ML20037A223
| ML20037A223 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/16/1978 |
| From: | Bosnak R Office of Nuclear Reactor Regulation |
| To: | Benaroya V, Lainas G, Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19220B244 | List: |
| References | |
| NUDOCS 7904250549 | |
| Download: ML20037A223 (41) | |
Text
["
UNITEo ST ATES y
NUCLE AR REGULATORY C(:MMISSION 7
t M /, $
W ASHINGTON. D. C. 20555
- h E
l s., s { j JAh I b 1978 MEMORANDD! FOR:
V. Benaroya, Chief, Auxiliary Systems Branch, DSS G. Lainas, Chief, Containment Systems Branch, DSS T. Novak, Chief, Reactor Systems Branch, DSS R. J. Bosnak, Chief, Mechanical Engineering Branch, DSS FROM:
THREE MILE ISLAND UNIT 2, REQUEST FOR RELIEF FROM CERTAIN
SUBJECT:
ASME CODE SECTION XI PUMPS AND VALVES INSERVICE TESTING REQUIREMENTS - DOCKET No. 50-3.O The Mechanical Engineering Branch, Division of Systems Safety, has received the request for relief from certain requirements of the ASME Code Section XI,1974 Edition including Addenda through Surner 1975 from Metropolitan Edison Company.
The information related to the proposed inservice testing program is attached. We request your assistance to evaluate the acceptability of the programs in your area, especially, in regards to the completeness, classification, and relief requests from testing of pumps and valves. We would appreciate your comments as soon as possible.
So that we can establish a realistic schedule for completion of this review effort, please advise us by return memo your best estimate as to when you will be able to provide your comments.
l l
R J. Bosnak, Chief Mechanical Engineering Branch f
Division of Systems Safety
Attachment:
Meted ltr GQL 001 dtd. Jan. 3, 1978 with enclosures cc w/o encl:
R. Boyd, FM R. Hartfield, MIPC R. Mattson, SS J. Knight, SS D. Eisenhut, OR D. Ross, SS R. Tedesco, SS S. Varga, PM L. Shao, OR F. Cherny, SS H. Silver, PM J. Kovacs, SS a
ec _
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2.8 P_ressurizer Level Indication A discussion of pressurized water reactor operatinc characteristics is provided in Section 2.9 below.
An important parameter in under-standingthe dynamic performance of the reactor coolant system is the pressurizer level and its relationship to the water level in other parts of the rector-e lene system.
In this Section we provide a description of the pressurizer level sensing instrumentation ONW arrangement at TMI-2, which is typical of-aM-B&W reactors.
We also discuss the reliability of this instrumentation for measuring pressurizer level during rapid depressurization of the reactor coolant system such as occurred at TMI-2.
The general layout of a typical pressurizer level instrumentation system is given in Figure 2.1.
Three systems are installed.
For i
each, two impulse lines connect to the pressurizer; one near the top and one near the bottom.
The lines are routed to a differential- '
pressure transmitter, located near the bottom of containment in the annular region between the shield wall and the containment wall.
There are several -athee factors which could affect instrument accuracy in a depressurization event,-and which are not normally automatically
. corrected in the level readoutP
- 1. A rapid reduction in pressurizer pressure could cause liquid to flash in the reference leg (the line connecting the transmitter to the pressurizer near the top of the vessel in Figure 2.1).
Such flashing, should it be significant, could cause the instrument to indicate a falsely high pressurizer water level.
- 2. Degassing of liquid in the reference leg could also cause an error.
Dissolved gases could rapidly be driven out of the reference leg by this mechanism, and the level instrument would again indicate a falsely high level, t
- 3. Should the pressurizer depressurization occur rapidly, a venturi effect could in principle be crea$ted at the point where the reference leg joins the pressurizer vessel.
If this occurred, liquid could be drawn out of the reference leg causing the same inaccuracies in level indication noted above.
The.importance of each of these effects has been assessed assuming conditions which existed at TMI-2 prior to and during the event.
Calculations were performed to estimate the effects of both flashing and degassing. While the calculations indicated that some flashing could occur, the reduction in water level in the reference leg due to flashing is estimated to be less than one foot.
