ML20136B359
| ML20136B359 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/28/1997 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2442, GL-96-06, GL-96-6, NUDOCS 9703100246 | |
| Download: ML20136B359 (23) | |
Text
.-
e.
~
c.*
l Y
CENTEREO3 ENERGY 4
s 5501 N. State Route 2 419-249 2300 John K. Wood Oak Harbor,OH 43449 FAX: 419-321-8337 Vce Presusent Nuclear
[
Davis-Besse j
Docket Number 50-346 License Number NPF-3 i
Serial Number 2442 February 28, 1997 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C.
20555-0001
Subject:
Response to NRC Generic Letter 96-06:
" Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions" Ladies and Gentlemen:
On September 30, 1996, the Nuclear Regulatory Commission (NRC) issued GL 96-06 (Toledo Edison letter Log Number 4919).
That letter requested licensees, such as those for the Davis-Besse Nuclear Power Station (DBNPS)
Unit Number 1, to respond within 120 days and to address the following generic issues:
(1) Cooling water systems serving the containment air coolers (CACs) may be exposed to the hydrodynamic effects of water hammer during either a loss-of-coolant accident (LOCA) or a main steam line break (MSLB). These cooling water systems were not designed to withstand the hydrodynamic' effects of water hammer and corrective actions may be needed to satisfy system design and operability requirements. Licensees are to determine if their plant's CACs cooling water systems are susceptible to waterhammer during postulated accident conditions.
(2) Cooling water systems serving the containment air coolers may experience two-phase flow conditions during postulated LOCA and MSLB scenarios. The heat removal assumptions for design-basis accident scenarios were based on single-phase flow conditions.
Corrective actions may be needed to satisfy system design and Iq/
plant'sCACscoolingwatersystemsaresusceptibletotwo-phased)k
]
operability requirements. Licensees are to determine if their flow conditions during postulated accident conditions.
(3) Thermally induced overpressurization of isolated, water-filled
\\
piping sections in containment could:
- 1) jeopardize the ability of accident-mitigating systems to perform their safety functions 9703100246 970228-lgggggggg
]
PDR ADOCK 05000346 p
PDR 3
. - ~.. ~ - - - -.
_.. - - ~..
=.
Docket Numbar 50-346 a.
Licensa Numbar NPF-3 Serial Number 2442
{.
Page 2 and, 2) could also lead to a breach of containment integrity via i
bypass leakage. Corrective actions may be needed to satisfy
]
system operability requirements. Licensees are to determine if the piping systems which penetrate their plant's containment are f
susceptible to thermal expansion of fluid so that overpressuriza-tion of piping could occur.
Generic Letter 96-06 required that a written summary report be submitted to l
the NRC stating:
(1) the actions taken in response to the requested j
. actions noted above, (2) the conclusions that were reached relative to susceptibility for waterhammer and two-phase flow in the CAC cooling water i
systems and overpressurization of piping that penetrates containment, (3) the basis for continued operability of affected systems and components, and (4) the corrective actions that were implemented or are planned to be implemented.
On January 28, 1997, Toledo Edison provided by letter (Serial Number 2439) an interim response to GL 96-06.
The summary report describing the actions taken to date in response to GL 96-06 and the results of those actions is attached.
Should you have any questions or require additional information, please contact Mr. James L. Freels, Manager-Regulatory Affairs, at (419) 321-8466.
i 4
Very truly yours, Y
('1 L}
/
/ aj f
attachment I
1 cc:
A. B. Beach, Regional Administrator, NRC Region III A. G. Hansen, DB-1 NRC/NRR Project Manager S.
Stasek, DB-1 NRC Senior Resident inspector
)
Utility Radiological Safety Board i
a j
Dockat Numb 3r 50-346 License Numbar NPF-3 Sarial Numbar 2442 Attachment j
I
RESPONSE
TO NRC GENERIC LETTER 96-06 I
FOR ME N
l DAVIS-BESSE NUCLEAR POWER STATION i
i UNIT NUMBER 1 This letter is submitted pursuant to 10 CFR 50.54 (f). Attached is information pursuant to NRC Generic Letter 96-06, " Assurance of Equipment
}
Operability and Containment Integrity During Design - Basis Accident Conditions" for the Davis-Besse Nuclear Power Station, Unit Number 1.
