ML19246A363
| ML19246A363 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/24/1979 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Case E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19246A360 | List: |
| References | |
| ACRS-SM-0102, ACRS-SM-102, NUDOCS 7906180631 | |
| Download: ML19246A363 (13) | |
Text
y s
.:' ' e f' t NUCLEAn nEGULATOnY COMMisslOrJ W A5mNGTON. D. c, Oc%5 f, &fl s
sr
%.". '
- e APR >.13)g L
E. G. Case, Deputy Director, Office of fluclear Reactor MEMORA*iDUM FOR:
Regulation D. F. Ross, Jr., Deputy Director, Division of Project FROM:
Manacownt. flRR
SUMMARY
OF MEETItiG WITH B&W REGARDIriG f1ATURAL CIR
SUBJECT:
C0i4SIDERATI0tlS the tiRC staff met with representatives of Babcock &
On April 18, 1979 Wilcox (S&W) Corporacion in Bethesda, Maryland to discuss severalA considerations related to natural circulation in B&W reactors. A list representative of Duke Power Company was also in attendance.
of attendees is attached (Enclosure 1).
The The meeting opened with a pr sentation of the proposed agenda.
following four (4) general s' eas were to be discussed:
?
1.
Concerns raised bylthe ACRS in a recen+. letter.
A recent B&W precaution recarding subcooling, RCP operation and 2.
0TSG 1evel while.ttempting to establish natural circulation in the RCS.
A report written'by C. Michelson, a consultant to the ACRS, 3.
regarding potential difficulties in the removal of core decay heat for certain small break LOCAs in the 205 class B&W reactors.
I Staff concerns related to the ICS and how it effects natural 4.
circulation in certain scenarios (loss of all RCPs, loss of off-site power).
I' l.
ACRS Concerns A.
Greater Understanding of flatural Circulation B&W explained the basic 'rinciples of natural circulation on the There are tcree (3) basic. criteria which must be met B&W system.
for natural circulation to be establish,ed in the RCS.
1.
There rust be an elevation head between the thermal centers of thy system.
/
The.RCS loops must be water solid without steam (the hot leg 2.
ter'.ierature, T, must be less than or equal to the saturation g
temperature for the pressure in the hot leg).
7906180631 p
228 217
\\
-~ E. G. Case 3.
There must be no interruption of flow by bubble of non-condensib'e gas (H ).
The partial pressure of hydrogen must 2
be below the pressure in the RCS.
B&W stated that as long as these three criteria are met, then natural cir ulation will occur, and the relative elevation of the pressurizer is not important. Using a drawing showing the elevations of the RCS components (applicable to all BSW plants except Davis Besse-1 (raised loop design), B&W showed that the Auxiliary Feedwater inlet location Al so, S&W (high on the OTSG) helps promote natural circulation.
stated that the amount of natural circulation flow would depend on the clevation difference and the system AT (which would deH on the heat input and removal).
l The staff asked B&W to plot the temperature as a function of tube length along the OTSG, and B&W,shcwed that under a natural circulation scheme, there would be a marked drop in primary temperature corresponding to the secondary fluid slocation, (water, steam interface).
5 B&W plants use an ICS'.>hich autcmatically raises OTSG level, to SC%
using the Auxiliary Feedwater system if all four (4) RCPs are tripped.
During normal operations, This feature is prese, t on all B&W plants.
when powar is bdow 15%, OTSG level is maintained two feet above the tube sheet ( ~
j6") for all plants.
The OTSG "cperating"' level is about 1/2 of the total tube bundle length, and a 50% level corrCacnds to roughly 1/4 of the total tube length.
In the " operating" level range, 05 corresponds to about 96" above the tube sheet, and 100T, corresponds to about 388" above the tube sheet.
The indicated level is temperature comoensated and represents a cross-section of varying density across the OTSG.
The staff noted that with respect to natural circulation criteria 3, hydroge' since the solubility of H decreases could ccme out if Jg approached TSAT 2
- markedly as tempeEature approaches the saturation value.
