ML20028G349
| ML20028G349 | |
| Person / Time | |
|---|---|
| Issue date: | 12/09/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| 020183, 20183, ACRS-2051, NUDOCS 8302080470 | |
| Download: ML20028G349 (28) | |
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12/9/82
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ACRS ECCS SilBCOMMITTEE MEETING l1.
$1 DECEMBER 2-3, 1982 SAN JOSE, CA PURPOSE lhe purpose of the meeting was to discuss with GE a revision to their ECCS EM codes (SAFER /GESTER), and to hear presentations on the pros and cons of the use of electrical vs nuclear heater rods for LOCA tests.
ATTENDEES: Principal meeting attendees are noted below:
ACRS GE M. Plesset, Chairman G. Sherwood J. Ebersole, Member J. Quirk D. Ward, Member J. Wood I. Catton, Consultant G. Dix V. Schrock, Consultant J. Anderson C. Tien, Consultant G. Potts T. Theofanous, Consultant B. Shiralkar Z. Zudans, Consultant M. Alamgir P. Boehnert, Staff *
- Designated Federal Employee NRC T. Collins, NRR MEETING HIGHLIGHTS, AGREEMENTS AND REQUESTS 1.
Dr. Plesset noted that we now have a great deal more understanding of the LB LOCA phenomenon and that an ordered approach to revision of Appendix K is timely. He said there are many physical errors in Appendix K.
Dr. C?tton expr-ssed concern with planned modification to the decay heat curve in Appendix K on a piecemeal basis without looking at the whole picture. Professor Schrock expressed displeasure with the slow pace of action by NRC in failing to promptly implement the state of the art knowledge inherent in the use of the new decay heat curve. Dr.
Plesset said Appendix K should be changed so as to enhance the potential of efficient nuclear plant operation while maintaining necessary licensing conservatisms.
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ECCS Meeting December 2-3, 1982
~ 2, GE (G. Sherwood) began their presentation by noting that the changes proposed in Appendix K and licensing analysis are designed to allow
- enhanced plant ' operation and will stop the " chasing of: a few degrees of PCT" since most'BWRs are licensed at the 2200*F PCT limit.
3.
Mr. J. Quirk provided an overview of the meeting purpose which was:
(1) provide technical descriptions of SAFER, GESTR-LOCA, and TRAC-BWR ECCS models, (2) provide ECCS model qualification and assessment results, and (3) describe ECCS evaluation methodology.
4.
Mr. J. Wood discussed the GE ECCS analysis approach. He reviewed the GE LOCA EM noting that the current approach of bounding each physical phenomenon in the model leads to violation of physical laws (conservation ofmass, energy,etc.). This approach has contributed to the present licensing difficulties. We now have sufficient knowledge to move towards a more realistic BWR plant system model (TRAC) and GE has submitted a revised licensing model for NRC review and approval (SAFER /GESTER). The
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new licensing model has been benchmarked with TRAC code calculations.
Mr. Ward asked why GE doesn't treat their new licensing model as BE then add margins to that. Mr. Wood said that practical constraints prohibit a true BE model for licensing use (1D vs 3D calculations).
In response to Dr. Catton, Mr. Wood said GE is attempting to develop a faster running TRAC model; however, even the faster running TRAC code would probably still be impractical for licensing use.
f Mr. Wood noted three bases for the new licensing EM. These are:
(1) use physically corisistent conservation models, (2) input expected value correlations, and (3) compare calculated margin with reasonable l
uncertainties. Mr. Schrock questioned the GE approach to the use of the 1979 ANS decay heat curve. He said he would like to see a clear state-ment from GE on just what they want approval for by NRC in this area.
In response to Dr. Plesset, Mr. Wood said the new GE licensing code is
2 ECCS Meeting December 2-3, 1982 ID and does satisfy all conservation laws. Dr. Tien suggested that the licensing models should track uncertainty propagations and sensitivity levels' in the submodels of the code. He also said that the code should have internal comparisons or checks on the calculatbns performed.