Because the distance between the taps is about 33 feet, the effect of this reduc-tion would be small.
Calculations also indicate that the effect of degassing of liquid in the reference leg is necligible. With regard to the venturi effect, it is estimated that gas velocities at the upper level sensing nozzle are too low to produce any significant effect.
We conclude from these assessments that the errors in level instrument indications during the event at 1MI-2 were not large.
In particular, potential effects that could cause falsely high indicated icvel were assessed not to be significant.
Therefore, the increasing level indicated by the instrumentation beginning about one minute 1
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into the event at TMI-2 is believ.ed to have been reliable and resulted from an increasing level of water in the pressurizer. However, as discussed in Section 2.9 below, water level in the pressurizer may not be a valid indication of water level in other parts of the reactor coolant' system for certain off normal conditions.
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1.0 INTRODUCTION
On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant experienced a loss of feedwater transient which led to a turbine trip and later a reactor trip.
Subsequently, a series of events took place that resulted in significant damage to portions of the reactor core.
It is believed that the sequence of events which led to core damage involved equipment malfunctions, design related problems and operational errors that, to varying degrees, all contributed to the incident consequences. Because plant conditions were substantially degraded, improvised operating modes for post-accident recovery were required.
On April 2,1979, while post accident recovery operations were taking place at THI-2, a task group was appointed to perform a generic assessment of feedwater transients in Babcock and Wilcox (B&W) plants in light of operating experiences,. including the TMI-2 accident, to determine bases for continued safe operation of these plants in both the short term and the long term. The charter for the group is as follows:
I
Given the operating experience with feedwater tran-sients in operating B&W designed r4 actors, assess whether reactor and plant systems at these plants provide adequate protection from design basis feed-water transients. This assessment should re-confirm Whether these pTant designs meet the requirements of NRC regulations, using appropriate staff guidelines for acceptable means of meeting these regulations.
This should include an evaluation of the safety margins of these plant designs to assure that spec-ified acceptd)1e fuel design limits are not exceeded as a result of feedwater transients.
1.1 Study Q)jective The initial focus of the study was on the following B&W designed plants for which utilities hold operating licenses:
Three Mile Island - Unit 1 Davis Besse - Unit 1 Crystal River. Unit 1 Oconee - Units 1, 2 & 3 Rancho Seco - Unit 1 Arkansas Nuclear One - Unit 1 The objective was to make an early assessment concerning those measures which might be necessary to prevent a recur-rence of the TMI-2 event at these facilities.
In particular, specific attention was giveq to the directives transmitted in Inspection and Enforcement Bulletins to utilities holding operating licenses for B&W plants to assure that the short term measures give adequate protection.
A second objective was to make an assessment concerning additional remedial measures of a short and long tenn nature which might be nec..:sary to correct design and operational deficiencies in B&W plants, including those which have not yet been licensed to operate. A third objective was added, namely to identify any weaknesses in the regulatory review process which may have contributed to failure to anticipate the sequence of events that led to degradation of core cooling in the early phases of the THI-2 accident.
1.2 Scope of Study The reassessment hero.in deals mainly with the generic impli-cations of the initiating eve it at TMI-2, i.e., the feedwater transient, or other initiating events of an anticipated nature which could lead to a similar transient in the reactor coolant system, e.g., an overpressure condition which opens pressurizer relief valves. Other aspects concerning the long term post-accident recovery sequence will be considered by another NRC task group that will deal with such matters as post-accident monitoring, operator actions, and emergency i
i pl an s.
j This reassessment involves the evaluation from the following standpoints:
1.
comparison of B&W plant design features 2.
plant operations 3.
licensing basis 4.
TMI-2 !aE Bulletins Item 1 is a comparison of the general design features includ-ing configurations, sizes, and safety and control systems o'f B&W operating plants to determine areas of uniformity and deviations. These are in turn related to plant characteris-tics which govern systems behavior under transient conditions.
Item 2 deals with event reports that have been reviewed where certain events of some similarity to THI-2 are dis-cussed in the interest of determining if we could or should have anticipated the TMI-2 event. The history of equipment malfunctions is reviewed. Operating procedures and operator training have also been reviewed in light of the TMI-2 event.