B&W noted that criteria #2 and #3 together should limit the amount of H coming out of solution. The staff agreed with this, but noted 2that during a normal transient, structural ' components may keep local' temperature up,.and could the temperature approach the sat cation value (as pressure decreases) thereby approaching a situatior where could ccme,out of solution?
H2 B&W didn't bcTieve enough H could come out of solution to block the 2
candy cane.. but they would check with experts (chemists) in Lynchburg.
228.218 L
\\
w
E. G. case 3
The staff asked B&W if during the transients experienced, the structural components (pipes, upper vessel internals, etc.) kept sy,; tem temperatures up.
B&W stated that the transients would have to be looked at individually, m mentarily, however the PPI is but temperatures may approach TSAT designed to replace lost volume due to flashing.
Also, the HPI response time is adequate to prevent significant voiding (by flashing).
B&W has conducted tests to determine the amount of natural circulation.
The tests are normally done during startup testing from an initial pwer level of about 20-253. The reactor is scrammed, the RCPs are tripped, the emergency diesel generator comes on, the steam and motor driven AFW pumps start, the ICS raises OTSG level to the 50%
value, and the plant is verified to be operating on natural circulation, without any operator action. The operator only has to monitor the systemstoensurethgirproperoperation. The RCS temperaturg levels out at about 550-560 F, the aT across the core is about 30-40 F, the pressurizer level (L ) steadies out at the normal value, and p
system pressure is 2000-2100 psig.
These tests have been conducted at Davis-Besse and Ocenee.
Al so,
Arkansas-1 suffercd a loss of offsite power (LCOP) from 100%, on 7/25/75 and natural circulation was established, without any operator action.
TMI-2 also had, two (2) unscheduled events in their startup testing program (LOOP te?t) which resulted in natural circulation.
The staff requested as much detail and description as possible on all the natural circulation tests and events.
The staff asked B&W to explain how system pressure is controlled if the LOOP results in a loss of pressurizer heaters.
B&W stated that the system pressure is mainly determined by the bulk system temperature which is a function of the energy input / removal.
If temps are constant, the pressurizer level is constant, and pressure is steady.
- Also, the makeup system, which shouldn't be lost in the LOOP, acts to maintain pressurizer level, which helps maintain pressure.
(B&W noted that during the ANO-1 event, the makeup pump was lost for an unknown reason.)
The staff asked B&W d at analysis had been conducted regarding the system response to natural circulation, and.had the various tests and unplanned events been compared to these analyses?
B&W stated that no fomal report regarding natural circulation had ever been generc:cd or submitted to the staf f, but significant in-house knowledge based on analcg and digital computer studies exists.
The staff asked B&W if these analyses included off-nomal situations, FMEAs, or sensitivity studies.
B&W noted that some sensitivity studies i
s:
" ~~
228 219
(
A
, E. G. Case had been done ( Auxiliary Feedwater initiation times, RCP inertial effects, and possibly initial pressurizer level effects), but no extensive sensitivity studies had been done.
For exarple, during the course of analyzing a loss of feedwater transient (the normal safety analysis) various single failures are considered, and the worst is assumed for the analysis.
The transient is analyzed for OtiB con-siderations, but natural circulation is not analyzed in this analysis.
Recarding the sensitivity studies of natural circulation, the staff asked if the PORV is manually openeo to control pressure during a LOFW? B&W stated that during the ANO-l event, the PORy was not opened, and in general, the PORV should not be challenged during a LOOP (or LOFW).
The staff noted that the plant response may be quite dependent on initial parameters, and asked if the B&W transient code, POWERTRAlfi, has been shecked against the Ari0-1 event.
B&W stated this had not yet been donc since they didn't yet have all the data.
The staff asked if any B&W reactor operator has any instructions on what to do if something failed during an attempt to establish natural circulation?
?
B&W noted that the gerator requalification program covers many instrument failures', but they don't knowt the details of the requalification program (training expert couldn't attend the meeting).
The plant operators have general instructions in the procedures to ensure that the automatic actions have occurred as designed, and this implies that if the automatic action has not occurred, to take corrective action. However, there are generally not statements in the procedures such as
"...if the AFW pump has not started within minutes, do..."
