ML20236F168
| ML20236F168 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/31/1987 |
| From: | ABB IMPELL CORP. (FORMERLY IMPELL CORP.) |
| To: | |
| Shared Package | |
| ML20236F148 | List: |
| References | |
| 01-0300-1625, 01-0300-1625-R01, 1-300-1625, 1-300-1625-R1, TAC-65471, NUDOCS 8708030236 | |
| Download: ML20236F168 (16) | |
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Tro,$an Nuclear Plant Document Control Desk
. Docket 50-344 July 27, 1987
' License NPF-1 Attachment B
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SUMMARY
REPORT INDEPENDENT REVIEW OF ACTION PLAN ITEM 5 T0 RESOLVE THE TROJAN MAIN FEEDHATER PIPING RESTRAINT FAILURE ISSUE Prepared For:
PORTLAND GENERAL ELECTRIC.
121 S.H. Salmon Street Portland, Oregon 97204 I
Prepared By:
Impell Corporation 350 Lennon Lane .
Halnut Creek, California 94598 l Job No. 0300-039-1336 ,
! Report No. 01-0300-1625 i l Revision 1 i i
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(T N July 1987 j l
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B70B03023b DR B70727 I ADOCK 05000344 PDR 1
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U l REPORT APPROVAL COVERSHEET Sumary Report - Independent Review of Action Plan Items Report
Title:
to Resolve the Tro.ian Main Feedwater Pioino Restraint Failure . Issue Report Number: 0300-162s Revision: o ,
0300-039-1336 .
CHent. Portland General Electric Job Number. q l
Project: Trojan: Start-up-Assistance RECORD OF REVISION DATE PREPARED REVEWED* APPROVED REVISON k 0 7//S/87 SW W%JW WS)"
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- Independent Review per QP 3.6
Sumary Report - Independent Review of Action Plan Items to Resolve the Trojan Main Feedwater Piping Restraint Failure Issue Revision Sumary Rev. No. Paae No. Description 0 - Initial Issue 1 6 Changed " generation" to " generator" in first paragraph changed"'...during the 1979 outage...." to l
...in October,1979,. . ." in second paragraph Expanded Action Plan title in last paragraph of Introduction; revised second sentence of second paragraph of Introduction.
1 7 Provided rise times of 100 psi- dp water-hamer loading in item 8.
Made slight revision to first paragraph of Sumary and Recommendations section.
Changed reference and items in iast paragraph from
... item 2 and 4..." to "... items 3 and 4 below..."
1 8 Clarified language and corrected typos by aaking minor changes to findings and recommendations 1, 2, and 3.
9 In first paragraph changed last phrase of first sentence to read " ..the modified support should be more than adequate for the expected loads."
Deleted last sentence of first paragraph of item 4a; added second sentence of third paragraph of item 4a.
Changed last phrase of second paragraph of item 4a from ."...were written off during that effort" to
...were determined to be of minor significance during that effort".
Added sentence to paragraph of item 4b to better highlight the differences in relative snubber locations.
1 10 Corrected deadweight reaction from "778 lbs" to "541 lbs" in second paragraph of item 4d.
, Rewrote item 6 to clarify impact of thermal l stratification calculation.
l (r),) 1 11, 12 Corrected Reference 9 to give PGE calculation number; added Reference 14.
Report No.01-030 % 1625 Revision 1 Page 3 of 12
e TABLE OF CONTENTS Section Pace Approval Coversheet 1 Report Title Page 2 Revision Sumary 3 Abstract 5 Review Program Introduction 6 Scope and Program Description 6 Sumary Results and Recommendations 7 References 11 O
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D Report No. 01-0300-1625 Revision 1 Page 4 of 12
ABSTRACT Portland General Electric (PGE) retained Impell Corporation to perform an independent review of certain items within the action plan developed to i resolve the Trojan main feedwater piping restraint damage issue. Specifically I reviewed were the thermal-hydraulic analyses, piping structural analyses, and pipe support evaluations performed by Bechtel Western Power Corporation staff and the support failure evaluation performed by PGE staff.