Item 3 treats the analyses as they are presented in the safety analysis reports and in response to specific questions.
The Standard Review Plan is discussed in terms of whether current licensing requirements would have required analysis of a TMI-2 type event.
The General Design Criteria and Technical Specifications are also considered; relative to the i
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-s-event.
Earlier studies, such as the React,or Safety Study' (WASH-1400) and subsequent Lewis Committee Comments are considered relative to any insight they may have provided.
Item 4 relates to the IAE Bulletin 79-05A addressed point by point. This bulletin provides a chronology of the event and identified areas for inr.ediate action by licensees to avoid' a recurrence. Near term action is focused in this area.
Consideration is also given to other PWR plant design feat-ures keyed to feedwater transients. This action provides insight into the generic applicability of the preliminary findings made on B&W plants as a result of the TMI-2 incident to Westinghouse and Combustion Engineering designed PWR plants.
The evaluation by the task group is presented as a set of findings and recommendations for further action in cach of principal areas investigated. These findings and recommend-ations will form the basis for more specific review on a long i
i term basis by the staff as well as with the reactor designers and utility license holders.
1 r
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e 1.3 Background Summary of the Three Mile Island Unit 2 Incident Although the current study is directed toward finding ways of preventing a recurrence of the TMI-2 incident the following summary of the entire sequence of events is summarized below to put the study into perspective and to emphasize the import-ance of ontrolling " anticipated" operational occurrences before plant conditions degrade to a point where core cooling capability is jeopardized.
At about 4 a.m. March 28, 1979, the Three Mile Island Unit 2 (THI-2) nuclear plant experienced a loss of feedwater which led to a tud>ine trip and a subsequent reacter trip on high pressure. Subsequently, a series of events took place that resulted in significant damage to portions of the nuclear core.
It is believed that these ensuing events involved equipment malfunctions, design failures, and operational errors that, to varying degrees, all contribut-ed to the incident consequences.
1 In the case of TMI-2 in the time period up to about 30 seconds, the sequence was generally normal and plant response j
was as expected.
The auxiliary feedwater system started-up and should have delivered secondary coolant to the plant's two steam generators to remove heat; however, the flow paths were l
blocked by closed valves. Auxiliary feedwater flow was established through operator action by opening the valves about eight minutes later.
In addition, the pressurizer relief valve should have closed as reactor pressure decreased; however, it failed to close.
As the reactor pressure reached a preset value (1600 psi),
the high pressure injection system (HPCIS) started as designed and began to inject cold water into the reactor. It was at this time period in the event that an indication of rapidly rising pressurizer level apparently led the plant operators to terminate the ECCS f1'ow. At this point the Three Mile Island l
incident had been underway for 11-12 minutes.
Between about 1 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the transient, the operators turned off the four large reactor coolant pumps which circu-late the reactor coolant through the reactor. This action was taken as a measure to prevent damage to the pumps.
It is following this action that we believe the severe damage to j
the nuclear fuel began. For the next several hours there was a very large temperature difference across the nuclear core indicating little flow of coolant through the core.
i
~
. i During this several hear period, when fuel damage was occur-
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ring, primary coolant from the reactor primary coolant system was being discharged to the reactor containment floor from flow out of the pressurizer relief valve and through the drain tank. Part of this coolant, which contained radioactivity,
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was automatically pumped from the reactor containment building I'
floor to tanks in the auxiliary building. The tanks overflowed
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permitting radioactivity to be vented from the auxiliary building. This discharge was secured in about 40 minutes.
The reactor containment was sealed (isolated) at about 9:00 a.m.
Through the afternoon and early evening of March 28, 1979, 1
the licensee isolated the stuck open PORY and tried to depres-s surize the reactor coolant system sufficiently to be able to
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turn on the residual heat removal system. Since this attempt i
failed, it was decided to repressurize the system.
After repressurization (about 8:00 p.m.); one of the main reactor coolant pumps in loop A was restarted and flow through l
the reactor core was established. Heat was being transferred i
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out of the reactor through the steam generator while using the I
condenser. The primary system was maintained at a pressure of 1000 psi and a temperature of 280*F.