The staff asked B&W if there is an iteration between plant operators and designers to ensure the procedures are adequate with respect to the design.
Duke Power Company representative stated that B&W design engineers reviewed all their procedures and made meaningful coments, and it is the policy of Duke to ask all equipment vendors to review procedures (generated by Duke) for the operation of that equipment.
B.
Fore Analyses and Experiments The staff noted that this agenda item has been discussed under Item A, and that we will probably request more detail'd analyses. This item will be discussed at the end of the meeting.
i j
228 220
~
C s.
E. G. Case C.
Better Procedures The staff noted that when at the TMI-2 site. many procedures for the situations occurring there had to be developed on the spot.
And the staff wonders if other plants must also develop these procedures in a like manner.
B&W recognizes tr ' concern regarding the detail to which the "what if" type analysis has oeen incorporated into the plant procedures.
The plant procedures are specific to each plant, and the indivicual most knowledgeable in this area could not attend, (training coordinator).
The staff noted that this item may be addressed in a bulletin.
D.
Information to Tell the Operat,r if Natural Circulation is Working The staff noted that this item is included in item C, above, and may be included in a bulletin. 83W stated that they do not have a document which could be used to tell the operator if natural circulation has been successfully achieved.
Duke Power Company stated they had looked at natural circulation with respect to plant security scenarios, and they have personnel who have thought through the details of natural circulation in the Oconee units.
E.
Role of Pressurizer Heaters (and pressurizer spray)
B&W stated that the pressurizer heaters are not eseential within the first 10-15 minutes following a LOOP, and they can be transferred manually to at energized power supply thereafter. They have calculated that there are about 1.5 MW of ambient heat loss at 532 F, and in an hour, pressure would decay about 40 psi. The staff noted tha'. TMI-2 estimated the heat losses from the pressurizer alone would be abeut 20 kw.
If B&W deemed it was necessary to have the pressurizer heaters powered by a vi al power supply, they would issue a " site instruction".
F.
AwarenessohImpendingSaturaticn To warn the operator that system temperatures are approaching the saturation temperatures for the system pressure, Duke Power utilizes the plant process ccmputer and a curve (Pressure vs. Temp with 50 F subcooling line). An explicit warning or alarm is not presently used.
85W is considering the need for an extra alarm ur warning.
Now, they conclude that the operator should definitely be aware of the status of piant subcooling, but they do not yet know if extra instrumentation is necessary.
a 228 221 e
N r---
s
.- E. G. Case The staff expressed concern over the validity of temperature readings That is, tie due to lag time in a natural circulation mode.
temperature being sensed is physically far frcn the core temperatures.
the 70 F subccoling was lost over 0
During the Davis Besse-1 event, a 5-minute span without operator intersention.
G.
Role of Exit Thermocouples The staff asked B&W stated that all B&W plants have core exit T/Cs.w as they relate to natural circulatica.
B&W responded that the core exit T/Cs may be important, but they're not yet ready to make a reccmaendaticn regarding their use in dec For example, if during natural circulation 80; of the Whig, reed,butnotedthattheT/Csarecertainlyasignthatnatural T
The staff also circulation may not be adequately established.
noted that this item may be further addressed in a bulletin.
i H.
Simulator Training The staff expressed #'s desire to cbserve the training taking place B&W simulator, and the degree to which "what if" type scenarios are investigated during ~ tural circulation simulation on the machine.
However, the staff dues not want to interfere with operator tra nor impact on any TML-2 simulation.
observe the simulator.)
.t B&W said they would investigate the availability of the simulator for observation.
J P
2.
B&W Precaution B&W explained that the ;rigin of the B&W precaution was from re f.
from Dr. Etherington, Uf the ACRS.
l tion.
tripping the RCPs because he expected the plant to go into natural c If this is true, tren how might he have known beforehand that natural circulat on was not achiev 2hle?
i Based on these comments, B&W decided it prudent to issue guidan operators ecgarding tripping of the RCPs.Fer concurrence, a and the nuclear ser;/ ices organization.
