ML19250C141
| ML19250C141 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 10/25/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 7911080283 | |
| Download: ML19250C141 (16) | |
Text
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SM RfC UNITED STATES NUCLEAR REGULATORY COMMISSION y
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october 25, 1979 Docket No.: 50-329 and 50-330 Mr. S. H. Howell Vice President Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201
Dear Mr. Howell:
SUBJECT:
10 CFR 50.54 REQUEST REGARDING THE DESIGN ADEQUACY OF BABC0CK
& WILC0X NUCLEAR STEAM SUPPLY SYSTEMS UTILIZING ONCE THROUGH STEAM GENERATORS (MIDLAND UNIT N05.1 & 2)
Several hardware and procedural changes have been made to operating B&W plants to reduce the likelihood of recurrence of a TMI-type accident. These changes have been in the area of auxiliary feedwater systems, integrated control system, reactor protection system, small-break loss-of-coolant accident analysis and operator training and procedures.
However, at this time, we are beginning to look more deeply into additional design features of B&W plants to consider if any further system modifications are necessary.
The use of once-through-steam-generators (OTSG) in B&W plants has an opera-tional advantage in that it provides a small degree of steam superheat, as contrasted with the conventional saturated U-tube steam generator.
In addition, it provides for less water inventory thus making a steam line break less severe. However, the relatively low water inventory with the presence of a liquid-vapor heat transfer interface in the active heat transfer zene closely couples the primary system to the steam generator conditions with a consequently high sensitivity to feedwater-flow rate perturbations. Enclosure 1 to this letter addresses system problems and staff concerns in this area. At present, we are investigating whether B&W plants are overlysensitive to feedwater transients, due to the OTSG concept, as coupled with the pressurizer sizing, ICS design, and PORV/ reactor trip set points.
As part of the post TMI-2 effort, detailed analyses have been made of under-cooling transients for B&W plants. However, due to the sensitivity of the OTSG design, B&W plants hsve also been experiencing a number of relatively severe overcooling events.
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Mr. S. H. Howell For your infomation, NRC is initiating a research task to quantitatively assess B&W system designs, including the integrated control system, aimed at identifying obvious accident sequences leading to core damage having a high frequency as compared to the Reactor Safety Study, see Enclosure 2.
( A complete determination of risk will not be attempted).
The objective of this assessment is to identify high-risk accident sequences (including TMI implications) utilizing event tree and simplified fault tree analyses.
Included will be estimation of release categories, approximai.e quantifi-cation of expected frequency of selected event sequences and sensitivity studies for reliability of operator response. The study will fo:us on the risk implications of the sensitivity of the B&W design and on the potential interactions arising from the integrated control system. We estimate this study to be completed in about six months. We will use the Crystal River, Unit 3 plant as the referenced facility to be analyzed.
We have been holding generic discussions with Babcock and Wilcox Company concerning this matter. However, system sensitivity to feedwater transients involves balance-of-plant equipment and systems as well as the nuclear steam supply system, and such plant-specific characteristics must be considered.
We are also considering whether it is necessary to halt portions of the construction of B&W plants, pending the outcome of the reliability assess-ment. As a preliminary consideration, we have identified those systems and components that may be impacted by possible design changes as a result of this study. Enclosure 3 is a preliminary listing of such systems and components.
Under the authority of Section 182 of the Atomic Energy Act of 1954, as amended, and Section 50.54(f) of 10 CFR Part 50, additional 1: '=ation is requested to allow us to determine whether it is necessary +.o halt all or portions of the construction of your plant pending the results of our study. We request you provide:
a)
Identify the most severe overcooling events (considering both anti-cipated transients and accidents) which could occur at your facility.
These should be the events which causes the greatest inventory shrinkage. Under the guidelines that no operator action occurs before 10 minutes, and only safety systems can be used to mitigate the event, each licensee should show that the core remains adequately cool eri.
b)
Identify whe;her action of the ECCS or RPS (or operator action) is necessary to protect the core following the most severe over-cooling transient identified.
If these systems are required, you should show that its design criterion for the number of actua+ ion cycles is adequate, considering arrival rates for excess 1ve cooling transients.
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. c) Provide a schedule of completion of installation of the identified systems and componcats.
d)
Identify the feasibility of hal'.ing installation of these systems and components as compared to the feasibility of completing installation and then effecting significant changes in these systems and components.
e) Comment on the OTSG sensitivity to feedwater transients.
f) Provide recommendations on hardv are and procedural changes related to the need for and methods for damping primary system sensitivity to perturbations in the OTSG.
