter 50 days In 1000 F sodium, flowing at 5 feet per 0
second. The results of this test, and other tests described in Section 15.4.2 lead to the conclusion that the probability of eroding significant quantitles of B C from cladding defects is extremely unlikely.
4 3.
Due to the high self-shielding of the enriched B C in control assemblies, 4
the loss of small amounts of B4C would produce no detectable change in control assembly worths. To reduce the worth of a control assembly 15 would require the loss of approximately 150 grans of enrtched B C 4
4.
The scran Insertion performance and shutdown margin is calculated assuming t% rice + reective control assembly is inoperative. Therefore, the B C A
from an entire control assembly could be lost without compromising tfie calculated scran Insertion performance or the shutdown margin.
Finally, the position of the operating control rod banks is constantly monitored and compared to the expected position.
if significant quantitles of B C were lost due to cladding rupture of the PCA pins the operating bank wbuld be withdrawn less than the predicted amount. Tbis trend would be 4
monitored, and the core would be shutdown bef ore the shutdown margin was reduced below an acceptable level.
QCS490.32-1 Amend. 69
_ _ - - _ _ _ _ _ -___- ___-- _ __- _ _ - __ _ _ __ _ _ _ __ _ _ _ _ Rm _ %GB _
P:ge 19 (82-0321) [8,22] #63 Ouestion CS490.33 l
Please show the quantitative delays in of fecting a reactor scran starting with the timo a real variable or quantity reaches its scran trip point and ending with the time the power just starts to decline.
Resoonse The time delay in ef fecting a reactor scran varies with event.
General ly speaking, this time delay is the sum of 1) Instrument channel, 2) Reactor Shutdown System (RSS), and 3) control rod delays.
The Instrument channel delay includes the sensor and transmitter delays. The delays range f rom 10 msec. for the neutron flux detectcrs to 5 seconds for the IHX and evaporator outlet thermocouples. Table 7.2-3 summarizes the Instrument channel delays used in the Chapter 15 safety analysis.
The RSS delay includes delays from the calculational units, compara1 ors, coincidence and final actuation logic.
A delay of 0.1 second conservatively envelopes these Individual f actors.
The control rod delay includes delays from unlatching and rod insertion spe eds.
Unletching time (start of CRDM stator current decay / current interruption to solenoid valves to start of primary / secondary rod motion) is approximately 0.1 second. The time to " turn around" an event (begin reduction of adverse temperature and/or power trends) varies, but can generally be esmed io occur before one dollar of reactivity has been Inserted. This occui s =; thin opproximately 0.31 and 0.46 seconds af ter the start of rod motion for the primary and secondary control rods, respectively. Figures 4.2-93 and 4.2-94 provide the minimum primary and secondary scran Insertion requirements used in the Chapter 15 safety analysis.
QCS490.33-1 Amend. 69 May 1982
Pcge 18 (82-0321) [8,22] #63 l
Ouestion CS490.34 It is our understanding that rod bundle-duct Interaction can cause substantial cladding stresses, at least f or blanket rod bundles.
Are these loads considered in evaluating f uel rod perf ormance by either the CDF or the ductility limited strain model?
If so, please provide a specific description of how th i s i s done, including examples f rr both steady-state and transient ocnditions.
Resoonse Bundle duct Interaction has been postulated to cause substantial local cladding bending stresses.
This conclusion is based on rod-bundle duct Interaction analysis. The analysis method and some conclusions have been reported in Ref erence QCS490.34-1. An earl ier report, Ref erence QCS490.34-2, contains details of the analysis methoa and was submitted to NRC. Cladding principal stresses due to rod bundle duct interaction have been calculated based on conservative assumptions, for example; assuming rigid pellets without ga p.
These calculated wire wrap stresses or strains are presently not being considered in the cladding danage analyses.
Actually, the f uel pellet material does creep significantly at typical CRBRP blanket rod operating conditions and allows local creep def ormations of the cladding due to wire-cl adding interaction. Claddin prof ilonetry measurements were perf ormed on experimental bl anket assembl ies o j A-20 and WB A-21 tested in EBR-il to determine the magnitude of the local cladding def ormation.
