ML19250C138
| ML19250C138 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/25/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Proffitt W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 7911080275 | |
| Download: ML19250C138 (17) | |
Text
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% PDR g# "*4 UNITED STATES 3g
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NUCLEAR REGULATORY COMMISSION 3( i j g WASHINGTON, D. C. 20555 e
October 25, 1979 Docket No.: 50-404 and 50-405 Mr. W. L. Proffitt Senior Vice President -
Power Operations Virginia Electric & Power Company P. O. Box 26666 Richmond, Virginia 23261
Dear Mr. Proffitt:
SUBJECT:
10 CFR 50.54 REQUEST REGARDING THE DESIGN ADEQUACY OF BABC0CK
& WILC0X NUCLEAR STEAM SUPPLY SYSTEMS UTILIZING ONCE THROUGH STEAM GENERATORS (NORTH AhNA UNIT NOS. 3 & 4)
Several hardware and procedural changes have been made to operating B&W plants to reduce the likelihood of recurrer.ce 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 mere 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 'J-tube steam generator.
In addition, it provides for less water invent ory thus making a steam line break less severe. However, the relatii.ely low water inventory with the presence of a liquid-vapor heat transfer interface in the active heat transfer zone closely couples the primary system to the steam generator conditions with a consequently high sensitivity to feedwater-flow rate perturbations. Enclosure I 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 th_e pressurizer sizing, ICS design, and PORV/ reactor trip set points.
As part cf the post TMI-2 effort, detailed analyses have been made of under-coolir.3 transients for B&W plants. However, due to the sensitivity of the OTSG design, B&W plants have also been experiencing a number of relatively severe overcooling events.
1274 237 7911 080
Mr. W. L. Proffitt For your information, NRC is initiating a research task to quantitatively assess B&W system designs, including the integrated centrol 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 canplete detemination of risk wili.ut 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, approximate quantifi-cation of expected frequency of selected etent sequences and sensitivity studies for reliability of operator resporse. The study will focus on the risk implications of the sensitivity if the B&W design and on the potential interactions arising from the integrated control system. We estimate this study to be completed in abott six months. We will use the Crystal River, Unit 3 plant as the refecenced facility to be analyzed.
We have been holding generic discussions w'th Babcock and Wilcox Company concerning this matter. However, system 'ensitivity to feedwater transients involves balance-of-plant equipment and rystems as well as the nuclear steam supply system, and such plant-specific rnaracteristics 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 information is requested to allow us to determine whether it is necessary to halt all or portions of the con'truction of your plant pending the results of our study. We request you provide:
a) Identify thu 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 ed.
b) Identify whether action of the ECCS or MS (or operator action) is necescary to protect the core following the most severe over-cooling i. Insient identified.
If these systems are required, you should show that its design criterion for the number of actuation cycles is adequate, considering arrival rates for excessive cooling transients.
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, c) Provide a schedule of completion of installation of the identified systems and components.
d)
Identify the feasibility of halting installation of these systems and components as compared to the feasibility of completing installation and then effecting significant chances in these systems and components.
e) Comment on the OTSG sensitivity to feedwater transients.
f) Provide recommenoations on hardware 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. natter 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.
Please call Dr. Anthony Bournia at (301) 492-7200 if you ha/e any questions concerning this letter.
u ncerely, Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
As stated cc: See next page 1274 239 e
Mr. W. L. Proffitt cc: Mr. Anthony Gambaradella Cl arence T. Ki pps, Jr., Esq.
Office of the Attorney General 1700 Pennsylvania Avenue, N.W.
11 South 12th Street - Room 308 Washington, D. C.
20006 Richmond, Virginia 23219 Carroll J. Savage, Esq.
Richard M. Foster, Esq.
1700 Pennsylvania Avenue, N.W.
211 Stribling Avenue Washington, D. C.
20006 Charlottesville, Virginia 22903 Mr. James C. Dunstan Michael W. Maupin, Esq.
State Corporation Commission Hunton, Williams, Gay & Gibson Comonwalth of Virginia P. O. Box 1535 Blandon Building Richmond, Virginia 23212 Richmond, Vi rginia 23209 Mrs. June Allen Alan S. Rosenthal, Esq.
412 Owens Drive Atomic Safety and Licensing Appeal Board Huntsville, Alabama 35801 U.S. Nuclear Regulatory Comission Washington, D. C.
20555 Mr. James Torson 501 Leroy Michael C. Farrar, Esq.
Socorro, New Mexico 87801 Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Comission Mrs. Vegaret Dietrich Washington, D. C.
20555 Route 2, Box 568 Gordonsville, Virginia 22942 Dr. John H. Buck Atomic Safety and Licensing Appeal Board Willi am H. Rodgers, Jr., Esq.