Include details on any design adequacy studies you have done or have in progress.
We are sending similar letters to all utilities holding construction permits for plants with B&W nuclear steam supply systems.
We request your reply by December 3,1979. We believe that a meeting with you and the other utilities together with the staff and the Babcock and Wilcox Company to discuss this matter would be beneficial to all parties.
At that time, we will provide further details on the Crystal River Study.
We are scheduling such a meeting for November 6,1979 at 10:00 a.m. in Room P-422 at our offices in Bethesda, 7920 Norfolk Avenue, Bethesda, Maryland.
It should be noted that this request is apart from and in addition to our earlier 50.54(f) request to you dated March 21, 1979.
Please call Dr. Anthony Bournia at (301) 492-7200 if you have any questions concerning this letter.
Sincerely, N
Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
As stated cc:
See next page 1274 268
I l_
Consumers Power Company ccs:
Michael I. Miller, Esq.
Isham, Lincoln & Beale Suite 4200 One First National Plaza Chicago, Illinois 60603 Judd L. Bacon, Esq.
Consuners Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Mr. Paul A. Perry Secretary Consunars Power Corapany 2125.. Michioan Avenue Jackson, Mich!gan 49201 Myron M. Cherry, Esq.
'One IBM Plaza Chicago, Illinois 60611 Mary Sinclair 5711 Summerset Drive Midland, Michigan 48640 Frank J. Kelley, Esq.
Attorney General State cf Richigan Environmental Protection Division 720 Law Building Lansing, Michigan 48913 Mr. Wendell Marshall Route 10 Midland, Michigan 48640 Grant J. Merr' r.t, Esq.
Thompson, Nicisen, Klaverksap & James 4444 IDS Center 80 South lighth Street Minneapclis, Minnesota 55402 t
Mr. Dan van Farowe, Chief Division of' Radiological Health Department of Public Health P. 0. Box 33035 Lansing, Michigan 48909 1274 269
Primary System Perturbations induced by Once Through Steam Generator I.
Introduction-B&W plants employ a once through st. m generator (OTSG) design, rather than U-tube 51 cam generators which are used in other pressurized water F.ach steam generator has approximately 15,000 vertical straight reactors.
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' tubes, with the primary coolant entering the top at 603-608 F and exiting the bottom at about 555 F.
Primary coolant flows down inside the steam 0
generator tubes, while the secondary coolant. flows up from the bottom on the shell side of the OTSG.
The secondary coolant turns to steam about half way up, with the remaining length of the steam generator being used to superheat the steam.
The secondary-side heat transfer coefficient. In the stear space of the OTSG, is much less than that in the bottom liquid section.
Inis results in a heat transfer rate from the p-imary system which is quite sensitive to the liquid level in the steam generators.
If a feedwater increase event occurs, the 4
liquid-vapor interface rises, increasing the overall heat transfer. This decreases the outlet temperature below 555 F and initiates an overcooling event, which can lead to primary system depressurization. By contrast, if a feedwater decrease event occurs, the overall heat transfer decreases, the outlet primery temperature increases, and a pressurization transient ensues.
In either of these cases, the response of tne primary system, pressure and,
pressurizer level to a change in main feedaater flow rate (or temperature) is comparatively rapid.
These rapid primary syste pressure chances due to changes in feedwater conditions is knowr. herein as syste:r " Sensitivity" and is 1274 270 me N A, A.W l
unique to the B&W OTSG design.
Following the incident at Three Mile Island, various actions were taken to increase the reliability of the auxiliary feedwater systems and improve plant System modifications to increase the reliability of the AFW transient response.
However, use of AFW results may have resulted in more frequent AFW initiation.
in introduction of cold (100 F vs. 400 F) feedwater into the more sensitive This may act to enhance system sensitivity.
upper section of the steam ge erators.
Further system modifications provide control-grade reactor trips based on While secondary system malfunctions, such as turbine or feedwater pump trip.
these reactor trips do serve to reduce undercooling feedwater traasients by reducing reactor power promptly following LOMFW, they may amplify subsequent overcooling.
A reexamination was made of small break and loss of feedwater events for B&W This resulted in a modificat' ion of operator procedures for dealing plants.
with a small break, which iriclude prompt RCP trip and raising the water level Both these in the steam generators to (95%) to promote r.atural circulation.
cctions are taken when a prescribed low pressure set point is reached in the reactor coolant system and for antic.ipated transients such as loss of feedwater these actions may amplify undesirable primary system responses.