1 For: John. K. Wood, Vice President - Nuclear By:
~
J
. Lash, D/vis-5 esse Plant' Manager Sworntoandsubscribedbeforemethis.28^dayc[b6b4Mh)9SI.
1" lA dh_.
Nota 7 Public, State of Ohio i
Ruth A. Anderson tote of Ohio February 28,1 j
... ~ _
i s
V' Docket Number 50-346 I
License Number NPF-3' L
Serial Number 2442
- Attachment Page1 i
i i
I h
DAVIS-BESSE NUCLEAR POWER STATION l
RESPONSE
TO L
GENERIC LETTER 96-06 l
-i NRC REOUEST I
Generic Letter 96-06: " Assurance of Equipment Operability and Containment Integrity During l
Design-Basis Accident Conditions" requested licensees to determine:
l (1)
If containment air cooler cooling water systems are susceptible to either water hammer or j
two-phase flow conditions during postulated accident conditions;
'l (2)
If piping systems that penetrate the containment are susceptible to thermal expansion of l
fluid so that overpressurization of piping could occur.
Generic Letter 96-06 also required that a written summary report be submitted to the NRC stating: (1) the actions taken in response to the requested actions noted above, (2) the conclusions that were reached relative to susceptibility for water hammer and two-phase flow in the Containment Air Cooler (CAC) cooling water systems and overpressurization of piping that penetrates containment, (3) the basis for continued operability of affected systems and components, and (4) the corrective actions that were implemented or are planned to be
' implemented. The following provides this requested information.
l l
t t
Docket Number 50-346 License Number NPF-3
, Serial Number 2442 Attachment Page 2 TOLEDO EDISON RESPONSE I
CONTAINMENT AIR COOLER WATER HAMMER OR TWO-PHASE FLOW A.
Background:
1.
Configuration The Davis-Besse Nuclear Power Station (DBNPS) containment (CTMT) consists of a large free-standing steel pressure vessel with an internal volume of approximately 2,800,000 cubic feet. The containment vessel is surrounded by a concrete shield building with which it has no supporting connections. Three essential powered, safety-related CACs are provided (see attached schematic). One of the CACs is an installed spare, while each of the others is aligned to one of the two open loop, safety-related Service Water (SW) trains. The same coolers are used for both normal and post accident cooling loads. For normal operation, the CAC fans run in high speed (approximately 1200 rpm). The fans are automatically switched / started to low speed (approximately 600 rpm) for specific accident conditions to allow for increased CTMT atmosphere density. The CACs have 90/10 copper / nickel cooling coils and are manufactured by American Air Filter Company. The CACs have a design heat removal l
from containment (at peak accident conditions)in excess of 70 MBTU/hr.
All three CACs (including the one spare in standby) are located next to each other on the 585' elevation in CTMT (see attached elevations diagram). The cooling water for the CACs is supplied by an open loop SW system using Lake Erie as its source. Supply and return piping for each CAC train is routed similar to that of the redundant trains. High l
points on the supply and return piping reach the 614' elevation inside i
CTMT. The SW return can be routed through any of several paths. The lowest elevation paths are the intake canal or Lake Erie. The intake canal and Lake Erie are normally interconnected and at the same level.
Although normal lake levels are above 570', with variation of a few feet due to wind, the DBNPS Operating License Technical Specifications
{
allow plant operation down to a level of approximately 562'. Under these conditions, approximately 20 feet of water column separation in the SW piping could result.
Docket Number 50-346 l
License Number NPF-3 Serial Number 2442 l
Attachment l
Page 3 When temperature conditions permit, SW is retumed to the circulating water Canal at approximately the 581' elevation. When the discharge is l
aligned to the circulating water canal, atmospheric pressure is sufficient to support a water column to approximately the highest point of the CAC piping. Therefore, column separation, ifit occurs, will be minimal under these conditions. In the event of an earthquake accompanied by collapse i
of the non-seismic return piping to the circulating water canal, the SW return is automatically routed to the intake forebay. Two SW pumps are l
normally in operation, one aligned to supply primary auxiliaries and one CAC, the other aligned to provide cooling to secondary auxiliaries and the other CAC. Therefore, both SW trains are available without any delay, except if a loss of power should occur.