U.S. Nuclear Regulatory Comission Georgetown University Law Center Washington, D. C.
20555 600 New Jersey Avenue, N.W.
Washington,D. C.
20001 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commis_sion Mr. Peter S. Hepp Washington, D. C.
20555 Executive Vice President Sun Shipping & Dry Dock Company Mr. Michael S. Kidd P. O. Box 540 U.S. Nuclear Regulatory Comission Chester, Pennsylvania 19013 P. O. Box 128 Spotsivania, Virginia 22553 Mr. R. B. Briggs Associate Director Dr. Paul W. Purdom 110 Evans Lane Department of Civil Engineering Oak Ridge, Tennessee 37830 Drexel University P's. ladel phi a, Pennsyl vani a 19104 id4 240
Mr. W. L. Proffitt cc: Dr. Lawrence R. Quarles Apartment No. 51 Kendal-at-Longwood Kennett Square, Pennsylvania 19348 Mr. Irwin B. Kroot Citizens Energy Forum P. O. Box 138
" clean, Vi rginia 22101 James B. Dougherty, Esq.
Potomac Alliance 1416 S Street, fi.W.
Washington, D. C.
20009 1274 241
ENCLO5tfRE 1 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 steam generators which are used in other pressurized water reactors.
Each steam generator has approximately 15,000 vertical straight 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 generator tubes, while the secondary coolant flows up from the bottem on the shell side of the OTSG. The secondary coolant turns to steam about half way uo, with the remaining length of the steam generator being used to superneat the steam.
The secondary-side heat transfer coefficient, in the steac space of the OTSG, is much less than that in the bottom liqui' section. Inis results in a heat transfer rate from the primary ~ system which is quite sensitive to the liquid level in the steam generators.
If a feedwater increase event occurs, the 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,ystem depressurization. By contrast, if a feedwater decrease event occurs, the~overall heat transfer decreases, the outlet primary temperature increases, and a pressurization transient ensues.
In either of these cases, the response of the primary system pressure and pressurizer level to a change in main feedwater flow rate (or temperature) is comparatively rapid. These rapid' primary syster pressure chances due to changes in fee:: water conditions is knowr. herein as syster " sensitivity" and is 1274 242
2-unique to the B&W OTSG design.
Following the incident at Three Mile Island, various actions were taken to increase the reliability of the auxild.try feedwater systems and improve plant transient response. System modifications to increase the reliability of the AFW may have resulted in more frequent AFW initiation. However, use of AFW results in introduction of cold (100*7 vs. 400 F) feedwater into the more sensitive 0
upper section of the steam generators. This may act to enhance system sensitivity.
Further system modifications provide control-grade reactor trips based on While secondary system malfunctions, such as turbine or feedwater pump trip.
these reactor tr.ps do serve to reduce undercooling feedwater transients by reducing reactor power promptly #illowing LOMFW, they may Omplify subsequent overcouling.
A reexamination was made of small break and loss of feedwater events for B&W plants.
This resulted in a modification of operator procedures for dealing with a small break, which include prompt RCP trip and raising the water level in the steam generators to (95%) to promote natural circulation. Both these actions are taken when a prescribed low pressure set point is reached in the reactor coolant system and for anticipated 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 the PCRV set point and lowering the high oressure reactor trip. While these actions have been successful in reducing the frequency of PORV operation, they I
i 1274 2'43 have resulted in an increased number of reacter trips. This occurs because the reactor will' now trip for transients it previously would have ridden through by ICS and PORV operation.
The staff is concerned by the inherent responsiveness of B&W OTSG design. While some specific instances are presented in the next section of this paper, the staff 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 prompt 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 On August 23, 1979 the staff met with the P&W licensees to discuss recent feedwater transients. One aspect which is of interest is the relationship of the operator to the functioning of the main feedwater system.
In at least one 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 th.e top of the steam generators, which increased the severity of '.he cooldown transient.
1274 244
4 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 underfeed to overfeed conditions, One causing a reactor trip and sometimes a high presture injection initiation.
undesirable 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 pressure). Thus, for a certain period of time the operato The same may not know whether they are having a LOCA or an overcooling event.
This type of behavior can be initiated by the nomal reactor control system.
was demonstrated by a December 1978 event at Oconee, where failure of a con recorder led to reactor trip, a feedwater transient, and ESF actuation.
grade T,yg A partial list of recent B&W transients and their effects is contained in the Appendix to this report.
Role of the Pressurizer Level Indicator III.