In addition to the post-TMI changes discussed above, actions were also taken to reduce the challenges to the power operated relief valve (PORV) by raising While tnese the PORV set point and lowering the high pressure reactor trip.
actions have been successful in reducing the frequency of PORV operation, they
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have resulted in an increased number of reactor trips. This occurs because the reactor will' n'ow trip for transients it previously would have ridden
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through by ICS and PORV operation.
The staff is concerned by the inherent responsiveness of B&W OTSG design. While so' e specific instances are presented in the next section of this paper, the staff m
concerns are also of a general nature.
It is felt that good design practice and maintenance of the defense-in-depth concept, requires a stable well-behaved system.
To a large part, meticulous operator attention and pcmot manual action is used on these plants to compensate for the system sensitivity, rather than any inherent design features.
The staff believes that the general stability of the B&W plant control systems should be improved, and that plant response to OTSG feedwater perturbations be dampened.
II.
Recent Feedwater Transients 0,n August 23, 1979 the staff met with the B&W licensees to discuss recent feedwater transients.
One aspect which is of interest is the relationship of In at least one the operator.to the functioning of the main feedwater system.
instance an operator manually opened.a block valve in series with a control valve (partly open but thought to be closed).
This resulted in an overfeed condition.
In several recent events the feed flow was reduced to the point where the reactor tripped on high pressure. Subsequent overfeed reduced pressure to below 1600 psi, where HPI was initiated, reactor coolant pumps tripped, and auxiliary feedwater flow introduced into the top of the steam generators, which increased the severity of the cooldown transient.
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It appears that in many cases the main feedwater control system does not react quickly enough or is not sufficiently stable to meet feedwater requirements.
Rather, the system will often oscillate from und.rfeed to overfeed conditions.
One causing a reactor trip and sometimes a high pressure injection initiation.
und'esirable element of this lack of stability is that overcooling transients on the primary side proceed very much like a small break LOCA (decrease in pressurizer level and pressurel.
Thus, for a certain period of time the operators The same may not know whether they are having a LOCA or an overcooling event.
This type of behavior can be initiated by the nonnal reactor control system.
was demonstrated by a December 1978 eva.t at Oconee, where failure of a recorder led to reactor trip, a feedwater transier.t, ard ESF actuation.
grade T,yg A partial list of recent B&W transients and their effects is cor.tained in the Appendix to this report.
Role of the Pressurizer Level Indicator _
- III, A major area of concern arising from the B&W OT5G sensitivity, is the respon Several B&W feedwater transients have led to of pressurizer level indication.
Most notatble was a November 1977 incident loss of pressurizer level indication.
The arrival at Davis Besse where level indication was lost for several minutes.
ral.e for this event appears to be on the order of.1.2 per reactor year, but could be on the increase due to the potential for more rea: tor trips and fee This is of water transients resulting from post-TMI-2 system modifications.
concern because an overcooling event could empty the pressurizer, the the potential for forming a steam bubble in the hot leg which may interrupt The staff feels that natural circulation, following RCP trip on low pressure.
the uncertainties associated with two phase natural circulation are somewh high for an event with a recurrence interval of a few years.
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Additionally, the' staff believes that good design practice and adherence to the defense-in-depth concept,'would require that plant operators be aware of the reactor's status during expected transients. A low-level off-scale reading on pressurizer level makes it impossible for the operators to assess system ihventory and more difficult to differentiate between an accident and an excessive cooldown transient.
The staff feels that the frequency with which this
. situation occurs is undesirable.
Some concerns also exist with regard to the operation of the pressurizet heaters when loss of level takes place.
Nonsafety grade control circuitry trips the heaters off when pressurizer level is low.
If these nonsafety grade cutoffs should fail, the heaters would be kept on while uncovered.
This situation has the potential of overheating the pressurizer to the failure point, as happened
.with a test reactor at Idaho Falls.
IV.
Role of ICS-M.W
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The ICS appe~ars to paly a significant role in the plant's feedwater response.
The staff is currently reviewing an FMEA study on the ICS.
However, review of operating experience suggests that the ICS often is a contributor to feedwater transients.
In some cases the ICS appeared inadequate to provide sufficient plant control and stability.
Some of the utility descriptions of feedwater transients (as sumarized in the minutes of a meeting on August 23,1979) emphasized the role of the operator in operating the MFW system.