The staf f 1sked Cu'Ie Power what their'acticns would be if they r Duke stated that they would guidance like the ene proposed by B&W.
l imediately brim their operattes, since the guidancc is only procedura and not a hardware change (other vise their plant safety ccmmitte be involved).
228 222 r..
gas.P r=w
E. G. Case j
The staff asked B&W to t plain what happened at TMI-2 with respect to OTSG level.
Apparently, the ICS was controlling A OTSG level at its startup level Then the ICS
( t 26") up until the time the RCPS were all lost.
raised A OTSG level to the 503 value, but B OTSG was isolated (manually beforehand) so its level was later manually raised.
noted that the B&W guidance says to ignore what the ICS will The staff do to OTSG level, and take manual control before tripping RCPs.
B&W agreed and pointed out that the bulletin said to cbserve all temperature and pressure limits while raising CTSG level, so no limits should be violated. This action just is earlier than the ICS automatic action (the ICS would raise OTSG level at about 15"/ min).
B&W stated that both guidances (paragrapn 1 and 2) are prudent actions, but are not necessary actions since the ICS does the same thing.
The staff asked if it would also be prudent to initiate HPI along with raising OTSG level prior to RCP stopping to increase the subcooling.
B&W said they would consider this.
B.
Role of ICS - Does it fleed Changing?
B&W stated that the ICS dccs not require hardware changes, and that the guidance being suggested is best carried out with procedural modifications.
C.
Exis+ing Guidelines or Criteria Provided to Custcmer B&W is resear:hing the guidelines associated with natural circulation and they'll.nform us.
3.
Michelson Recort It was B&L"s initial assessment that the phenomena discussed in the Michelson report have no real bearing on the B&W reactors, and that the B&W reactors for the break size of interest do not suffer a loss of natural circulation Also, for the (intermitter.t or sustained) as the report suggests is possible.
range of break sizes discussed in the report, the core is not uncovered.
The effect of non-condensible gas:
--iing out 'of solution is not appreciable, and would not effect natural circula; cn.
The only aspect of concern associated with the etack open P')RV is theNa misleading L indication.
P When asked by the staf f if it is good or bad to shut would not t'e lost.
the PORV MOV if the operator notices the stuck open PORV, B&W stated that remains on, it wouldn't matter since natural circulation as long as HP:
wouldn't be lost.
228 2E t-s
\\
E. G. Case.
Another PUR vendor stated that for a 21/2" steam space break, if AFW is unavailable, and HPSI is available, the core remains covered, but if the HPSI is not available (and the AFW unavailable) the core uncovers.
B&W stated this was probably true in their reactors also.
They note that a key piece of infomation regarding the TMI-2 event is the times when the HPI was unavailable.
B&W stated that their analyses (small break LOCA) assume thu availability of Auxiliary FW at about 40 sec, and the OTSG level goes to that set by the ICS (for loss of RCPs), abcut 17 ft.
B&W noted, in response to several questions regarding the ICS, that there are three (3) pcwer supplies to the ICS: (1) batte:y pack, (2) offsite power and (3) energency diesel generator.
1 The staff asked B&W to discuss the sensitivity of AFW initiation time on the result s of a PORV stuck open LOCA.
B&W said that even without AFW, and the OlSG boiling dry, the core goes into pool boiling, and, as long as HPI is available, the core remains covered.
Voids would occur in the system and the safety valves.wnld cpen and pass water.
Y The staff su riarized the B&W response:
although not specifically analyzed, 2
if there viere a loss of fee water w/a small LCCA (21/2-3" or ~ 0.04 ft ),
if both HPI ptmps are avaiable, the results are satisfactory, (based on B&Ws judgment).
B&W agreet with this summary.
o In response to a previous concern on the H coming out of solution, B&W reported that the normal hydrogen ccncentration in the system is about 40 cc/kg, as set by the nakeup tank cover gas pressura.
This corresponds to about 440 standard ft3 of hydrogen, or abcut 20 ft at 300 psia if 3
all evolved.
At a ccncentration of 40 cc/kg, at a temperature of 410 F at 300 psia, significant H,,
wouldbereleasedfromthesgstem.