All pertinent calculations and reports are now essentially complete and have been reviewed ~. Impell agrees with the conclusion that the failure was caused by a- thermal-hydraulic event, most likely a steam bubble collapse, perhaps acting in conjunction with thermal stratification within the line. Further, all analytical techniques used were appropriate and the modification- to SR-8 vill strengthen the support sufficiently to mitigate possible damage to the !
piping system under similar events in the future. (
1 Before the issue is completely closed, minor discrepancies as noted in the results and recommendation section, should be resolved. These do not invalidate the results and should not prevent plant heat-up, however. j O
G i Report.No. 01-0300-1625 Revision 1 Page 5 of 12 _ - _ _ _ _
INDEPENDENT REVIEN PROGRAM AND StM4ARY RESULTS AND RECOMMENDATIONS I
l REVIEN PROGRAM j l
. Introduction -
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' Inspections of the Trojan feedwater system piping. conducted during the 1987 )
. refueling outage revealed that rigid restraint SR-8 on the "B" steam generator '
main feedwater line.had failed sometime since the previous refueling outage.
The support had been pulled away from the concrete deck to which it was attached.
The severity of this event and its similarity to a failure at the same support (SR-4) on the "A" main feedwater line, discovered in October,1979, indicated
-that water-hammer events may have occurred. After the SR-8 failure, an action:
plan was developed to determine the extent of possible damage to the feedwater system, determine the likelihood of any.further events wnich could' impair the-system's-functionality, and to develop corrective actions to preclude future occurrences.
Important portions of the action plan entail piping thermal-hydraulic analyses, piping structural analyses, and-pipe support evaluations, largely performed by the Bechtel Hestern Power Corporation (BHPC). To confirm the validity and technical adequacy of these analyses, Impell Corporation was contracted to perform an independent, or third party, review of this analytical work.
Specifically, the review was intended to confirm the following:
- 1. Validity of the evaluations of the potential root causes
- 2. Technical adequacy and appropriateness of the analytical techniques utilized.-
- 3. ' Applicability and reasonableness of the results achieved.
The review program is item F1 of PGE's Action Plan for the Main Feedwater Restraint Failure and is described more completely in the following sections.
Scone and Procram Description The following action plan (Reference 13.0) activities are verified under this independent review program:
Activity No. Description Action Party A 2. Evaluate load on SR-4 (SR-4 found to BNPC/PGE be in compression; pipe moved 33 mils when restraint was removed)
B 2. Calculation of failure load of SR-8 PGE B 3. Piping structural analysis of "B" line with and BHPC gm without SR-8 under existing design conditions B 4. Perform bounding thermal-hydraulic analyses BHPC Report No. 01-0300-1625 Revision 1
} Page 6 of 12
The inputs into the review program consist of the following:
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- 1. Original piping and support design bases
- 2. Restraint SR-8 failure load evaluation
- 3. Loop B analyses with SR-8 intact
- 4. Loop B analyses with SR-8 failed
- 5. Loop A analysis for 33 mils cold spring at SR-4
- 6. Thermal-hydraulic analyses of possib1'e bounding water-hammer events
- 7. Loop B thermal stratification analysis i 8. Linear elastic analysis of Loop B with a 100 psi dp water-hammer loading with rise times of 0.01 sec. and 0.001 sec i
- 9. Bechtel report on the evaluation of the main feed line seismic restraint failure
- 10. Bechtel report on the evaluation of the displacement at main feedwater restraint SR-4
- 11. Bechtel final report on the SR-8 failure root cause evaluation Each of the inputs is reviewed by engineers experienced in the areas of piping structural analysis, pipe support design, or thermal-hydraulic analysis.
Their findings, conclusions, and recommendations for corrective actions are then summarized and reviewed by senior supervisors and technical managers for completeness, applicability and accuracy.
SIM4ARY RESULTS AND RECOMMENDATICi4S l As of July 13, all items of the scope have been essentially completed and reviewed.
Impell concurs with the conclusion that the support failure was caused by a
, thermal-hydraulic dynamic event, most likely.a steam bubble collapse, perhaps
! acting in conjunction with thermal stratification within the fluid of the ,
J line. All analytical techniques utilized and results to date are appropriate I and adequate, and the corrective action to be taken with respect to the piping configuration, i.e., repair of the concrete anchorages of SR-8 by employing through bolts, is reasonable and will mitigate possible damage to the piping system under similar events in the future.