The following sequence illustrates the type of event and system response which the staff feels could potentially occur.
1.
Reactor at'100% power.
2.
Reactor trip, from arbitrary cause (does not matter).
3.
Plant stabilizes in hot shutdown, for a few minutes, heat rejection to condenser (and/or secondary dump valves).
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4.
Overfeed transient (MFW) (not uncomon to B&W) causes overcooling; pressurizar level shrinks, pressure reaches 160 psi, RS actuates; RCP tripped; AFW on.
(Possible RCP seal failure).
5.
Operator manually controls AFW (possibly MFW instead or in addition, if MFW n' c isolated such that OTSG level comes up to 95% of operating range.
This massive addition of cold water may lead to emptying of pressurizer and interrapi.isa of natcral circulation (or, the hot leg may flash due to depressurization and interrupt natural circulation even if pressurizer does not empty).
6.
HPI delivers cold water, no heat transfer in OTSG; vapor from core leads to system repressurization; steam may condense or PORV may lift.
7.
No pump restart criteria available, circulation may not be reestablished.
It appears th?t an upgraded safety quality ICS, which is designed to balance
. power to OTSG 1evel in a better fashion, could reduce the sensitivity, illustrated in the above sequence.
V..
Role of ECCS and Auxiliary Feedwater it is known that some feedwater transients result in overcooling to the extent a that the HPI actuation setpoint is reached. Traditionally, the operator isolates letdown and turns on an extra makeup pump following trip so as to avert this actuation.
If this manual action is not perfomed quickly enough, or if the cooldown transient is too severe, the HPI set point will be reached and the pumps automatically started. Following procedures, the operator would then trip all mai-coolant pumps and utilize recovery procedures based on the plant symptoms, If the incident was actually a feedwater event and not a small LOCA, he would then presumably go to the loss of forced circulation procedures. When pressure has recovered such that the coolant system has become 50 F subcooled, the operator can secure.4PI.
One problem is the difficulty in differentiating between a small 1274 275
~ ' break LOCA and an. excessive feedwater transient. The operator would be forced to assume a small LOCA until proven otherwise. However, following the small
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break procedures a'nd introducing cold auxiliary feed / ster, may increase the severity of an overcooling event.
Initiation of AFW and delivery to the OSTG, especially if accompanied by filling to the high level required by new pro-cedures (95%) will continue the cooldown and depressurization. Thus, the AFW system acts to increase the responsiveness of the reactor to feedwater transients where excessive cooldown is occurring.
VI.
Conclusions The staff believes that the current B&W plants are overly responsive to feedwater transients because of the OTSG design, pressurizer sizing and PORY and high pressure trip set point.
Some of the sensitivity also arises from
. inadequacies in the ICS to deal with expected plant perturbatior.s.
Regardless of the reasons, B&W plants are currently experiencing a number of feedwater transients which the staff feels are undesirable. The staff believes that modifications should be considered to reduce the plant sensitivity to these events and thereby improve the defense-in-depth which will enhance the safety of the. pla...
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APPENDIX FEE 0 WATER TRANSIENT
SUMMARY
FACILITY TRANSIENT DAIE DESCRIPTION CR-3 8/16/79 (0259 Reactor Trip on liigh Pressure - 4 to 3 RCP, A-5/G underfed -721 Pw 8/16/79 (1125)
Reactor Trip on liigh Pressure - 3 RCP - A-S/G underfed - 45% Pwr.
8/17/79 (0706)
Reactor Trip on liigh Pressure
. 3 RCP - A-S/C underfed - 48%Pwr.
8/17/79 (1825)
Reactor Trip on liigh Pressure - 3 RCP - A S/G underfed - 26% Pwr.
8/02/79 (02.2)
Reactor Trip on Low-Low Level in both S/G - 10% Pwr.
ANO-1 8/13/79 (1749)
Turbine Trip - Antic. Trip did not. work - Rx Trip on 111 Press - 75 Oconee-1 6/11/79 (0333)
ReactorTriponAnti. Trip (LOFW)-99%Pwr.
6/11/79 (0752)
Reactor Manually Tripped when FWPT "lB" Tripped Oconee-2 5/07/79 (0346)
Reactor Trip on liigh Pressure - feedwater oscillations - 18% Pwr.
6/03/79 (2046)
Reactor Trip on liigh Pressure - feedwater oscillations - 30% Pwr.