If the concentration were 68 cc/kg, then at a temperature of 405 F at 300 osia, significant H would evolve.
2 4.
Staff Concerns I
l-5 The staff noted that many of the questions were already answered in the dis-cussion of the preceding items, and that only a few remain.
6.
Thermal Shock Considerations The staff is concerned for the potential for themal acck of the reactor vessel 'em relatively cold UPI water entering the dcwnccmer, during a loss of naturai irculation.
What would be the minimum RCS temperature with the pressure being maintained by the RCS code sa fe ties.
22 224 s
s
___ E. G. Case B&W said they would look at giving the operator more instructions regarding the closing of the PORV MOV since the thermal shock problems depend on the tipi flow into the system (as well as system
(
pressure).
A 1-hour break was taken for the staff to discuss t ie necessity (and co of a bulletin to B&W reactors, and other information/ analyses required from B&W.
The staff concluded that, based on the information presented at the meet a bulletin to all B&W reactors would be issued oy cob Thursday, 4/19/79.
The bulletin would have four (4) ingredients:
Operating plants must reduce the likelihood of opening the PORY 1.
in the event of an anticipated transient.
Operating plants should incorporate a scram feature sensing a loss (The scram of heat sink (loss of load or loss of feedwater).
2.
may be manually performd.)
General guidance to operators regarding the necessity of determining the degree of subcooling before attempting to initiate natural 3.
circulation.
fr General guidance to - erators regarding the OTSG level during an 4.
attempt to establish natural circalation.
The staff noted that wit $ respect to item 1, reducing the overpressure scram setpoint to below the PORV setpoint may result in the necessityB&W for steady state operation at reduced pressure.
of the study the possibility with respect to the continued validity ECCS analyses.
B&W agreed to look at this aspect, and suggested the possibility of Such a scram signal, however, an automatic scram on loss of feedwater.
the may originate frca "un' qualified" instruments which might violate B&W GDC (regarding mixture of safety /ncn-safety grade systems).
said they would contact the staff in the morning of 4/19/79 to discuss the most preferable means of achieving the goals of 1
/
s
f
[
.f.
228 225 L_
\\
\\
E. G. Case :
TIME
_ ITEM
(
A.
Perform calculatiuns, worst-case break without AFW for 30 min.
2-3 days B.
Document natural circulation tests conducted at Davis Besse & Oconee 2 1/2 weeks C.
Document all occurrences of natural circulation which happened inadvertently; include a description of unexpected behavior 2 1/2 weeks D.
Document natural circulation analytical methods 4 weeks y
E.
Summarize and document sensitivity in key parameters (defin' tion and agreement with staff in tvo wee;.s regarding scope) 8 weeks F.
Deleted
[
i G.
Define and focument thermal shock criteria for operr+
at 10 ' ~ unerati re with HPI punips renn oig and no C + ural..rtf -^ ion 2 weeks The staff ag-eed wits
...A item's and schedule, and requested that items A-D p to f,'. S t:%on, and tnct items E and G not t ' started until the re; ort. :ing ca.erated by R. Tedesco is ccmplete, around May 1, 1979 si - nis report will define the scope of work desired.
t-Also, the stafi rc 4.nts an additional item, called H, to be an assess-ment of the safety concerns raised in the report of Dr. Michelson, and that this assessment be submitted within 21/2 wecks.
B&W agreed to our requests.
t The staff met with S&W on 4/13 to discuss alternative means of reducing PORV actuations in the event of the type of transients that have been exper.enced at B&W designed plants.
B&W considered four alternative means of achieving the above objective.
1 1.
Restrict Initial, (operating) Pcwer o
Experience shows PORV would lif t even at low power level ( 4 9%)
and thus this liternative would be ineffective.
f 2.
Lowering High' Pressure Reactor Trip Scipoint In order tc' minimize or eliminate PORV opening following feedwater transients, the reactor trip setpoirt would have to be lcwered to 228 220 N
w
\\
a _. E. G. Case 2215 psig which wot.1d result in only 60 psig margin in pressure between the nominal reactor operating pressure and the high pressure Bec.ause of uncertainties in these parameters; increased scram setpoint.
spurious reactor scrams would be expected.