Measurements on WBA rods indicate a maximum ovality of 0.6% D/D superimposed on a peak average disnetral strain of 1.15% D/D. This Inf ormation is contained in the f inal perf ormance test report f or these tests. The response to Question CS490.9 gives the status of these tests and a date f or a plan to release this inf ormati on.
Since cladding damage due to rod bundle-duct interaction was of concern, fuel assembly and radial blanket assembly tests are planned which will exhibit significant rod bundle duct interaction.
These tests are included in the test plans which will be provided in response to Question CS490.15 for the FFTF fuel surveillance progran, in response to Question CS490.9 for blanket Irradiation program, and as part of the Irradiation test program in support of CRBRP.
However, the Initial concern about The damaging ef fect of wire wrap-cladding interaction on cladding perf ormance decreased because the test assembly P53 In EBR-il with significant rod bundle-duct interaction did not exhibit a rod f ail ure, in addition, Irradiation tests on f uel assemblies with 217 rods and very severe rod bundle-duct Interaction did not result in cladding f ailures (see Ref erence QCS490.34-3).
In summary, no cladding f ailures have been attributed to wire wrap-cladding Interaction. One of the reasons may be that cladding f ailures are caused by exceeding membrane stress and strain limits. Cladding loads due to wire wrap-cladding interaction are compressive and reduce the local membrane stress and strain but introduce secondary bending strains which relax due to creep as Indicated by the blanket Irradiation tests in EBFhil.
QCS490.34-1 Amend. 69 May_1982
Pcge 19 (82-0321) [8,22] #63 Because of the above considerations and since testing to evaluate these ef fects are planned, the calculated wire wrap-cladding stresses or strains are presently not being considered in the cleading damage analysis.
Ref erences: QCS490.34-1, E. C. Schwegl er, Jr., "Cl adding Response to Uni f orm Radial Compaction of Clinch River Breeder Reactor Plant Rod Bundles", Nuclear Engineering and Design 61 (1980) pgs. 223-235.
QCS490.34-2, CRBRP-ARD-0149, E. C. Schwegl er, " Wire Wrap-Cl eddi ng interaction in LMRBR Fuel Rods", dated December 1977.
QCS4 90.34-3, J. Rousseau, et al, "Def ormation of Fuel Reds with Wire Spacer in the Presence of Swelling and Creep Due to Irradiation", International Conf erence on irradiation Behavior of 8
Metal lIc Materials f or Fast Reactor Core Components, Corsica, France, June 4-8, 1979, pg. 291.
QCS490.34-2 Amend. 69 May 1982
)
Pzg2 3 82-0222 & 82-0307 [8,22] 27
_0uestion CS490.36 Part As in the uncertainity analyses presenteG In section 4.4.3.2 of the CRBR PSAR, the rationale used to determine 2 and 3 uncertainty factors for thermal and hydraulle data is discussed. The discussion does not include a quantitative justification f or non-statistical factors nor does it provide Information about the methods used to determine statistical factors.
Please Indicate for the data presented in Tables 4.4-18A through 4.4-31 which of the uncertainty f actors are determined statistically and which are not. Also, for the non-statistical f actors please provide a quantitative basis and for the stati stical f actors please provide a detailed description of the methods used and of the data base.
Part B:
In addition to uncertainities in material property data, design tolerances, and similar data there are uncertainties associated with the numerical methods (including model) used in the various computer codes. Are uncertainties in numerical methods (including models) included in the uncer.alnty factors presented in Tables 4.4-18A through 4.4-31?
If uncertainties In numerical methods are included in the overall uncertaintles, please provide a detailed mathematical description of the methods used to determine these uncertainties.
If numerical method uncertainties are not accounted for, please explain why they are not.
Resoonse A topical report "CRBRP Core Assemblies Hot Channel Factors Preliminary Analysis", CRBRP-ARD-0050 has been submitted to the NRC and was issued by TIC / DOE in February 1980. This report discusses the methodology, rationale and bases of the hot channel / spot f actors used in the CRBRP core assemblies thermof luids design.
The specific questions are f'ully covered in the report.
QCS49C.36-1 Amend. 69
- Y 0
PegI2b (82-0321) [8,22] #63 Ouestion CS490.37 According to the CRBR PSAR (Section 4.4.2.5) the procedure used to determine assembly crificing f or the heterogeneous core is based on a 3 loop natural circulation transient with an imposed maximum coolant temperature of 15500F.
