ML20023B370
| ML20023B370 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/18/1983 |
| From: | Mattson R, Thompson H Office of Nuclear Reactor Regulation |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20023B372 | List: |
| References | |
| NUDOCS 8305030587 | |
| Download: ML20023B370 (15) | |
Text
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UNITED STATES y
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NUCLEAR MEGULATORY COMMISSION f..
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APR 101cS3 N IORANDUM FOR:
Carrell Eisenhut, Director Division of Licensing FRCM:
Roger J. !!attson, Director Division of System Integration Hugh L. Thompson, Jr., Director Division of Human Factors Safety
SUBJECT:
GENERIC FOLLOWUP EVALUATION TO BOARD NOTIFICATION BN-83-21 FOR B&W PLANTS P,eference:
1.
Memorandum,11attson and Thomoson to Eisenhut,
" Board Notification", dated February 18, 1983.
2.
Memorandum, Mattson and Thompson to Eisenhut,
" Followup Evaluation to Board Notification SN-83-21 for TMI-1", dated March 11, 1983.
Ourmemorandum of February 18, 1983 recuested that you notify licensing boards associated with reactors designed by Babcock and Wilcox of new infomation involving auxiliary feedwater effectiveness. Our memorandum of March 11, 1983 provided our evaluation of this matter for TMI-1 which concluded that the infomation does not adversely affect our present conclusions regarding the ability of TiII-1 to achieve and maintain decay heat removal by natural circulation through the steam generators under transient and accident conditions.
We have now completed a generic evaluation for all S&W designed plants and have reached the same conclusion as for TMI-1.
The generic evaluation is enclosed.
We request that the generic evaluation be provided to the remaining licensina boards Yb wW
.4 (
-s
'oger J. !1attson, Director Divisi n of ~yste s Integra-' n
$P Huch
. Thompson, irector Di ion of Huma Fa. ors Safety Er. closure:
As stated c::
see next page i
1
O. Eisenhut -
cc:
W. Dircks, E00 H. Sullivan V. Stello G. Lauben H. Denton W. Jensen T. Speis R. Minogue, RES
0 a-ENCLOSURE Backaround On February 18, 1983, the staff issued Board Notification BN 83-21.
This board notification identified information which had recently come to the attention of the staff that was potentially significant with rescect to achieving and maintaining natural circulation in B&W-designed reactors.
The genesis of the hencerns was the staff review o' the GPU-B&W 1awsuit trial transcript.
In this transcript, testimony by two individuals raised questions on two technical areas concerning natural ci rcul ation.
The Issues The details of the technical issues identified were discussed in SN S3-21 and are repeated here.
During the trial, testimony by Dr. R.' Lahey of Rensselaer Polytechnic Institute (RPI) and Dr. G. Wallis of Dartmouth College identified two concerns.
These are (1) the adequacy of emergency operating orocedures to assure that a sufficient condensing surface woulc be established in the stean generators under all design basis conditions for which decay heat removal by the steam generators was recuired ano (2) the ability to I
-c-establish an effective condensing surface at the elevation of the auxiliary feedwater sparger ring in light of new data which shows limited penetration into the tube bundle of feecwater entering the stean generator from the emergency feedwater soarger ring.
The first concern was raised by Dr. Lahey.
It deals with procedures and relates to whether or not the operators have sufficient instructions and training to assure that they will raise the secondary level of the steam generator to 95 percent of the operating level under all conditions necessary to assure natural circulation.
Following the TMI-2 accident, it was learned that the then current procedures instructed operators to raise the secondary level to 50 percent of the operating range.
Under certain circumstances, it was possible to postulate that natural-circulation would not be reestablished with the secondary level at 50 percent.
Subsequently, it was determined that raising the level to 95 perce.nt of the operating range would assure natural circulation if the RCS was saturated.
However, because of overcooling considerations, it is not desirable to raise the level to 95 percent for all cases of loss of forced circulation.
Thus, specific plant circumstances dictate the aporopriate steam generator level and the manner to achieve this level.
The operating procedures and training to describe the correct actions I
are, therefore,'important to the issue.
1 A discussion of this issue was presented in NUREG-0565 (" Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior in Babcock and Wilcox Designed 177-FA Operating Plants," dated January 1980) and was provided in Attachment 2 to BN-83-21.
A cooy of the relevant i
m sections of Dr. Lahey's testimony was orovided as Attachment 3 to EN-83-21.
Dr. Wallis' testimony was provided to the Acceal Board and carties to the recoened Ti1I-1 restart proceding on March 8, 1983.
The second concern was raised by Dr. Wallis.
It involves recent test data frcm the Alliance Research Center which shows that auxiliary feedwater entering frem the soarger ring does not penetrate uniformly into the steam generator' tube bundle but only contacts a small percentage of the tubes.