During design basis accident conditions, CTMT cooling is also provided by each of two CTMT spray systems. The cooling capacity of each CTMT spray train is comparable to that provided by each of the two CACs. Two trains of both CTMT spray and CACs are required to be operable in plant i
operational Modes 1 through 3 per the Technical Specifications, however, only one train of CTMT spray and one CAC are credited in the design analysis.
2.
LBLOCA Resnonse A large break loss of coolant accident (LBLOCA) is the bounding CTMT temperature / pressure transient. Main Steam Line Breaks (MSLBs) are considerably less severe, in part due to the limited inventory available in i
once-through-steam generators. During the LBLOCA, a concurrent loss of offsite power is assumed, with assumed single failure of one emergency diesel generator. This results in the loss of one train of most safety related loads, including one train of emergency core cooling system (ECCS) pumps, one CTMT spray train, and one CAC. Under the DBNPS design basis, a seismic event is not postulated concurrent with a LBLOCA, however, LBLOCA loadings are combined with seismic loads in the CTMT and Reactor Coolant System structural analyses.
l The ECCS and the CTMT spray pumps take suction on the Borated Water i
Storage Tank (BWST) until it is depleted. Following its depletion, suction is taken on water collected in the CTMT sump. Assuming worst case initial temperatures for cooling water and CTMT heat sinks, CTMT conditions peak at approximately 35 psig and 255 F in approximately 60 j
seconds. Neither CTMT spray nor CACs are capable of mitigating the
Dockct Numbcr 50-346 License Number NPF-3 Serial Number 2442
' Attachment 5
i Page 4 i
initial temperature / pressure spike, since it is dictated primarily by the CTMT volume and RCS blowdown characteristics. After 10 seconds, as blowdown decreases, containment passive heat sinks begin to become effective, followed by actuation of one of the CACs and one of the CTMT spray pumps.
In the existing CTMT analysis, no cooling from the CAC is assumed to be available for 45 seconds. Containment cooling, provided in varying degrees by passive heat sinks, CACs, and CTMT spray, is effective after approximately 1000 seconds, when CTMT temperature and pressure begin to sharply decrease. With Se limiting assumptions of the CTMT analysis, CTMT sump temperature remains nearly saturated for a long period.
Therefore, following depletion of the relatively cool BWST and initiation of sump recirculation, energy is added to the CTMT atmosphere by the CTMT spray system. This results in a secondary repressurization of CTMT, which is bounded by the initial pressurization transient. If no cooling is provided by CACs, the secondary peak will be higher than if cooling from CACs is available, but CTMT integrity would not be challenged since the secondary peak will remain lower than the initial peak. (However, long term equipment environmental qualifications inside CTMT could be challenged.) If actual initial heat sink and cooling water temperatures are assumed, CTMT sump temperatures would be lower and the repressurization may not occur.
3.
CAC Ooeration During Loss of Offsite Power (LOOP) Transients Following a LOOP, the CAC fans will be de-energized. A seismic event and LOOP are assumed to occur concurrently. Assuming a single failure of an EDG, one CAC fan will be re-energized within 10 seconds when the output breaker of the available EDG closes. During the interim period, the fan will begin to coast down from its high speed rpm. During LOCA l
conditions, the fan is expected to coast down to approximately the low
)
speed setting by the time of restart. Therefore, air flow would not be i
interrupted.
With loss of power, the SW pumps will also begin to coast down. Recent testing indicates that the coast down time of the SW pumps is approximately 5 seconds. In the case of a LOOP without LOCA, the SW pumps will be restarted by a separate relay with a 40 second delay. In the case of a LOCA, the SW pumps will be started by the EDG load
_ _ _ _ _ _ ~ _ _ _ _ _. _ _ _ _
(
i V
Docket Number 50-346 License Number NPF-3 Serial Number 2442 Attachment
. Page 5 -
5
}
sequencer. Since the SW pumps are not restarted until step four of the s
l EDG load sequencer (20 seconds after EDG output breaker closure),
{
forced SW flow will be interrupted for approximately 25 seconds of the j
l LBLOCA. If no LOOP occurs, the SW pumps should not trip and would 1
I remain running.