While the analytical results to date are valid, minor discrepancies, as noted in items 3 and 4 below should be resolved before this issue is completely closed. ,
Report No. 01-0300-1625 Revision 1 Page 7 of 12
The complete resolution to this issue also includes actions to alter plant operations, augment maintenance procedures, and develop a monitoring program
'to preclude and minimize future water-hamer events. These are beyond the scope of this review but their implementation should greatly help prevent water-hamer events of sufficient magnitudes to cause piping and support
' failures.
Impe11's findings and recommendations are:
14 The bounding thermal-hydraulic analyses include preliminary evaluations of water-hamer loadings due to fast valve closure events These loadings produce lower thermal-hydraulic forces and help confirm that steam bubble collapse is the most likely cause of the water-hamer event. This analysis should'be finalized and included in the documentation of the resolution to this issue to complete the analytical files.
- 2. The equivalent static failure load of 40 kips calculated for SR-8 should be viewed as an upper bound for the following reasons:
- The Phillips redhead anchors were designed only to be embedded 2-1/2" into the concrete slab which is on the order of magnitude of concrete cover, thus only unreinforced concrete was engaged.
- While the interaction between pullout and shear at anchor bolts is complicated and difficult to quantify, the very high shears the anchorage experienced could have acted to reduce the strength of the concrete failure cone and caused failure at a load less than 40 kips.
- The allowable load of the strut of SR-8 is 20 kips under faulted conditions. Under a 40 k*p applied load some distress, or even failure, would be expected, however, no distress was indicated to have been observed.
- 3. In the discussion of the restraint failure engineering analyses submitted to the NRC (Reference 12.1) it is indicated that the design capacity for SR-8 in its repaired configuration is greater than 40 kips. The support qualification calculation, however, employs a faulted load based on the latest analyses of the existing design conditions which is on the order of 10 kips. PGE should clarify to Bechtel if the entire support should be qualified for 40 kips or only the anchorages, and, if only the anchorages, is the 40 kip load to be applied as only a pullout force or some combination of pullout, shear, and moment? The support in.its current configuration is not qualified for 40 kips as the strut employed is only good for a 20 kip faulted load. Also it is not readily apparent that the anchorages, welds, and members could take such a force.
O Report No. 01-0300-1525 Revision 1 Page 8 of 12
Since the strut has a significant margin of safety.(10k applied load versus 20k allowable), the proposed modification greatly-strengthens the anchorages, and measures to be taken will greatly i reduce the likelihood of occurance of future water-hammer events, ;
the modified support should be more than adequate for the expected loads. .The resolution of the above issue should not-delay plant heat-up.
Also, the details of the modified support currently shown in the calculation are incomplete as they.do not indicate that the 5/8 inch diameter threaded rods. are grouted into the 1 inch diameter holes through the concrete deck. This.information should be :
officially conveyed to Bechtel and incorpor ted into the calculation to.close out the support evaluation. l
- 4. The new analyses performed with SR-8 in place should be finalized and items listed below resolved before closing out this issue. The following specific findings and recommendations are made:
- a. The original-1970 stress analyses incorrectly modeled SR-8 as a two-way XY stop located on the horizontal portion of the elbow while the 1975 stress analysis thermal runs did not include SR-8, even though this is an active support under thermal loads.
i The IE 79-14 walkdown results indicate the supports in their correct positions and geometries. The discrepancies between l
. the as-built and the original as-designed configurations were determined to be of minor significance during that effort.
l A' review of the new loads also showed that these are compatible with the support design and confirm that the support failure was not due to an inadequate original design.
Both the 1970 and.the 1975 analyses are now superseded.
- b. Loop B was qualified by comparison to analyses of the Loop A piping, which.is generally mirror image to Loop B. The three snubbers on Loop B, however, .are in sufficiently different locations to warrant a separate seismic analysis. The
- relative snubber locations differ by up to 5 feet while SS-2 of Loop A is located on a horizontal run of piping while the-comparable snubber, SS-6, of Loop B is on a riser. It is recommended that this separate analysis be performed,
, e:pecially since the new analyses will become the analyses of record.
- c. In the thermal stratification calculation, the ' temperature
- l. difference at the two horizontal elbows of the model is l applied in the horizontal rather than the vertical direction.
If this calculation is to become a part of ths permanent plant
- record and not be just a scoping study, this discrepancy
- should be corrected.
Report No. 01-0300-1625
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Revision 1.