This has the effect of lowering the elevation of the effective condensing surface in the steam generator.
Previous analysis mode ~1s assume good cenetration of auxiliary feedwater spray into the tube bundle but recent B&W models account for the new data.
The issues raised by Dr. Lahey and Dr. Wallis are of concern only for B&W design plants with a lowered loop configuration.
This arrangement is shown in Figure 1 and is applicable to ANO Unit 1, Oconee Units 1, 2 and 3; Crystal River Unit 3, TMI Units 1 and 2; Rancho Seco and Midland Units 1 and 2.
Other plants designed by B&W have a raised 1000 configuration (see figure 2).
This arrangement is utilized for Davis Besse Unit 1; Bellefonte Units 1 and 2 and WNP Unit 1.
The 3ellefonte and WNP Plants are further identified as 205 fuel assembly plants whereas all other S&W plants have 177 fuel assemblies.
For plants with the raised loop configuration the entire secondary level would provide a condensing surface above the core and operator action to raise the steam generator leyel from 50% on the operatinc range to 95",
would not be reouired.
The effectiveness of the condensing sur# ace at l
1 the elevation of where auxiliary feedwater enters the steam cenerator wculd not be significant for plants the raised loop design since any auxiliary feedwater which did not provide heat transfer by boiling above the secondary level would act to increase the depth and therefore the effectiveness of the secondary level as a condensing surface.
The Bellefonte and WNP plants have never taken credit for auxiliary feedwater effectiveness above the secondary level since AFW enters the bottom of the steam generators for these plants.
The concerns raised by Dr. Lahey and Dr. Wallis are applicable only to plants with the lowered loop design and these plants are addressed genericly in the following i
NRC staff evaluation.
Staff Evaluation 1.
EFW Soray Effectiveness On February 23, 1983, the B&W Owners Regulatory Response Group (RRG) met with the staff to present information on the above two technical issues.
Subsequent to this meeting, the owners group submitted a technical report, " Evaluation of SSLOCA Operating Procedures and Effectiveness of Emergency Feedwater Spray for B&W designed Operating NSSS," (Reference 1) which documented the j
information presented at the February 23, 1983 meeting.
The staff has reviewed this report and our evaluation follows:
?
d
A.
Effectiveness of Emercency Feedwater Sorav In the Once-Through-Stean-Generator (OTSG) design for the l'owered loop plants, emergency feedwater enters the steam generator throuch seven nczzles located circumferentially around the OTSG shell.and at an elevation just above the top tube support plate.
This is shown in Figures 3 and 4, taken from the B&W report.
Also shown on Figure 3 is the operating range for feedwater level (item 24).
Analysis models used by B&W and the staff have assumed that emergency feedwater injected at the sparger elevation would be uniformly distributed within the tube bundle and provide effective heat transfer to all tubes within-ths bundle.
Data obtained from testing performed by B&W 'at its Alliance Research Center (ARC) shows however, that the emergency feedwater spray does not effectively penetrate the tube bundle providing uniform wetting and unitann heat transfer throughout the bundle.
Rather, the emergency feedwater only contacts those steam generator tubes in the immediate vicinity of the in.iection nozzle.
The emergency feedwater would then pool. on the tube support plate and spread out, draining down the flow holes where the steam generator tubes penetrate the tube support plate.
As can be seen from Figure 3, at least 6 tube support plates exist between the injection location and the top of the operating range.
In Figure 5, B&W shows the emergency feedwater axial wetting profile measured in Oconee 1.
This shows that as the emergency feedwater drains down the tube bundle, the tube support plates tend to scread the flow tewards the center of the tube bundle.
The wetting profile could be envisioned as an " inverted cone."
The imoact of this incomolete wetting of
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-c-the tubes is that the heat transfer rates attributed to soray cooling above the secondary side pool level can be reduced from what they were originally, based on the 100 percent wetting assumotion.
In the B&W report, data and data correlations are presented which allow the percent of tube area wetted above the secondary side pool level to be calculated as a function of EFW f1cw.
In the following section, the reliance on EFW spray effectiveness will be discussed.
Subsequent to the TMI-2 accident, the staff investigated the possible causes fo natural circulation not being established in
~
TMI-2 once the reactor coolant pumps were shut off.
In reference 2, it was postulated that natural circulation did not commence following RCP trip because the secondary side steam generator level
.as not high enough to establish a condensing surface which would w
- allcw natural circulation flow over the pump entrance.
In references 3 and 4, the staff further explained and documented this concern and concluded that in order to assure natural circulation, (in particular boiler-condenser), the secondary side steam l
generator level must be raised to above the elevation of the i
l reactor coolant pump.