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Reactor cooling has essentially been in this mode since that time. Efforts have mainly been devoted to maintaining,
this condition while a series of analyses have been conducted and while measurements have been taken to confirm a variety of parameters. These efforts have been directed toward preparing for the next steps in the cooldown process. A detailed chronology of the incident is given in Enclosure I
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v 11 B&W Plant Comparison _
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b General Configuration The reactor coolant system typically consists of the reactor vessel, J
two vertical once-through steam generators, four shaf t-sealed reactor 4
coolant pumps and one electrically heated pressurizer. The system is-1 arranged in two heat transfer loops, each with two reactor coolant 3 <
pumps and one steam generator. Figures II.1 and II.2 provide plan and elevation views of the primary reactor coolant system arrangement.
These are typical of all but one of the B&W plants currently operating. \\
Davis-Besse-1 is the first of a series of " raised loop" configurations.
s Its configuration is shown in Figure II 3 and II.4. The " raised loop" was introduced initially to improve post-LOCA characterics and to permit removal of the internals vent valves. The internals vent valves have been retained,though the number was reduced from eight to four. Other s
than an improvement in the already adequate natural circulation
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characteristics, B&W went ahead with the raised loop conf'iguration because of mechanical design and support configuration improvements.
Davis-Besse-1hasexperiencedeventssimilartokheTMI-2eventfrom lower power levels.As will be discussed later, the plant response was
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similar but saturation conditions were terminated by the operator before core damage occurred.
11.2 Comparison Table Key characteristics of the nine licensed B&W plants are listed in Table II.2.1.
The core thermal power ratings varied from 2452 MW to I
2772 MW although the core size and configuration are all essentially 1
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1 identical. The primary coolant system volumes are essentially the The pressurizer represents about 13 percent of the total same.
V system volume.
N The ;'ower operated relief valves on the pressurizer are set to relieve at 2255 psig. The valve for D-B-1 was mada by Crosby while I
those for the other plants were Dresser valves.
(The two code safety valves are selected to provide a 100 percent
, dargin in flow relieving capacity.
The high pressure injection pumps (HPI) were made by four different manufacturers. Other than D-B-1, each HPI would provide 450 gpm at 1600 psi (ECCS actuation on low primary system pressure).
D-B-1, uniquely, has separate make-up and high pressure injection pumps.
JheHPI while providing only 200 gpm, each at 1600 psia, would provide significantly more flow against lower back pressure. Their lower shut off head (4000 ft) would not lift the: power operated relief valves or pose af significant a repressurization problem following JCCS actuation.
, The once-through steam generators are all essentially the same. The
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i secondary side inventory is a function of po.ver level, but generally
'all operate within the same range. The column indicates the time to I boil ^off from high water level.
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Generally, the emergency feedwater system consists of one 100%
capacity steam turbine driven pump and two 50% capacity motor driven pumps where the capacity is related to the decay heat when actuated.
~M The column indicates same variation in design.
q During normal power operation, the water level in the once-through steam generator is maintained around 30 inches. With auxiliary feedwater,
'n it the level is controlled to around 180 inches to promote natural
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The reactor coolant pump trap is indicated where the RCP is raised
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relative to the cold leg centerline approaching the reactor.
Cany cane elevation refers to the rise of the hot leg to where it loops back down to the steam generator.
The internals vent valves are check valves inside the reactor vessel which would open to equalize pressure should the hot leg pressure exceed the cold leg pressure under post-LOCA conditions.
The high pressure reactor trip set point is 2355 psig on all plants.
A The high containment pressure trip is 4 psig for all plants (containment isolation).
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l There are no direct reactor trip signals generated by turbine trip.
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TABLE 1.
COMPARISON OF KEY CHARACTERISTICS OF UEEjnt1NG_B&'LPLAR11_ lit.LAl VE 40 itic L-p I
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II.3 Steady State Operation The PWR reactor coolant system (RCS) behavior during steady-state (constant pressure and temperature) operation principally involves flow and heat transfer mechanisms to maintain equilibrium. The heat cutpoint of the core is essentially balanced by the heat removed by the steam generators. The reactor coolant liquid volume (inventory) is maintained relatively constant by a small (compared to. total system volume) letdown / makeup flow rate. Normally, no mass (energy is removed from the system via the relief-safety valves. Thus, the predominant energy equilibrium between core heat given and steam heat removal maintains the reactor in thermal equilibrium, i.e.,
PRCS, TRCS = constants and pressurizer water level (which reflects the average temperature in the reactor coolant system) remain constant.