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- d. In requalifying the pipe supports the original seismic loads were used rather than the loads from the latest analyses.
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Since the new loads.are smaller, the support design loads are conservative. This should be noted in the calculations or corrected.
Also, at support SR-8 the dead weight reaction of 541 lbs was not considered. Since the old seismic load used in the j support qualification is 721 lbs larger than the new seismic load this discrepancy is offset and doesn't impact the support qualification. This discrepancy should be noted in the calculation, however.
- 5. The steam bubble collapse analysis was performed-using the dynamic slug model described in NUREG-0291. The analysis assumed two symetrical slugs on each side of the bubble moving toward each l other, and included sensitivity studies for the bubble volume, pressure- difference, and slug length. For completeness this I analysis should include:
l a. Sensitivity of the calculated loads to variations. in the height of the bubble (i.e., void fraction at the location of the bubble).
- b. A study of the situation where a single slug on the steam generator side will be accelerated toward the bubble must be
-- included. This is a more realistic representation of the i situation and will indicate the level of conservatism in the symmetrical slug model.
- 6. The thermal stratification calculation is based on a unit " load" of a 100*F temperature difference across the piping cross-section. At this temperature difference a 5.4 kip force acts at SR-8.
Due to the large differ 9nce between the MFH and AFH temperatures, variations larger than 100*F are possible, however. Under this situation relatively high loads at SR-8 could occur. It would then not take a significr.nt hydro-dynamic type event to overload the support.
This issue should be enmined in more detail over the next fuel cycle to determine the range of temperature differences the piping actually experiences under plant operation. Depending on the magnitude of the resulting loads, a decision as to whether or not this condition should be incorporated into the design basis could be made at that time. Since the support anchorages have been significantly strengthened and measure will be taken to greatly reduce the likelihood of water-hammer events, this issue should not impact plant haat-up.
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! Report No. 01-0300-1625 Revision 1 Page 10 of 12
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REFERENCES
.1.0 i Original 1970 Piping Structural Analysis of Loops A and B of the Trojan Main Feedwater Piping,. Bechtel Calculation 2-1, Rev. 0 2.0 Supplemental 1975 Thermal Analyses of Loops A and B of the Trojan Main .
'Feedwater Piping, Bechtel Calculation 2-16, Rev. 0 '
3.0 Trojan Main Feedwater IE79-14 Walkdown Results (no document no. assigned) 4.0 The following 1987 Analyses-of Loops A and'B, Bechtel Calculation 2-16, Rev. 1:
4 . 11 Loop A Analyses,with a 0 033 in. Cold Spring at the Location of SR-4
- 4. 2- Loop B Analyses with Support SR-8 Removed 4.3 Loop B Analyses with Support SR-8 in Place 5.0 The following Bechtel pipe. support calculations:
5.1 EBB-3-1-SS5, Rev. 0 5.2 EBB-3-1-SS6, Rev. 0
.5.3 EBB-3-1-SS7, Rev. O
\ 5.4 EBB-3-1-SR8, Rev. 0
,,,/ 5.5 EBB-3-1-H7, Rev. 0 5.6 EBB-3-1-H8, Rev. 0 -
5.7 EBB-3-1-H9, Rev. 0 5.8 EBB-3-1-H10, Rev. 0 6.0 Thermal-hydraulic Force Time History Analysis of Loop B with a Sudden Assumed 100 psi Drop in Pressure, Bechtel Calculation 11760/SS02, Rev. 0 7.0- ' Thermal Stratification Structural Analysis of Loop B with an Assumed 100*F Differential Temperature Across the Pipe..Bechtel Calculation 11760/SS03, Rev. 0 8.0 The following Bechtel thermal-hydraulic analyses:
8.1 " Steam Condensation Induced Hater-Hammer", Bechtel Calculation SP-54 (102), Rev. 0 8.2 " Forcing Functions Analy.iis for Feedwater System Valves Closure", SP-54 (101), Rev. 0 (Preliminary) 9.0 Portland General Electric Evaluation of the EBB-3-1, SR-8, Restraint Failure, PGE NPE Civil Branch Calc. TC-450, Rev. O O
Report No. 01-0300-1625 Revision 1 Page 11 of 12
10.0L The following Bechtel Hestern Power Corporation Reports:
10.1 " Trojan Nuclear Plant Evaluation of Main Feed Line Seismic Restraint Failure", dated June 11, 1987 10.2 " Evaluation of Displacement at Main Feedwater Restraint SR-4 in the Trojan Nuclear Plant", dated June, 1987 10.3 " Final Report, Trojan Nuclear Plant, SR-8 Failure Root Cause Evaluation Steam Condensation - Induced Hater-hammer", dated July,1987.