5.
Mr. Dix (GE) described the major BWR LOCA/ECCS safety research facilities both in the US and Japan (Figure 1) that were used to support the new GE licensinC model. He indicated that the test results obtained from these various facilities were complerantary despite the fact there were both single and multiple channel facilties used, as well as both heated and unheated pins simulated. Single channel tests identified large margins in PCT via the lower plenum bundle inlet CCFL phenomenon and high heat l
transfer rates. Mr. Dix detailed the core region water level response seen in the TLTA facility during a LB LOCA (Figure 2) to illustrate the above point. Dr. Tien asked if there is alot of data available on the lower plenum CCFL since it is so important to the core cooling. Mr. Dix replied in the affirmative. Multiple channel experiments show good agreement with singis channel results and tow pin. temperatures were seen for all break sizes tested. Dr. Catten questioned whetner the single i
bundle SAFER code model could adequately describe multiple bundle effects.
He suggested GE use a three parallel channel model in order to more fully encompass the effects of core power variations.
In summary, Mr. Dix said the above experiments show:
(1) excellent empirical understanding'of BWR LOCA/ECCS response, and (2) the experimental basis is diverse and complete enough to challenge and qualify models. The GE modeling approach is to develop the BE TRAC model and a realistic EM (SAFER). SAFER will be qualified with test data and TRAC BE calculations.
As a result of numerous Subcommittee questions centering on the details of the SAFER /GESTER models and the verification of these models via experimental data, Dr. Plesset said that another Subcommittee meeting should be held in the nece future to fully discuss these details. [ Note m.
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ECCS Meeting December 2-3, 1982 l
a Subcommittee meeting is tentatively scheduled in February. The meetirg will be closed in order to discuss GE proprietary information.]
l 6.
Mr. Anderson (GE) discussed the development of the TRAC best estimate model.
Figure 3 shows the TRAC model capabilities. The GE version of tne code, TRAC B02, is based on the INEL BD1 Version 12 code.
In response to.
Mr. Schrock and Mr. Ward, Mr. Anderson said the GE TRAC version is an order of magnitude faster than the "D" (cetailed) TRAC version. Figures 4 and 5 list the basic and component TRAC models used by GE.
Development of selected models and assessment of the TRAC code were reviewed.
Details of the jet pump, steam separator, steam dryer, and upper plenum models were g'ven. Mr. Ebersole asked how, given a LOCA, the non-jet pump (JP) plants are cooled without reflooding as is seen in JP plants. GE said the core spray adequately cools these plants which are of lower power density than later model (BWR 4-6) JP plants.
The upper plenum model was shown under Subccamittee questioning to be lacking in detailed docurentation. The void frhetion model calculations and associated CCFL predictions (interfacial shear model) were shown com-pared against data. Dr. Catton requested details of the bundle inlet CCFL model. GE was not prepared to address this item at this meeting; this material 13 GE proprietary. Dr. Plesset requested that this information be provided in written form to the Subcommittee and discussed at the future meeting noted above. Mr. Schrock said that the two-fluid model used in TRAC has limitations, particularly for reflood calcul ations. GE agreed that the model has problems at lcw pressures and that develoment of the GE TRAC code has some way to go yet. Dr. Catton noted that all advanced codes have problems with reflood predictions.
7.
Mr. Potts (GE) discussed the GESTER fuel rod model. The GESTER single I
rod model provides rod initial conditions at the onset of a LOCA (fuel stored energy and fission gas inventory). NRC still has the mode) under review.
Figure 6 shows the GESTER model elements. The thermal and a.
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ECCS Meeting December 2-3, 1982 mechanical model, and fistion gas release and rod internal pressure l
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calculations have all been qualified from Halden IFA test results.
8.