See Figure 11.3-1 II.4 1.oss of Feedwater Event (General)
In order to maintain the RCS energy level (P,T) within acceptable upper limits during a loss of normal heat sink (.0FW transient),
systems are incorporated to limit the rate of energy deposited in the system (reactor protection system) and to provide for adequate alternate /
backup paths for energy removal (relief-safety valves and auxiliary feedwater supply). The relief-safety valves involve both energy and I
mass removal while the auxiliary feedwater supply provides for only energy removal.
Since the relief valves remove RCS mass, a makeup 1
system (High Pressure Injection System) is provided to restore (RCS inventory) lost through the valves.
In addition, the HPI provides
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11.5 Interactions During -Gompicte-Loss of feedwater Event _
The behavior of the RCS during this event may be divided into two phases.
It is important to understand and distinguish between these two phases since the first phase is handled predominately. automatically by the NSSS while the latter phase must be handled by the automatic /
inherent aspects of the NSSS.together with the reactor _ operator.
In the firstphase, which occurs during the early (15 sec) part of the transient, more energy is being deposited in the system 'than is, being i
removed (see Figure 11.3-1). That is, during this phase the core continues to put energy into the system at a constant rate while the steam generator energy removal capability diminishes. This growing imbalance results in an increase in energy stored in the RCS. 'This is manifested by an increase in both pressure and temperature of the reactor coolant. The rising temperature of the primary coolant in turn results in its thennal expansion observed as a level swell in the pressurizer.
This rapid energy increase in the RCS builds until the energy input
[
of the core is essentially terminated (via reactor trip) together with the alternate / backup energy removal devices relief-safety valves operation and auxiliary feedwater system comes on. Thus, in essence, j
there is an initial period up to reactor trip where more energy is l
l being added to the system than is being removed (see Figure 11.5-1) which is immediately followed by a subsequent period b which more
energy removed from the RC system by the relief-safety valves and
~
auxiliary feedwater supply than is being put into the system by the shutdown core (see Figure II.5-2). The first phase is characterized by a pressure and temperature increase in the RCS resulting in a rapid pressurizer level swell. The second phase is characterized by a depressurization and cooldown of the RC water involving a rapid pressurizer level drop. Again, it is important to note that the inittal phase is accommodated by the inherent / automatic features of the NSSS while the latter phase involves a combination of. inherent /
automatic features as well as operator actions.
It should be further be noted that depending on the ability of the inherent / automatic aspects of the NSSS to handle the first phase and parts of the second. phase of the event, greater or lesser burden is put on the reactor operator,
to handle his responsibilities during the recovery cooldown-depres-surization phase. Thus, it may be stated that the ability to ultimately safely recover from this event depends on (1) the inherent / automatic aspects of the NSSS to present the operator with. a relatively stable controllable system still in a dynamic-state, and (2) the ability of the operator to correctly interpret and act upon conditions as they exist in the system. Put in a simple way, the operator should be tossed a small rock which he can see and is physically strong enough to catch without dropping. The weight of the rock depends on the plant design, his strength depends on the adequacy of h.'s training and experience, and his eyesight depends on his ability to be fed accurate infonnation in a timely manner.
1
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l From the above discussion a comparison can be made between different PWR plant designs to safely recover from a LOFW transient by examining the ability of,the inherent / automatic characteristics of the design to present the operator with a relatively mild dynamic state from which i
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II.6 Comparison of Loss of Main F.W. Event with Total Loss of Feedwater
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Event It should be noted again, that the operator does not have control over this heatup pressurization phase and that the severity of this phase can only contribute to the burden placed on:the operator during the second (cooldown recovery) phase. Since the first phase involves an increase in RCS energy, equipment which serves to lessen the heatup.
rate and pressurization should be examined.