11.0 Portland General Electric Memorandum TEB-55-87M, Trojan Nuclear Plant, 1 Main Feedwater Restraint Failure Action Plan", T.E. Bushnell to R.E.
Fowler, dated June 3,.1987 12.0' The following Portland General Electric Co. letters to the NRC with attachments:
12.1 " Main Feedwater Restraint Failure. Engineering Analyses", David H.
Cockfield (PGE) to USNRC, dated. June 12, 1987 12.2 " Main Feedwater Restraint Failure" David H. Cockfield (PGE) to John B. Martin (NRC), dated June 16, 1987 l 13.0 Main Feedwater Restraint Failure Action Plan, dated June 10, 1987.
( ( 14.0 PGE Trojan Nuclear Plant Maintenance Request MR 87-3253.
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l 0 LJ Report No. 01-0300-1625 Revision 1 Page 12 of 12
Trojen Nuclear Plcint Documint Control Dask Docket 50-344 July 27, 1987 License NPF-1 Attachment C Page 1 of 4 PCE RESPONSES TO IMPELL RECOMMENDATIONS / FINDINGS Recommendation / Finding 1 Finalize the evaluations of water hammer loadings due to fast valve closure events, and include this analysis in the documentation of the resolution to this issue.
PCE Responso PGE agrees with the Impell recommendation that the fast-valve closure water hammer analysis should be finalized and included in the reso-lution documentation flie. Bechtel is presently in the process of finalizing these analyses and expects &han to be complete by Auguct 31, 1987.
Recommendation / Finding 2 The equivalent static failure load of 40 kips calculated for SR-8 should be viewed as an upper bound.
PGE Response No action required.
Recommendation / Finding 3 Clarify to Bechtel whether the entire seismic restraint SR-8 or only j the anchorages should be qualified for 40 kips. If only the anchor- 1 ages, identify whether the 40-kip load is to be applied as only a f pullout force or some combination of pullout, chear and moment. l 1
Officially convey to Bechtel additional details of modified support SR-8. For example, indicate that the 5/8-inch-diameter threaded rods are grouted into the 1-inch-diameter holes through the concrete deck. This information should be incorporated into the calculation to close out the support evaluation.
PGE Response The design load for SR-8 continues to be based on dead weight, ther-mal and seismic conditions. Even though water hammer has occurred in the past, corrective action to minimize the possibility of water ham-mer has been taken so that restraint design loads need not include ;
this load case. Thermal stratification could contribute to restraint i
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l Trojan Nuclear Plcnt Docum:nt Control Desk Docket 50-344 July 27, 1987 License NPF-1 Attachment C !
Page 2 of 4 loading but will only be included as a design load case if data col-1ection confirms that stratification is a significant contributor to restraint loads. See Responses 4.c and 6 below for further discus-sion on this matter.
The previous discussion on SR-8 capacity in the range of 40 kips was presented, not to suggest that this is the " design load" for the support, but rather to demonstrate that significant margin exists to accommodate unusual loads from water hammer and stre.tification above design loads. The current design load for SR-8 is less than 10 kips.
Contrary to the Impell report, the strut currently installed in SR-8 has a maximum recommended design load of 15.7 kips with a significant safety factor to failure. The Level D faulted load for the strut at i 350*F is 26.2 kips.
As-built details of SR-8 will be forwarded to Bechtel in accordance with PGE's standard as-built packago processing procedure and PGE's standard as-built drawing distribution procedure. Finalization of design calculation will be officia11*/ completed in accordance with the Bechtel As-Built Reconciliation Program. However, PGE and Bechtel have already prepared calculations to justify Plant operation based on the final SR-8 configuration.