B&W proposed raising the steam generator secondary level to 95 percent of the operating range in reference
- 5. By raising the level to 95 percent of the operating range, a l
condensing surface that is above the elevation of the pump and is sufficient to remove all decay heat in assured.
By establishing a condensing surface above the puno elevation, condensation of L
o 1
primary steam and the buildup of a condensate level on the crimary side of the tubes sufficient to cremote flow over the Duma inlet is also assured.
In addition, because the liouid levels in the core, downcomer, and steam generator are equalized due to the vessel vent valves, the 95 percent secondary level assured that a condensing surface will be established before core uncovery could occur.
In other words, to assure that natural crculation (boiler-condenser) would commence and reartablish decay heat removal before core uncovery could occur, no credit for EFW spray effectiveness needs to be taken.
The EFW spray could be postulated not to provide any heat transfer, as long as it contributes to the secondary cool inventory.
Thus, while the effect of the reduced heat transfer due to the reduced penetration of EFW spray during a SBLOCA would be to change the degree of initial overcooling and thus the initial system pressure response, the overall conclusions regarding core cooling would not change.
Inherent in this conclusion however, is the assumption that the pool level on the secondary side of the steam generator is raised to 95 percent on the operating range in e titcely manner following a SBLOCA.
Presently, the EFW is autors43ct :y controlled to establish the level at 50 percent
- c.arating range. Operator action is required to raise the level from the 50 to 95 percent level. To estimate the time available for the operator to initate actions to raise the EFW level to 95 percent, analysis by 3&W in reference 5 shows that for the largest break size for which steam generator heat renoval would be recuired,
(.01 sq ft.), boiler condenser heat transfer was calculated to commence after 25 minutes.
tioreover, at the time boiler-conde'nser i
ccmmencec, S&W analyses indicated 105,500 lb. would still remain above the core.
If this amount of coolant were assumed to exit the primary system via the break (.01 sq ft.) as saturated liquid, it would still require at least an additional 35 minutes before core uncovery.
This is a total of at least 60 minutes available to establish a condensing surface for the limiting break.
If the level must be raised from the 50 percent to the 95 ogreent level and it takes approximately 1 minute to raise the level one foot,
then we estimated'it would take approximately 12 minutes to raise the level to 95 percen't.
Therefore, there is estimated to be in excess of 13 minutes available for the operator to recognize the event and initiate filling of the secondary side of the steam generator with EFW to achieve the 95 percent level for the most limiting small break to establish boiler-condenser in the time period assumed in the B&W analysis.
A still longer time would be available before core uncovery could occur.
l
- This estimate is used by B&W and has been confirmed by the staff.
e
.g.
Tigure a, which is recroduced from Figure 3-3 of the B&W submittal, orovices the results of mass and energy balances which demonstrate that even with reduced EFW penetration, the EFW scray is still sufficient to remove decay heat.
The solid curve, labeled EFW scray, recresents the coint at which the overall heat transfer rates from EFW scray, combined with the crimary to secondary temperature difference at the indicated RC pressure, can remove all of the decay heat at the indicated time.
This curve is basically the energy balance requirement.
The dashed lines represent the points at which the HPI flow can match the core
~
boiloff at the prevailing pressure.
The significance of this curve is that orior to any core uncovery, the break must uncover, and discharge steam.
The source of the steam is the core boiloff; therefore the dashed line represents the point at which HPI flow
.will match break flow (the mass balance).
The intersection of the two curves is the point at which both all decay heat can be removed by EFW spray and HPI can fully make up all mass loss through the break.
These occur for all times beyond about 1000 seconds for the 100% HPI case and beyond 3000 seconds for the 70% HPI case.* It is therefcre only necessary to show that core
- The 70% HPI case refers to 1 HPI train with a 30% reduction assurec to result from scillage of HPI out the break.
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~.
q uncovery does not occur for any small breaks less than.02 square
- t (i.e., those breaks which recuire the steam generators for heat removal).
As previously shown, core uncovery cannot occur prior to at least 60 minutes, or 3600 seconds, which is in considerable excess of seconds as indicated above.
2.
Emercency Procedure and Ooerator Trainina Adecuacy The staff's board notification SN-83-21, dated February 18, 1983, stated the importance of operating procedures and operator training in assuring that a sufficient condensing surface is established in the steam generators under all design basis conditions. The concern raised by Dr. Lahey is whether or not coerators have sufficient instructions 'and training to assure that the secondary level of the steam generators is increased to 95 percent of the operating range in a timely manner for all conditions necessary to assure natural circulation.
Subsecuent to board notification of this issue, the staff met with
(
members of the B&W Owner's Regulatory Response Group on February 23, 1983, to discuss design features, the emergency coerating procedures and operator training.