Rancho Seco 7/12/79 (1714)
Reactor Trip on Antic. Trip 'LOFW) - 100% Pwr.
Davis-Desse NONE N
O N
N N
ENCLOSURE IREP - INITIAL PLANT STUDY We have attempte.d to develop a general framework for the conduct of a limited risk assessment of a B&W reactor aimed at identifying any unique risk-impacting sequences relative to the Reactor Safety Study. An absolute determination of risk is not intended.
We have selected Crystal River 3, a plant owned and operated by Florida Power Corporation, for analysis.
The architect-engineer for this Babcock and Wilcox reactor was Gilbert Associates.
It began conrnercial operation in March 1977.
The project, as presented in Figure 1, will require the following tasks:
A survey of the LER files as now established in ORNL and A0 reports, as 1.
well as the Sandia and Fluor-Zion systems interactions studies to identify interactions and common mode failures which have occurred in similar plants.
This survey should parallel construction of system logic models and event trees since it will ensure that actual experience is incorporated into the assessments performed.
Event trees for loss-of-coolant accidents and transient conditions.
Specific 2.
attention will be given to more frequent LOCAs and these will include a feed-water transient tree which incorporates experience at B&W plants and will explore the post-TMI modifications.
Emphasis will be given toward under-standing the human coupling interaction between systems at the event tree sequence level.
Fault trees for the key systems identified in the event trees.
They will
- 3.
be constructed to the component level and will include control, actuation, and electric power considerations.
Human errors will be included as well as the ability of the operator to cope in the time span available.
Our preliminary opinion is that simplified fault trees will be required for auxiliary feedwater and secondary steam relief, the following systems:
high pressure emergency core cooling in the injection and recirculation modes, low pressure emergency core cooling in both injection and recirculation modes, containment spray and containment heat removal systems and a limited study of loss of AC power, considering the 480 and 4160 busses and the emergency diesel generators, with limited analysis of high voltage switch-Separate fault trees will probably be required for ECCS and yard faults.
AFWS initiation logic and the system trees must include the contribution from auxiliary systems such as instrument air, ventilation, component Truncation of the fault trees cooling, etc., and control-induced failures.
This basis will will be permitted provided a written basis is provided.
present the rationale why no coupling of cutsets or event sequences is expected from further development of the tree.
An investigation of the adequacy of high pressure-low pressure interfaces.
4 Analysis of the physical phenomena associated with cominant sequences to 5.
This obtain estimates of the magnitude of releases fro-tne containment.
5;ill aid in categorizing r21 eases into apprc:riate release categories.
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2 To conduct a program of this magnitude in a short time period, delays assoc-iated with acquiring and transferring information must be minimized. Optimally, the event tree and fault tree analysts should share a cannon location during the initial portion of the project. As the fault trees progress below the top logic, however, the analysts should be located at or near the site with innediate access to as-built drawings and procedures as well as a representative of the plant operations staff. This will permit verification of engineering and procedural details and will minimize information transfer and print re-production. Access should also be arranged between the fault tree analysts at the site, the remaining team in Bethesda, the architect-engineer, and the vendor.
In addition to basic y'. ant data, deterministic calculations may be required to
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understand the behavior of the plant under off-normal conditions.
This may also involve -ea'l-time simulation at an appropriate simulator to the extent possible. The arrangements with the vendor should cover this possibility and it may be desirable to have confirmatory calculations made by one of the NRC contractors on a se'?cted basis 1274 279 4
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FICllRE I ANALYSIS OF ACCIDENT 4
p EVEllT TREE C0t3STRUCT10ti PROCESSES AS NEEDED T.ER SURVEY P'JLL TOGETilER EXISTING INFO p
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LICENSING FAULT TREE 4
D,UAllTIFICATI0li 0F EVEtiT
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REC 0tatFNDATIONS C0tlSTRUCTI0t1 TREE SEQUENCES tr DETEllMINISTIC CALCULATIONS BY VEllD0lt
^3 AllALYSIS OF
SUMMARY
RETO IIIGil PRESSURE -
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ENCLOSURE 3 PRELIMINARY IDENTIFICATION OF SYSTEMS AND COMPONENTS THAT PAY BE BY DESIGN CHANGES HPI System EFW System DHR System CFT System RCS Pressure Control System Makeup / Letdown System SG Pressure Control System Steam Generator Pressurizer Quench Tank Control Room Layout RCS Piping 1274 281
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