Using this method, minimum required flows are calculated and used to determine flows f or 12 orificing zones. The above procedt.re resulted in a minimum core flow of 93.07% of total flow out of a maximum allowed core flow of 94% of total flow. What would the result have been if, Instead of using PEOC, THDV at 3 had been used to def ine the temperature Tg Resoonse The 15500F maximum coolant temperature was only a conservative guideline to quantify transient constraints to be accounted for in the orificing process which optimizes flow allocations.
Other constraints, i.e.,
lif etime, outl et temperature and gradients are also consider 3d and subsequently constraints are quantitatively expressed on an equal basis. The minimum flow required to satisfy the most restrictive of the various constraints is then calculated and the orificing configuration is selected.
Since all the constraints are put on an equal basis, the orificing conf iguration and relative f low al location among the various zones is completely Independent whether PE00 or THDV plant conditions are considered, in particular, Tu is a 2a, PEOC temperature; the equivalent 3, THDV temperature is reported in Table 4.4.3 of the CRBR PSAR.
The same orificing would have resulted using either one of the two temperatures, provided the other constraints were on the same basis (i.e., 26, PEOC or 3sr, THDV).
Using the flows selected in the orf fIcIng process, the resulting Tel paraneters (e.g., temperatures, flow and pressures) are then quantitatively predicted in detail for all the core assemblies as described in Section 4.4.3.3.
QCS490.37-1 Amend. 69
( % _ M B R____
Paga 23 (82-0321) [8,22] #63 Ouestion CS490.38 in Section 4.4.2.6 of the CRBR PSAR there is a discussion of reactor coolant flow distribution at low flow conditions, it is stated there that a system of three computer codes (DEMO, COBRA-WC and FORE-2M) was used to assess the of feet of alI natural circulation cooling on the maximum coolant temperatures in CRBR. Please provide a detailed description of the geometry modeled by each of the codes and of the data coupling between them, i.e., output used as input, for the calculations discussed in the above section. The geometry model Information should include the number and type of assemblies modeled, the number of fuel or blanket rods in each assembly that are modeled explicitly, the LIM model, and the UlS model. Also, please provide detailed results, i.e., temperature distribution and flow rates as a function of time, for the calculations used to arrive at the conclusions presented.
No experimental evidence of natural circulation cooling for the CRBR heterogeneous core geometry is presented in this section. Are there any experimental data?
If not, what type of experiments are planned to demonstrate the conclusions presented?
Resoonse Before presenting the direct response of Question CS490.38 concerning the Section 4.4.2.6 (i.e., Reactor Coolant Flow Distribution at Low Reactor Flows)
Information, the foilowing three points need to be made:
o A detailed DEMO, COBRA-WC and FORE-2M model of CRBRP has not been used for the Section 4.4.2.6 analyses.
Section 4.4.2.6 wil I be amended to clarify tnis point.
o The Natural Circulation transient is the only event where significant flow redistribution would occur; other design events have 57.5% full flow from pony motors and buoyancy ef fects are insignificant.
o CRBRP worst case design and safety predictions have neglected the benefIclal of fects of Inter-and intra-assembly flow and heat redistribution with regard to lowering maximum core temperature predictions. Thus, the phenomena described in Section 4.4.2.6 have not bean used I* the PSAR Naufral Circulation predictions.
Details were explained at the January 26, 1982 NRC/CRBRP meeting and in topical report WARD-D-0308.
With the above f acts in mind regarding the Section 4.4.2.6 Information, the following discussion provides the response to the questions Cerves provided in Section 4.4.2.6 and other independent studies (Ref s.
o QCS490.38-1 and 2) which found similar trends exemplify mnservatism of neglecting flow and heat redistribution.
Information on flow redistribution shown by Figures 4.4-66 and 4.4-67 was calculated by a preliminary model using the C2RINT11 code.
Figure QCS490.38-1 shows a schematic of the core parallel flow network modeled for these studies. An average channel in each type assembly was analyzed. The LIM model used is described in Sections 4.2 and 4.4.