i While a SW pump is de-energized during a LOOP, column separation in i
the CAC inlet and outlet piping high points (614' elevation) will occur for l.
all CAC trains if the SW return is aligned to Lake Erie. In the unlikely event that the lake is at the minimum level of 562', the water will fall in 4
both inlet and outlet piping by approximately 20 feet. The CACs are j
located at the low point of a loop seal between the inlet and outlet high points. Therefore, water will not gravity drain out of the CAC cooling coils. However, if a LBLOCA accompanies the LOOP, water is available within the CAC coils to be heated and, potentially, to boil. Boiling will occur in the CAC coils well before restoration of forced SW flow for design basis LBLOCAs. Thus, the Generic Letter 96-06 issues related to CAC boiling are applicable to the DBNPS.
B.
Water Hammer Effects:
Water hammers can be classified in several manners depending on system configuration, fluid conditions, and fill velocities. A compilation of seven major types of water-hammer mechanisms is provided in EPRI document NP-6766, " Water Hammer Prevention, Mitigation, and Accommodation."
Essentially, water hammers are driven either by rapid condensation of steam
. voids or by pumped liquid velocity. In either case, propelled liquid columns or slugs eventually impact stationary objects such as closed (or throttled) valves or another liquid column. In cases where the maximum condensation rate is high, the potential for large water hammers is particularly high. High condensation rates are promoted by large temperature differences between liquid and vapor phases, large contact surface area, and especially by high purity steam (i.e., little non-condensable content). In some closed loop systems, especially where j
significant heat input causes the potential for all of the above conditions, j
particularly severe water hammers can occur. Unlike the open loop SW system which serves the CACs at the DBNPS, some closed loop systems may have very little non-condensable gas content.
During a LOOP without a LOCA, water column separation will begin within a few seconds as the SW pumps coast down and pressure in the CAC piping high points decreases toward the vapor pressure. The separation results in a low i
s
,--aq-s
i' i
F Docket Number 50-346 I
License Number NPF-3
[
, SerialNumber2442
+
Attachment j
Page 6 i
i i
i pressure void formation in both vertical and horizontal CAC piping high points.
t In the CAC piping following pump restart, the condensation surface area will be j
small because the fill velocity will be sufficiently high (Froude number >l) to l
maintain a filled pipe, even in horizontal sections. Temperature differences L
tend to be relatively small and non-condensable gas will be present, evolved from the SW in the low pressure regions, to reduce the heat exchange between phases, even in the vertical downward sections. Investigations by Fauske and j
Associates, Incorporated (FAI) (an engineering finn hired by Toledo Edison) i indicate that the liquid column will maintain essentially a plug flow profile.
Therefore, there would be little fluid acceleration due to condensation effects L
and the fluid velocity will not be especially high.
Water hammer can also occur when voids pass through throttled valves. The j
fluid accelerates as gas voids are passed, and then abruptly decelerates as liquid 1
impacts the throttle point. Since the CAC system does not have throttled valves i
during LOCA conditions, this phenomenon will not occur. In the absence of a
. LOCA, the smaller, very low pressure voids will tend to collapse prior to j
reaching the throttle valves. Thus, peak pressures would be those associated
' with column rejoining (at the pumped fluid velocity), as the moving fluid impacts stationary fluid.
j The resultant water hammer pressure is a direct function of the impact velocity and the sonic velocity of the liquid. Unlike closed loop systems, open SW systems always have considerable quantities of non-condensable gases in l
solution. In fact, the gas content would typically be near the solubility limit of i
air in water at the suction temperature. For a typical column separation event, I
the pressure is reduced approximately to that of the vapor pressure of the liquid.
i For the CAC cooling coils in normal operation, this would correspond to a j
pressure ofless than 2 psia for either inlet or outlet. At this pressure, a significant portion of the non-condensable gas would leave the solution. This i
amount of non-condensable gas would reduce sonic velocity by an order of magnitude. Therefore, the presence of non-condensable gas has a strong beneficial effect for mitigating column rejoining type water hammers. Based j
on preliminary results, the maximum pressures appear well within the piping capacity.