The SAFER model was detailed by B. Shiralkar (GE). SAFER is a "better-estimate" thermal-hydraulic transient code for calculation of long term inventory analysis given a LOCA or other off-nonnal reactor transient.
SAFER is based on a combination of the currently used EM codes SAFE and REFLOOD.
He noted that SAFER addresses what GE believes are excessive conservatisms in:
(1) current LOCA EM conservatisms, and (2) excessive conservatisms addressed by SAFER.
Figure 7 outlines the BWR EM char.ges as.a result of incorporating the new SAFER /GESTER codes into the improved evaluatirn model.
In response to Dr. Catton, Mr. Shiralkar said SAFER calculations will be compared to TRAC multi-channel analyses. SAFER provides Appendix K calculations with an uncertainty " adder" on the PCT result. Figures 8 and 9 describe the 5AFER model improvements and the impact on PCT (major or minor). The major impacts are in the areas of:
(1) improved inventory distribution and (2) steam cooling, tranrition tniling.
9.
Mr. Shiralkar provided an overview of the SAFER models (hydraulic, regional (bypass leakage, upper plenum, hot fuel assembly), external core flows, t. eat transfer, and fuel rod). Key discussion items noted ficluded:
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- Dr. Catton questioned the assumption of perfect mixing in the lower plenum. GE said they would provide details of this for the next Subcommittee meeting.
- There was detailed discussion of the CCFL correlation used
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(modified Wallis). Additional discuJsion was postponed to l
the next meeting as noted above.
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- Drs. Tien and Schrock commented on the use of the Dittus-Boelter correlation for heat transfer. They both felt use of this correlation is questionable and a better correlation needs to be developed.
ECCS Meeting December ?-3,1982 i
GE believes the SAFER /GESTER codes are a great improvement over the current EM now in use.
- 10. Dr. W. Craddick (ORNL) discussed ORNL investigations into the differences in thermal behavior of electric and nuclear fuel rods. ORNL addressed two objectives:
(i) determine how power should be varied through time in an electric rod to best simulate nuclear rod behavior, and (2) analyze post-test electric rod behavior to determine what could be inferred about nuclear rod behavior. A code called PINSIM was used to analy:e rod behavior. Results of the investigations indicated that current electric rods cannot match nuclear rod behavior. ORNL concluded that it was doubtful that the cost of "more realistic" electric rods is justified by the benefit. Dr. Plesset noted that while it is difficult to simulate nuclear iods with electric rods, che use of analyris can overcome the differences between them. Dr. Craddick agreed with Dr. Plesset's comments,
- 11. Mr. T. Knight (LANL) also addressed the use of electric versus nuclear rods. Key points of his presentation include:
- LANL considers power generation in the core to be a boundary condition. Deposition cf the enc'gy into fluid is affected by heat transfer in the rod and by the heat-transfer co-efficients to the fluid. The heat-transfer coefficients are common to both nuclear and electrical rods.
Nuclear fuel rods and electrical rods have different heat capacities and conouctivities.
These parameters together with distributed heat generation affect the amount and the location of the stored energy.
If we assume that the rod j
configuration is not changing, these differences can be modeled easily in the heat. conduction model of the rod.
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ECCS Meeting December 2-3, 1982 The nuclear fuel rod can exhibit cladding swelling and rupture and fuel cracking. These phenomena can change the heat-transfer characteristics of the fuel rod. Generally, the state / condition of the fuel at the initiation of a transient is not well established. Therefore, the stored energy in the fuel is not well known and may affect peak cladding temper *+ure and quenching.
LANL believes that any problems associated with nuclear fuel rods can be resolved by adding a more detailed fuel-rod model to the TRAC code and by running a steady-state fuel-rod code to establish initial conditions. Separate-effects fuels tests should be sufficient to support code development and assessment in this area.
i
- 12. Mr. M. Alamgir (GE) reviewed the GE effort on qualification assessment of the TRAC B02 cme. The qualification data base is shown in Figure
- 10. Figure 11 shows the specific essessment objectives of the tests.