Recommendation / Finding 4 Resolve the following items and finalize the new analyses performed with SR-8 in place:
- a. The original 1970 stress analyses incorrectly modeled SR-8 as a two-way KY stop located on the horizontal portion of the elbow, while the 1975 stress analysis thermal runs did not include SR-8, even though this is an active support under thermal loads,
- b. Loop B was qualified by comparison to analyses of the Loop A piping, which is generally mirror image to Loop B. Perform separate seismic analyses to account for the different locations of three snubbers in Loop B. The relative snubber locations differ by up to 5 feet, and Snubber SS-2 of Loop A is located on a horizontal run of piping, while the comparable Snubber, SS-6, of Loop B, is en a riser.
- c. If the thermal stratification calculation is to become part of the permanent Plant record, correct the discrepancy in which the ,
temperature difference at the two horizontal elbows of the model l is applied in the horizontal, rather than the vertical direction.
- d. In the calculations for requalifying the pipe supports, use the seismic loads from the latest analyses. Alternatively, note in the calculations that they are conservative, since the original seismic loads were used rather than the loads from the latest analyses, which are smaller.
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.Trojcn Nuc1 car Plcnt Docum:nt Control Desk Docket 50-344 July 27, 1987 License NPF-1 Attachment C Page 3 of 4 Also note in the SR-8 qualification calculation that the old seismic load, which was 721 pounds larger than the new seismic !
load, was used. This offsets the fact that the dead weight reaction of 541 pounds at Support SR-8 was not considered.
PGE Response Action taken or to be taken is as follows:
- a. Bechtel has annotated the 1970 calculation as being superseded by the 1975 calculation and has annotated the 1975 calculation as being superseded by the June 1987 calculation.
- b. PGE agrees with Impell that the differences between the 'A' and
'B' Loop configurations warrant separate seismic analysis rather than relying on mirror image comparisons. However, since the differences are relatively minor and since the seismic snubber support designs include significant margin above design loads, this supplemental analysis will be completed after Plant startup.
- c. System monitoring to be performed over the next operating cycle will include collection of data needed to assess the degree, if any, of thermal stratification. If this data leads PGE to con-clude that thermal stratification contributes significantly to the system design basis, the Bechtel calculation will be final-ized (including resolution of the Impell-noted discrepancy) and made a part of the system design basis documentation.
- d. Since the new seismic loads are lower than previously calculated loads, the Bechtel calculation will either be annotated to reflect the lower load without revising the calculation or will l
be revised to factor in the lower load. If the calculation is revised, the dead weight reaction at the support will be
} considered.
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l The above actions which have not already been completed, will be completed by October 31, 1987.
Recommendation / Finding 5 l l
l The steam bubble collapse analysis was performed using the dynamic slug model described in NUREG-0291, "An Evaluation of PWR Steam Generator Water Hammer". The analysis assumed two symmetrical slugs on each side of the bubble moving toward each other, and included sensitivity studies for the bubble volume, pressure difference, and I slug length, complete the analysis, including:
f f a. Sensitivity of the calculated loads to variations in the height of the bubble (ie, void fraction at the location of the bubble),
l and l
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4-Trojan Nucicar pltnt Docum:nt Control Desk Docket 50-344 July 27, 1987 License NPF-1 Attachment C Page 4 of 4
- b. A study of the situation where a single slug on the steam gene-rator side will be accelerated toward the bubble.
PGE Response PGE agrees with the Impe11 recommendation. Bechtel hr.s incorporated this recommendation in the analysis and the results and conclusions did not change.
Recommendation / Finding 6 The thermal stratification calculation is based on a unit " load" of a 100*F temperature difference across the piping cross-section. At this. temperature difference, a 5.4 kip force acts at SR-8. Due to i the large difference between the main feedwater and auxiliary feed-water temperatures,. variations larger than 100*F are possible, however. Under this situation, relatively high loads at SR-8 could occur. It would then not take a significant hydro-dynamic type event to overload the support.
Examine this issue in more detail over the next fuel cycle to deter-mine the range of temperature differences the piping actually exper-ionces under Plant operation. Depending on the magnitude of the resulting loads, decide whether this condition should be incorporated into the design basis.
PGE Response As noted in our response to Recommendation 4.c above, the impact of thermal stratification on SR-4 and SR-8 loading will be evaluated as operating data are collected. In the meantime, system operation is justified based on PG8 and Bechtel calculations Which show signifi-cant margin in the restraint designs capable of accommodating maximum l credible thermal stratification loads.
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