As can be noted in the figure, the UlS was neglected for this initial study. The typical fuel hot rod transient temperature data exempilf fed on Figure 4.4-68 were calculated for FFTF f uel QCS490.38-1 Amend. 69
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - L2 m M h 8 ______ __________A
Prge 24 (82-0321) [8,22] #63 rods under natural circulation cooling using the system of computer codes:
DEM6, COBRA-WC and FORE-2M, described below. Due to prototypicality of the f uel and system designs, the same trends as found in FFTF would occur for TBRP.
CRBRP has a system of three computer codes (DEH5, WBRA-WC and F6RE-2M) o verified to reduce maximum core temperature predictions from those presented to-date which neglect Inter-and intra-assembly flow and heat redistribution. Current analyses (which conservatively did not use the system of three codes) result in a large 150 to 4000F margin to the coolant boiling criterion used for judging natural circulation capability (described in WARD-D-0308 and at the January 26, 1982 NRC/CRBRP meeting for heterogeneous core design). Since the margin to coolant boiling is so large on the conservative basis, there is not need to use the codes for PSAR predictions. However, predictions from this system of codes will be included in the PSAR.
Details of the system of three codes (including coupling between them and experimental verification) are provided in
" Verification of Natural Circulation in Clinch River Breeder Reactor Plant
- An Update". The analysis procedure with regard to the input / output and sequencing between the three codes is shown by Figure QCS490.38-2.
o Figure QCS490.38-3 shows experimental data and results of analyses with DEMO, COBRA-WC and F6RE-2M f or the highest temperature FFTF f uel rod during a prototypic natural circulation tests. This Information was discussed at the January 26, 1982 NRC/CRBRP meeting. The top two curves of this figure show the ef fect of Inter-and intra-assembly flow and heat redistribution reducing worst case temperature predictions such as those in WARD-D-0308.
It can be noted that both of these predictions are extremely conservative (due to the uncertainty factors applied) relative to the experimental test data and expected predictions (nominal) glen by the low curves on the figure.
o Extensive experimental data (e.g., core component pressure drop and sodium heat transfer testing in f uel and blanket assemblies over wide range of conditions, decay heat, pump coastdown characteristics, FFTF and EBR-ll natural circulation tests, etc.) are already available as described in the report, " Verification of Natural Circulation in CRBRP Plant - An Update".
i
~
o ?.:::;t:n : T::t Phase natural circulation experimental will be performed to
- ".. :.:^r :t: CCP natural circulation capability.
Ref erences: A)
A. K. Agrawal, et al, " Dynamic Simulation of LMFBR Plant Under Natural Circulation", ASME Rapan 79-HT-6,1979.
B)
M. Khatib-Rahban and K. B. Cady, " Establishment of Buoyancy-induced Natural C!rn.lation in Loop-Type LWBRs", Trans.
Amer. Nucl. Soc.
28, pp. 432-433, June 1978.
QCS490.38-2 Amend. 69 May 1982
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OCS490.38 6
,Page 23 (82-0321) [8,22] #63
_Ouestion CS490.39 in Section 4.4.2.8.5 there is a discussion of fuel-cladding gap ef fects on peak cladding temperatures reached during an undercooling transient. The discussion concludes that under LOF conditions with scram it is conservative to overestimate heat transfer to the cladding early in the transient, f.e., a higher peak cladding temperature will be calculated. Please provide quantitative justification, i.e., transient temperature results, for this conclusion.
Resoonse The worst case undercooling transient of PSAR Section 15.3 is the " loss of of f-site electrical power" event reported in Section 15.3.1-1.
This transient has been updated with FORE-2M in Section 15.1.4 for the heterogeneous oore design using a fixed gap conductance model. For fuel assembly #52 which contains the highest cladding temperature hot rod f or any oore location at any time in life, a maximum cladding temperature of 14550F was calculated in a 36 basi s.
In response to the above question, this analysis was repeated with the FORE-IM variable gap conductance model (both the Section 15.1.4 update and this new evaluation having the same initial gap conductance at time zero).
Because of the mechanisms described in Section 4.4.2.8.5 and referred to in the question, a 50F decrease Ir. maximum cladding temperature was f ound due to the lowering of the gap conductance f rom its initial value as the cladding expands proportionately more than the f uel. A comparison of the maximum cladding temperatures f or the two cases is given by Figure 490.39-1.
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QCS490.39-1 Amend. 69 May 1982
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