During a LOOP plus a LOCA, initially the event would proceed with column separation as described previously. After the CAC cooling coils reach the boiling point, steam production will drive hot water and steam out of the CAC coils. This heated water and steam will be somewhat cooled as it displaces and mixes with less heated liquid in the vertical legs near the CAC inlet and outlet.
1
Docket Number 50-346 l
License Number NPF-3 Serial Number 2442 l
- Attachment l
Page 7 l
As the steam generation progresses, the pressure in the piping which was voided by column separation will increase and the amount of voided pipe will increase.
l During this process, a significant additional amount of non-condensable gas in L
the heated liquid will come out of solation. Following the EDG and load sequencer delay, a SW pump will start in approximately 30 seconds. At this i
point, as forced SW flow resumes, the pumped water velocity will result in full I
pipe flow, as for a LOOP without LOCA, so that condensation-induced water l
hammer is not expected to be significant. A greater amount of non-condensables a
will be evolved than for a LOOP without a LOCA. Also, the initial expulsion of i
heated liquid from the CACs will reduce the interphase temperature difference, i
which will reduce the condensation rate after SW pump restarts. As column j
l rejoining occurs, the added steam is expected to be compressed. Without very high condensation rates, this leads to a cushioning effect, and could result in lower peak pressures for the LOOP-plus-LOCA event than for LOOP alone.
However, since the total void size in the LOOP-plus-LOCA event is larger, the bounding event is not readily apparent.
j FAI has conducted a number of scaled tests featunng similar geometries and fluid fill velocities. These tests indicate that the above descriptions of the phenomena are correct. They also indicate that water-hammer pressures on the order of tens rather than hundreds of psi should be expected. These qualitative assessments are supported by preliminary quantitative results.
As described above, the anticipated peak pressures will not result in piping system failures. However, the motion of fluid slugs and voids in the piping will result in dynamic shaking of the piping in addition to pressure spikes. The magnitude of water hammer-induced and slug-induced dynamic loads is being calculated by FAI using their proprietary computer code TREMOLO.
TREMOLO is a transient, two-phase, node and junction, thermo-hydraulic code.
It was originally developed by FAI to analyze two-phase pressure transients resulting from closure of motor operated valves for the Generic Letter 89-10 program. With the addition of mechanistic heat transfer calculations to simulate a fan-cooler, it has been used to simulate two-phase flow in CAC systems.
TREMOLO was developed under the FAI quality assurance program which meets 10 CFR 50 Appendix B requirements. Output from the TREMOLO l
code will include pressures, velocities, and reaction forces versus time for the J
CAC piping. Preliminary output for CAC No.1 from TREMOLO has been l
input to the structural code ME101. ME101 is a Bechtel piping structural l
response code, also developed under 10 CFR 50 Appendix B.
I l
\\
l l.
c Docket Number 50-346 License Number NPF-3
., Serial Number 2442 Attachment Page 8 C.
Two-Phase Flow in CAC Piping:
_In the bounding design basis LBLOCA conditions, the outlet water temperature of the CACs is subcooled at all times, with substantial margin to boiling conditions at design heat removal and flow rates. Due to the large CTMT volume at the DBNPS (>2,800,000 ff), the peak CTMT pressure and temperature are relatively low. When SW pumps are running, the available inlet pressure to the CACs is greater than the saturation pressure in CTMT.
Therefore, the CACs should not be capable of vapor binding against SW pump discharge pressure. Following post-LOOP-plus-LOCA restart, the SW pumps will be capable of suppressing further boiling and sweeping any voids out of the CACs. Because the CAC effectiveness is low early in the transient compared to other heat removal mechanisms and the CACs are not credited in the CTMT analysis for 45 seconds, any reduction in initial effectiveness is inconsequential.
During normal operation, discharge valves in the CAC outlet lines may be throttled if desired for CTMT temperature control. However, accidents which involve significant containment heating, such as steam leaks or LOCAs, will result in automatic opening of the outlet throttle valves. Therefore, when significant containment heatup has occurred, no throttling points exist in the CAC system where prolonged flashing can occur.