The data base includes TLTA and SSTF results as well as Peach Bottom turbine trip transient tests. Dr. Catton asked if GE participated in the International Standard Problem Program. GE replied in the negative.
In response to Mr. Schrock, Mr. Shiralkar said B02 will not be the final f
code version, rather it will be used internally by GE to develop the i
final produ:t.
i Mr. Alamgir reviewed the comparisons of TRAC calculations for the tests noted in Figure 11, In general, re alts shown were in good agreement wito the data. There was some question on tne capr5111ty of the reficod model results. Further discusCon led the Subcommittee to recommend GE take a closer look at their test facility to assure adequate high quality data is available to verify the code models being assessed.
In cc...clusion, P.r. Alamgir said the assessment study is still on-going.
Areas for improvement cited by GE were:
(1) droplet field for dispersed
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flow, and (2) liquid withdrawal near a pool due to the Bernoulli effect. Mr.
Schrock indicated that he was not satisfied with the tone of the conclusion
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that "all is well" with the code at this stage. There are development areas that nec) improvement and should be acknowledged. He tempered his remark by noting that he favors the GE effort to make use of a best-estimate code for licensing use. Dr. Plesset.said he vas encouraged to see GE cooperating with INEL to improve the TRAC BHR code.
Dr. Tien felt that GE has, in gene al, done a good job with the code but he too urged GE to be more forthcoming in discussing the specific problem areas in cade development that we all know exist. Dr. Catton felt all vendors shoulct, as GE is doing, move toward development and use of best-estimate codes. Dr. Plesset agreed with this comment.
- 13. The SAFER assessment effort was detailed by B. Shiralkar. He :howed results from the TLTA facility and noted preliminary comparisons with p
data from the Japanese ROSA-III facility. GE believes SAFER captures the significant phenomena and trends of these tests.
Figure 12 lists the Tt.TA tests selected for SAFER assessment. Mr.
Shiralkar detailed the results of the code comparison with the test data.
Figure 13 shows a comparison of predicted vs measured PCT's for the 1LTA experiments. For the ROSA tcsts, GE only had oral informa-tion for the SB LOCA results. The agreement between predicted and measured PCT was described as excellent. Dr. Catton urged GE to consider participation in the International Standard Problem Program via the use of a blind pretest prediction as a good test of the GE codes described
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at this reeting (TRAC and SAFER /GESTER).
- 14. The results of TRAC B02 calculatt.ns for a BWR/6 plant were reviewed.
The transients analyzed and assumptions mat's for the calculations are shown in Figure 14.
Figures 15 and 16 show the details of the code nodalization used by GE. The highest PCT seen for the calculations shown 4
(MSIV, HPCS, and Recirculation line breaks) was 782*F. Dr. Plesset as'.ed if GE had looked at the relative probabilities of the above events.
GE said these events are of low probability and are not substanstial risk contributors.
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15.
B. Shiralkar discussed the evaluation methedology proposed by GE to address unce"
' ties in the SAFER /GESTER PCT calculation. GE proposes the use of an " adder" to the nominal PCT calculation to obtain an upper bound PCT. Figures 17 and 18 detail the actual adder calculation metho-dology. The adder would account for code model biases and unc2rtainties.
A plant uncertainty contribution is also included to account for, amung other things, the Appendix K recuirements (Figures 18. Item a). A poten-tially questionable aspect of the methodology is the plan to combine the contribution of the Appendix K requirements in t?.e adder vis SRSS (square-root-of-the-sum-of-the-squares). As noted below, this whole topic of treatment of tile uncertainties is still under NRC review.
Mr. Ward felt that the GE approach to the treatment of uncertainties is an excellent one since all the constitutive elements are layed out in the open for examination and review.