The rate.at which the heat sink is lost is of prime importance. Plant designs which have a relatively large water inventory on the secondary side (CE & W) will result in a gradual loss of. heat sink when feedwater supply is lost.
This will result in a relatively gradual pressure and temperature increase in the primary system allowing time to establish auxiliary feedwater supply to prevent the pressure and temperature rise from lifting the relief valve setpoint. Thus, the steam generator inventory is ext emely
- i important in terms of precluding the need for a P32V or SV to lift.
Thus, plants with small steam generator inventory (B&W plants) will be more susceptible to lifting a relief safety valve during the event thereby inducing the possibility of a relief or safety valve failing to reseat. Thus plants with small steam generator inventory would tend to substantially increase the probability of providing the operator with an additional failure to contend with (in the form of a stuck open relief-safety valve) during the cooldown/depressurization recovery phase.
Secondly, plant designs which incorporate an anticipatory (direct) reactor trip for loss of heat sink conditions on the secondary side (e.g. low steam generator water level, low feedwater pump discharge pressure) could help prevent the imbalance between energy added to the system and energy removed from the system from growing so large as to cause PORY actuation. NSSS designs which rely on indirect reactor trips (e.g. high reactor system pressure) will always result in PORY actuation since the high pressure trip set point is above the PORY actuation set point.
The simmer margin of the PORV relative to nonnal operating pressure is also of significance for those plants with anticipatory (direct) i reactor trips.
If there is enough simmer margin in conjunction with a large steam generator water inventory and direct anticipatory reactor trip feature, the liklihood of lifting the PORV can diminish significantly. This is because the resulting gradual pressure and
. temperature in 'he PCS rise can be terminated by a reactor trip before the PORV set po'... is reached.
Finally, since the initiation setpoint time to reach full flow and capacity of the auxiliary feedwater system is geared toward removing decay heat from the system after the reactor trip it does not play an important roll in the initial heat-up/ pressurization phase of the
~
event.
,w!
Should auxiliary feedwater supply not be available (as was the case C) n
/
at7MI-2) $ the ' pressure peak will be somewhat higher and wil and close! o remove energy from the system whenever system pressure is t
above the valve reseat set point.
II.7 Comparison of Designs of Operating B&W Plants in Regard to Reactor Coolant System Dynamics The operating B&W plants have been compared in those areas of their design which play a significant role in the thermal and hydraulic behavior of the RCS system during and subsequent to LOFW transient.
Comparison values for various parameters are presented in Table II.1 The significant characteristic of the B&W plant design relative to the severity of the heatup-pressurization phase of the transient is the relatively small water inventory in the steam generators during power operation.
Rapid boiloff of this inventory (typically in less -
than 30 seconds at HFP) results in a rapid loss of normal RCS heat sink resulting in relatively rapid and severe RCS pressurization in the first few seconds of the transient. For all 0TSG B&W plants, it would be i
expected that the power operated relief valves would lift at their 2255 psi set point prior to a indirect reactor trip on high RCS pressure, which would occur at 2355 psi.
With regard to safety valve actuation for a B&W plant, although the FSAR would predict that they would lift (since calculated RCS transient pressure exceeds the set point) this would not be expected in actuality -
[1 s
TMI-2 LOFW event sequence. The difference lies in the. conservatisms incorporated in the FSAR analysis winich include pessimistic equipment time delays, scram effectiveness, BOL reactor kinetics, steam generator dry out models, etc.
Thus, one would not be expected to be confronted with a stuck open safety valve during the cooldown depressurization part of the transient, even for a B&W plant, since the valves would not be f
- 1) expected to lift in the first place.
In summary, therefore, the l
combination of the relatively small steam generator water inventory f
(inertia), 2) lack of a direct reactor trip on secondary side conditions (e.g. low steam generator water level, low feedwater pump discharge
'l pressure, and 3) relatively low PORV simmer margin would result in G
PORY actuation everytime a loss of feedwater event occurred, based on the automatic / inherent characteristics of the system. Some plants operating procedures (e.g. Oconee-1, 2, 3) require a " soft-wired" operator action' to' trip reactor immediately in a loss of feedwater. This action, if fast enough, could prevent PORV actuation. However, based only on the automatic / inherent characteristics of the B&W design, all B&W plants would be expected to actuate the PORV.