All owners of B&W design coerating reactors were represented.
The staff was informed by l
representatives of each plant that instructions necessary to bring i
the stean generator levels to 95 percent of the operating range were included in procedures and that the totality of training and procedures were sufficient to ensure that necessary coerations l
would be performed.
fioreover, the staff recuested and obtained emergency operating procedures from Oconee, Arkansas Muclear One, l
P.ancho Seco and Crystal P.iver to ascertain what instructions are
' provided relative to maintaining steam generators levels.
The Davis-Besse plant procedures were not requested since the raised loop plants do not have the problem.
The review included procedures for responding to loss-of-coolant and natural circulation.
The procedure organization and formats were different among the plants and represented the individual owners' procedural philosophy.
Howevdr, our review concluded that the procedures provide specific instructions to increase steam generator levels to 95 percent of the operating range when the reactor coolant pumps are tripped following a loss-of-coolant.
It should be. noted that Arkansas Power & Light Company's event-based emergency procedures have been replaced with emergency operating procedures based onr the B&W Owners' Group Abnormal Transient Operating Guidelines These new procedures do not require the determination that a
. lass-of-coolant be identified to enter the procedure.
Instead the new emergency procedure is entered upon receiot of a scrant.
PTant symptoms direct the operator to the appropriate procedural steps for coping with the threat to loss of a safety function.
The small break loss-of-coolant symotoms direct the operator to actions for loss of subcooling which recuires as the first action reactor l
coolant pung trip.
The operator is instructed to raise steam generator levels. to 95 percent on the operating range as the second action.
In addition, the composite of each licensee's procedures contained instructions and guidance to assure the steam generator levels will
l
.c-1 i
be raised to 95 percent on the operating range for conditions of no forced cooling flow and existence of indications of sucerheat in the RCS.
Discussions with licensee representatives present at the February 18, 1983 meeting indicated that operators have been trained in the use of these procedures.
The staff did not review the training programs in detail; however, licensee representatives stated that operator training on use of emergency operating procedures included instructions to remain in the aopropriate emercency procedure in use unless the procedure specifically instructs the user to exit it or unless there is another valid reason to exit it.
This gives further confidence that the operator will not exit a procedure prior to establishing 95 percent level in the steam generators, if recuired.
Based on the staff review of procedures and discussions with representatives of B&W designed operating plants regarding operator training, the staff concludes that there is reasonable assurance that the operator will increase steam generator level to 95 percent of the coerating range under the conditions for which it is necessary to establish natural circulation.
Conclusions The staf# has evaluated information provided by the B&W owners group regarding the effect of reduced EFW cenetration on decay heat removal caoability for the reactors with lowered looos.
Based on
13 this evaluation, we have concluded that-for the design basis SBLOCA scenarios, EFW spray cooling need not be relied uoan to assure adequate decay heat removal, and decay heat renoval solely by, primary steam condensation in the region of the secondary side pool, which is at an elevation of 95 percent of the operating range, is adequate.
This conclusion assumes correct coerator action within approximately 10 to 15 minutes to initate raising the steam generator lev'el to the 95 percent level.
For scenarios not normally considered in the design basis, including delayed EFW, credit for EFW spray cooling is relied upon to assure core cooling.
Analyses by B&W show that after accounting for the reduced EFW penetration into the steam generator tube bundle, the EFW spray cooling will still provide effective decay heat removal.
Based on staff review of the procedures for loss-of-coolant and
. natural circulation and discussions with representatives of the licensees for B&W designed plants with lowered loops regarding operating training, the staff concludes that there is reasonable assurance that the operators will increase steam generator levels to 95 percent of the operating range under the conditions for which it is necessary to establish natural circulation.
.c REFERENCES 1.
" Evaluation of SBLOCA Operating Procedures and Effectiveness of Emergency Feedwater Spray for B&W-Designed Operating NSSS," B&W 00C. ID. 77-1141270-00 dated February,1983 2.
Memorandum, B. W. Sheron to Z. R. Rosztoczy " Pool Boiling
-Condensation Natural Circulation In TMI-2," dated July 23, 1979.
3.
" Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior in Babcock and 'lilcox Designed 177-FA Operating Plants" NUREG-0565 dated January,1980.
4 B. W. Sheron, " Generic Assessment of Delayed ' Reactor Coolant Pump
. Trio during Small Break Loss-of-Coolant Accidents in Pressurized Water Reactors" NUREG-0623 dated November,1979 (Appendix A).
5.
Letter from J. H. Taylor (BaW) to R. J. Mattson (NRC) " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in l
the 177-FA Plant," Volumes I and II, dated May 7, 1979.
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