- 16. GE, in su. mary, said they believe they have developed a good analytical effort closely coupled to their experimental programs for implementi,ng a more realistic approach to ECCS licensing anelysis. Dr. Plesset said that the GE effort it commendable for moving to a best-estimate approach for ECCS/LOCA analysis.
17.
T. Coll'ns (NRC-NRR) p.ovided a summary of the status of the NRC review of SAFER /GESTER.
The GESTER review is well along and the NRC reaction is generally favorable. The 'GESTER code review can be completed in January, pending timely receipt of additional information fr'n GE.
The SAFER review is not as far along as GESTEn. NRC has a number of outstanding questions. The most important issue appears to be the treatment of uncertainties (adder) as noted above. Mr. Collins speculated that the treatment of decay heat and fuel stored energy will probably be the major issues of the review. Assuming receipt of the required information from GE by a scheduled January 26, 1983 NRC/GE meeting c' ate, NRC can t.omplete its SAFER review by mid-March 1983.
4
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ECCS Meeting 10-December 2-3, 1982
- 18. The Meeting was adjourned at 2:30 p.m.
A transcript of the open portice of the meeting is available in the NRC Public Document Room at 1717 H Street, N.W., Washington, D.C., or can be obtaired at cost from Alderson Reporting, 400 Virginia Avenue, S.W.,
Washington, D.C. 202/554-2345.
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f PLAJOR BWR LOCA/ECCS SAFETY RESEARCH FACILITIES
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18' SECTOR TEST FACILITY - (TOSHIBA)
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Increased Flexibility for Water Makeup System Simulation, Operator Action.
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GREATER DETAIL AND ACC'JRACY ALL MAJOR REGIONS MODELED MINOR CORE AXIAL SUBDIVISION MINOR HOT CHANNEL CALCULATION 6ESTR FUEL MODEL MINOR SAFE /REFLOOD INTERFACE ELIMINATED
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- 'l s
s s
=
3 t
l m_
l a
?
t I
g
_.l..
_1 l
l_
g
.l.
1 I
I i
TRAC _. BWR/s WobhLFfATt04 f/6l[y)
5:
~
.=...:..
........,_.n...
J*
TRAC NODALIZATION VESSEL:
11 Axial Levels 4 Radial Rings 44 Vessel Cells s..
1 8 Sector Total Number of Components = 33 CORE:
3 groups of Channels (11 cells, 9 heated)
O l
s.
{
t
..s:.. :. :. -
. =-..
. T.
~
t, ADDER CALCULATION PCTJpger Bound ".PpJSAFER/GESTRNOMINAL
+T[
E 2
2 2
Jgder = A2+A2 + tr + r
)
A2-Average Blas of Experiment-TRAC values of PCT.
Assumed to apply in plant.
Ac,coun' t
P 'for A2-Average Blas of (TRAC) plant values of PCT relative Model to SAFER /GESTR Plant values for the same LDCA.
Bias
~
Accounts for simplified models in SAFER /GESTR, 2
.r
= Adder contribution due to variance of TRAC -
Experiment Values.
Account for Model Uncer-r Adder contribution due to variance of SAFER /GESTR tainties.
2 l
TRAC values.
(
t f
l"F/&17
............~.......u....
......w.
=
rj =
Adder contribution due to variance of distribution of uncertainty of SAFER /6ESTR Plant values.
This reflects uncertainty in:
a)
Va,"lables whose values are conservatively specified in Ap'andix K
. Decay Heat Maximum Temperature for Transition Bolling Break Flow Model e
g, Metal Water Reaction Rate Coefficients b)
Varlobles whose values were much better known in the experiments than in a plant Core Power Peak Linear Heat Generation Rate Bypass Leakage Coefficients Minimum Critical Power Ratio ECCS Water Temperature ECCS Initiation Signals y
c)
Variables which were not involved in the experiments Pellet-Clad Gap Conductance
(
Fuel Rod Internal Pressure.
i
,