~
For a normal foss of feedwater event at a B&W plant in which the relief valve reseats, pressure-temperature conditions would not be expected to reach saturation conditions. Thus, voiding in the RCS f
would not be expected to present the operator with difficulty during the recovery.cooldown phase. However, whenever the PORV lifts, does not reseat and is not manually isolated, the resulting cooldown-depres-surization would probably reach saturation conditions for all B&W plants.
1
This is because the total HPI injection capacity and head characteristics is not sufficient to overcome the combined depres-surizing effects of the energy lost through the PORV and the heat lost through the steam generators with auxiliary feedwater available. This situation can be observed from the 9/24/77 Davis-Besse incident. Even if the steam generators are not available to remove energy from the system such as was the case at TMI-1, saturation conditions will be achieved whenever the PORV valve remains stuck open, because of the limited capacity HPI is incapable of maintaining system pressure above uturation pressure. Thus, for a B&W plant, during the RCS cooldown phase of a LOFW event a PORV sticking open generally always invariably presents the operator a RCS at saturation condition whether or not auxiliary feedwater is available. This is perhaps also true of W and CE plants but the likelihood of a valve sticking open is substantially less (see discussion of the LOFW event for CE and )[
plants). Since the plant will generally invariably reach saturation conditions even with HPI on continuously, the areas of concern become the effects of voids in the system relative to misleading pressurizer level readings, the potential for losing natural or forced circulation due to large bubble formation and frequent adverse operator actions based on his training. These concerns are true of most B&W plants.
Voiding in the system as it progresses due to saturation conditions being achieved in more parts of the system when pressure drops will always result in a pressurizer level swell. This was observed from the Davis-Besse incident as well as the TMI-2 incident. Thus, the combination
of mistakenly high pressurizer level and temperatures at or'near i
<aturatinn ennditions will frequently result in the operator turning offHPI(topreventblowingwateroutthepressurereliefvalves)and possibly shutting down one or more of the reactor coolant pumps (to prevent cavitation) because of his training.
In essence, evary time a LOFW event would occur with a stuck open PORV, the operator will be thrown a heavy rock (P+Psat) which he frequently cannot see very well (misleading prepressurizer level) and is not strong enough to catch (trainedactionswhicharecounterproductive). Since the likelihood
.of a stuck open valve would appear to be fairly high for loss of heat sink events for B&W plants, this difficult,,if not impossible, situation for the operator would be expected to occur quite often were no other actions taken.
Additional design aspects of the B&W system which would tend to compound the difficulty during recevery from such an event is the potential for forming large bubbles in key locations of the RCS 1
I which would inhibit either forced or natural circulation capability.
These areas are the peak elevation location of the hot leg riser going into the once-through steam generator (" candy cane") and the peak elevation of the pump impeller location. These are both potential vapor trap locations which could inhibit circulation - either forced or i
i natural-through the RCS, thereby possibly losing energy transfer capability out of the system via the steam generators.
This woul'd require that the shutdown decay heat energy generated in the core be
l.
i 1
I removed continuously through the relief valves (or safety valves if the relief valves are isolated). All B&W reactor coolant systems are virtually identical in these layout aspects, except for the Davis-Besse plant which has raised loops resulting in greater elevational pressure drop at the height of the " candy cane." Thus, Davis-Besse would be expected to exhibit a somewhat worse potential for forming vapor pockets at this point in the primary loop where pressure is low.
Finally, relative to the clarity of the operator's vision of the' actual condition of the reactor coolant system (i.e. whether or not voids are in the core or other parts of the system) all pWR designs will result in a characteristic level swell whenever voids form. This phenomena is not a function of pressurizer surge line layout.