A major area of concern arising from the B&W OTSG sensitivity, is the response 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.
rate for this event appears to be on the c. der of.1.2 per reactor year, but could be on the increase due to the potential for more reactor trips and feed-This is of water transients resulting from post-TMI-2 system modifications.
concern because an overcooling event could empty the pressurizer, thereby the potential for feming 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 somewhat high for an event with a recurrence interval of a few years.
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5 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 pressurizef 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-MFW
~
The ICS appears 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 illustra:es 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; pressurizer level shrinks, pressure reaches 160 psi, RS actuates;-
(Possible RCP seal failure).
5.
Operator manually controls AFW (possibly MFW instead or in addition, if MFW n' t 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 interruption of natural 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 that an upgraded safety quality ICS, which is designed to balance power to OTSG level 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 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.
.I f 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 HPI. One problem is the difficulty in differentiating between a small 1274 247 e
break LOCA and an. excessive feedwater transient. The operator would be forced to assume a small LOCA until proven othemise. However, following the small break procedures and introducing cold auxiliary feedwater, 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 coolcown 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 p0RV and high pressure trip set point. Some of the sensitivity also arises from inadequacies in the ICS to deal with expected plant perturbations.
Regardless of the reasons, B&W plants are currently experiencing a number of feedwster. transients which the staff feels a.e 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 plant.
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SUMMARY
FACILITY TRANSIENT DATE DESCRIPTI0il CR-3 8/16/79(0259 Reactor Trip on liigh Pressure - 4 to 3 RCP, A-S/G underfed -72% Pwr.
8/16/79 (1125)
Reactor Trip on liigh Pressure - 3 RCP - A-S/G underfed - 451 Pwr.
8/17/79 (0706)
Reactor Trip on liigh Pressure - 3 RCP - A-S/G underfed - 481Pwr.
8/17/79 (1825)
Reactor Trip on liigh Pressure - 3 RCP - A S/G underfed - 26% Pwr.
8/02/79 (0202)
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 ll1 Press - 75%
Oconee-1 6/11/79 (0333)
Reactor Trip on Anti. Trir (LOFW) - 99% Pwr.
6/11/79 (0752)
Reactor Manually Tripped when FWPT "lB" Tripped Oconce-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.
h[$
Davis-Desse NONE A
N h
4
ENCLOSURE IREP - INITIAL PLANT STUDY We have attempte'd to develop a general framework for the conduct of a limited risk assessunt 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 commercial operation in March 1977.
The project, as presented in Figure 1, will require the following tasks:
1.
A survey of the LER files as now established in ORNL and A0 reports, as well as the Sandia and Fluor-Zion systems interactions studies to identify interactions and comon mode failures which have occurred in similar plants.
This survey should parallel construction of system logic models and event trees since it will enure that actual experience is incorporated into the assessments performed.
Event trees for lo:;s-c '-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.
- 3.
Fault trees for the key systems identified in the event trees. They will 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 the following systems: auxiliary feedwater and secondary steam relief, high pressure emergency core cooling in the injection and recirculation modes, low pressure emergency core cooling in both injection and recirculacion 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.
will be permitted provided a written basis is provided. This basis will 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 from tr.e containment.
will aid in categorizing releases into apprec-iate release categories.
1274 250 e
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 comon 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 imediate access to as-built dr:. wings and procedures as well as a representative of the plant operations staff. This will permit verification of engineering and procedural deuils and will minimize informazior, 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 plant data, detenninistic calculations may oe required to understand the behavior of the plant under off-nortnal conditions. This may also involve real-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 selected basis 1274 251 f
FIGilRE 1 ANALYSIS OF ACCIDENT EVEllT TREE 4
C0!ISTRUCT10tl PROCESSES AS tlEEDED LER SURVEY Ptn,L TOGETilER EXISTING INFO 9
ALREADY C0t!!'II,ED if CATECORIZAT10tl
)
LICEllSIllG FAULT TREE 4
QUAllTIFICATI0li OF EVENT REC 0telEllDATIONS C0tlSTRUCTI0tl TREE SEQUENCt'S 1,
DETERiittilSTIC CALCllLATI0tlS BY VENDOR A3 " '
AllALYSIS OF
SUMMARY
htEPOR1 i
ilIGli PRESSURE -
0F RESULTS LOW PRESSURE INT.
y N
4 hJ l
'n
.O NRR AND PEER REVIEW 4
ENCLOSURE 3 PRELIMINARY IDENTIFICATION OF SYSTEMS AND COMPONENTS THAT MAY BE IMPACTED BY DESIGN CHANGES HPI System EFW System DHR System CFT System RCS Pressure Control System Makeup / Letdown System SG Pressure Control Systeni Steam Generator Pressurizer Quench Tank Control Room Layout P.CS Piping 1274 253 f
e e