3.
Plant Operation with Reduced Operating Pressure To eliminate lifting the PORV during transients, the normal operating pressure would have to be reduced to about 1900 psig with a comparable reduction in high reactor pressure trip to 2100 psig. A reduction in system operating temperatures would also be required to preserve margins to departure from nucleate boiling (DNB) in the core. These changes would necessitate an extensive reanalysis of all transients and accidents.
/
Reduction of High Pressure Trip Setpoint and Increase of Pilot-4.
Operated Electromatic Pressurizcr Relief Valve Setpoint By lowering the reactor' coolant system high pressure trip setpoint of the pilot-operated relief valve, and raising the setpoi.
B&W stated that it is ;;ossible to eliminate the lifting of theThey PORV following transients which have occurred on B&W plants.
had performed analyse. to support this conclusion.
I The pressure trip set'oints for this alternative are:
?
Proposed Present l-2500 psig 2500 psig safety valve n
2300 psig 2355 psig reactor trip PbRVsetpoint 2450 psig 2255 psig L
2155 psig 2155 psig gperatingpressure A tabulation of the pros and cons of this alternative is attached.
B&W recommended the approval of this alternative since it covers essentially all transients and maintains the safety analysis of the plants.
The staff reviewed e'ach alternative and concurred with the D&W
'h f
e-D.
. Ross, Jr., Deputy Director
/
Division of Project Management
/
Enclosure :
As stated 228 227 A
s.
~.
ENCLOSURE 1 i
LIST OF ATTENDEES B&W MEETING 4/18/79 Organization Name L. B. Ma rsh NRC/f!RR/ DOR S. Newberry NRC/f;RR/nSS C. Graves NRC/NRR/ DSS R. C. Jones B&W/ECCS Analysis D. F. Hallman B&W/ Plant Performance Svs.
J. H. Taylor B&W/ Licensing E. A. Womack B&W/ Engineering R. E. Ham B&W/ Customer Service Thadani NRC/QSS
[AF. Ross NRC/DPM W. A. Smith Bechtel F. Odar NRC/ DSS /AB H. A. Wilber NRC/I&E S. Israel NRC W. Minners NRC E. V. Imbro NRC J. A. Castanes B&W/C&I Engineering R. W. Winks B&W/ Plant Design D. H. Beckham NRC/DPM L. Beltracchi NRC/ICSB M. Fairtile NRC/ DOR L. R. Cartin B&W/ Plant Design G. N. Lauben NRC/ DSS C. Berlinger NRC/NRR/ DOR /RSS S. Carody Duke Power E. Case NRC/NRR R. Mattson NRC/ DSS F. Schroeder NRC/ DSS.
A. Szukiewicz NRC/ DSS A. Schwencer hRC/ DOR A. Oxfurth NRC/ISE B. Clayton NRC/NRR/DPM J. Calvo NRC/NRR/ DSS f
f 228 228 C
\\
~. _.
ASSESSMENT OF ALTERNATIVE 4 l
PRO CON t
1.
Eliminate PORV actuation following 1.
Somewhat potential for essentially all anticipated transients.
sourious reactor trips 2.
Preserves validity of analyses serving 2.
Eliminates runback as basis for current operating licenses.
capability on lo load.
Turbine trip (i.e.,
increases number of trips by desian).
3.
Reduces (relative to current 3.
May still open on some setpoints) probability of PORV infrequent AT's (i.e.,
and PSV actuation (PORV is its still in the system) isolatable).
4.
Preserves venting capac?ty for 4.
PORV may open spuriously high pressure transients (e.g., ATWS)
(i.e., its still in the system).
5.
Can be implemented immediately with setpoint change in control coom.
6.
Lessens probability of actuating PORY for infrequent AT's (e.g., rod withdrawal) 7.
More forgiving of delay in auxiliary feedwater supply.
(Longer time to OTSG dryout.)
.a 228 229 r
s, a