In any s
NSSS design there will be an insurge and level swell in the pressurizer-whenever voids form elsewhere in the system. Thus, misleading infor-mation about total liquid water volume will be. transmitted to the operator whenever voids form - regardless of pWR NSSS design. Be that as it may, if sufficient voiding occurs so that bnly vapor interfaces with the pressurizer line hot leg junction, the B&W design will continue to provide misleading information to the operator about total liquid water content of the RCS while CE and W plants will not. This is because of the loop seal (manometer) piping layout of the B&W surge line prevents water from dropping out of the pressurizer when sufficient vapor pressure exists at the hot leg - surge line interface. On the other hand, the " straight down" biping layout of the surge line of the 1
CE and W plants would result in a level drop in the pressurizer whenever l
steam exists at the surge line interface. This feature woul'd prevent the B&W operator from ever knowing from pressurizer level.whether or.
not he had low RCS inventory (as long as vapor pressure were high enough in the reactor coolant system piping). On the other hand, pressurizer level would drop rapidly in the W and CE designs as soon as a vapor interface occurred at the surge line exit, regardless of pressure.
In summary, therefore, all of the currently operating B&W once-through steam generator reactors would be expected to behave very similarly to a loss of feedwater transient given credit for all systems performing as designed. A single failure of a relief valve failing to reset would also have similar effects with any differences predominantly attributable to HPI flow capabilities. The plants with the higher HPI capacity might be able to avoid saturation conditions but an analysis would have to be performed to show this. The major differences in plant dynamics recovery could depend on operator actions during this phase.
l I
I
4 A comparison of the main and auxiliary feedwater systems is provided for three operating B&W plants--Oconce, Crystal River, and Rancho Seco.
Main feedwater: Common to all three units (1) 2 steam turbine-driven pumps with the same flow and pressure ratings; (2) full flow condensate demineralizers and each has a bypass.
There are differences in the operation of the bypass: Oconee--air-operated, auto-open on Hi AP, fail open on loss of air; Crystal River--air-operated, auto open on'high AP, fail open or closed (air to both sides of diaphragm); Rancho Seco--motor-operated (local operation).
Auxiliary feedwater: Oconee One steam-driven centrifugal pump per unit, suction from three sources not designed to seismic Category I.
Auto start on loss of both main FW pumps (detected by discharge pressure below 750 psig or FW turbine stopvalveposition). Auxiliary service water pump (3000 gpm at 75 psig) from Class IE bus, one pump for all three units.
4 Failure mode on loss of air--switches to 14" main feed ring.
Failure mode on loss of power--valves and solenoids powered by batteries I
(non-lE).
There is no auto feedwater isolation.
Auxiliary feedwater: Crystal River Two centrifugal pumps, one motor (Class IE), and one s, team turbine-driven, suction from three sources (not desigr.ed to seismic Category I). Auto
?
. t start on loss of both main FW pumps as detected by' low control oil pressure; will start turbine-driven pump if the motor-driven AFW pump is not running.
Failure mode on loss of air--valves fail as is (air accumulators at valve -
for three cycles).
Failure mode on loss of, power--(NA Emergency Busses).
~
Stei,m li,ne failure matrix isolates all,FW to SG.
Auxiliary feedwater:
Rancho Seco Two centrifugal pumps, one motor-driven Class IE and one motor and turbine tandem (motor Class IE), suction from three sources only one is seismic Category I.
Auto start on loss of both FW pumps below 850 psig or loss of all RCps as detected by pc.ver moniotr (voltage, current, and phase).
Turbine-driven pump starts on SFAS signal.
Failure mode on loss of air--FCV fails open on loss of air.
1 Fail Jre mode on loss of power--FCV fails to 50% position. There is a Class IE MOV bypass around FCV on SFAS.
Steam line failure matrix does not isolate aux, FW.
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UNITED ST ATES NUCLEAR REGULA10RY COMf91tSSIOrJ bl~
W ASHINGTON, D,
- c. 20555 007 1 61975 Docket No. 50-320 H. Silver, Project Manager, Light Water Reactors Branch #2-2 THREE MILE ISLAND 2 SER INPUT In accordance with your request dated October 10, 1975, I have,
reviewed Amendments up through #33 of the Three Mile Island No. 2 FSAR.
Several outstanding items-remain to be resolved before we could issue The a satisfactory SER for chapter 14, Initial Tests and Operations.
outstan' ding items are identified in the minutes for the September 12, 1975 meeting we held with applicant to discuss these matters.
W Robert J(McDermott Quality Assurance Branch Division of Reactor Licensing cc:
D. Skovholt K. Knici to 9
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