Therefore, the CACs are not susceptible to two-phase flow concems.
D.
Responses to NRC Items-1.
Actions Taken:
As described above, the Service Water System (SWS) is an open-loop system which provides cooling water to the DBNPS CACs. The CAC cooling water system was evaluated for susceptibility to either water hammer or two-phase flow during postulated accident conditions. This effort'aas been conducted by Toledo Edison staff and their consultants, Fauske and Associates,Inc. (FAI).
FAI has performed small-scale modeling experiments to study the water-hammer effects which may occur for geometries similar to that which exist at the DBNPS. A qualitative evaluation of these water hammer effects has been I
performed utilizing those experiments and DBNPS-specific parameters. The results of these experiments confirm the peak water hammer pressures will be low, and provide a firm basis for the assumptions of the follow-on plant specific quantitative evaluation. Quantitative evaluations for water hammer
Docket Number 50-346 License Number NPF-3 l
Serial Number 2442 L
' Attachment l
Page 9 t
i i
. hydraulic loadings have been performed for CAC Train No. I utilizing FAI's hydraulic code TREMOLO. Output from this code has been input into the i
piping structural dynamics code ME101 by the DBNPS staff. Preliminary results are described below. Similar analysis is in progress for CAC Trams No. 2 and 3.
2.
Conclusions Relative to Suscentibility for Water Hammer or Two-Phaw Flow:
The DBNPS CACs are of the open loop design and are susceptible to the water hammer effects described in GL 96-06. Two modes of voiding may l
occur, column separation and voiding due to heat input at the CACs. Column separation in the CAC inlet and outlet piping high points (614' elevation) will occur for all CAC trains due to the de-energization of the service water l
pumps. For the minimum lake level of 562', the water will fall in both inlet and outlet piping by approximately 20 feet. Column separation will occur l
with or without heat addition. The CACs are located at the low point of a loop seal between the inlet and outlet high points. Therefore, water is available within the CAC coils to be heated and potentially to boil if a Large Break Loss of Coolant Accident (LBLOCA) accompanies the Loss of Offsite Power (LOOP). Boiling will occur in the CAC coils well before restoration of forced SW flow for design basis LBLOCA events.
Loadings created within the CAC inlet and outlet lines produced by water hammer in a LOOP or LBLOCA are currently under evaluation. The CAC Train No.1 inlet and outlet piping has had preliminary stress analysis performed to quantify piping stress levels, pipe support loads and CAC nozzle i
reactions. Piping stress levels have remained within the applicable American
)
Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code j
Section 111 stress allowables for both inlet and outlet piping. Pipe support reactions, in general, reflect loads which are within the design margins of the current design calculations. However, several supports have elevated loads beyond the current design loads and efforts are currently underway to evaluate j
these loads. Those efforts are expected to demonstrate the continued functionality of these supports. The CAC nozzle reactions remain to be evaluated, but based upon the other loadings within the piping systems, it is l
expected that the nozzles will be shown to be within their design basis.
l Analysis of CAC Trains No. 2 and No. 3 is in progress and is expected to produce results similar to CAC Train No.1. Thisjudgment is supported by
]
3 Docket Number 50-346 License Number NPF-3
. Serial Number 2442 i
Attachment Page 10 system configurations similar to CAC Train No. I and the sharing of some of 1
the piping between the CACs.
l In the bounding design basis LOCA conditions, the outlet water temperature of the CACs would be subcooled. During significant containment heating -
events, the temperature control valves will move to their full open position in order to achieve maximum cooling. No throttling points will exist under these conditions in the CAC system and no prolonged flashing will occur. Even if i
two-phase conditions should develop during a loss of power scenario, single-phase flow will be restored at approximately the time that the CACs are credited in the containment analysis. Due to the large containment volume at the DBNPS, the peak containment pressure and temperature are relatively low.
When the Service Water (SW) pumps are running, the available pressure to 1
the CACs is greater than the saturation pressure in containment after an
]
accident. Therefore, the CACs are not capable of vapor binding against SW l
l pump discharge pressure. Following restart, the SW pumps will be capable of suppressing further boiling and will sweep any voids out of the CACs.
Because the CAC effectiveness is low early in the transient compared to other l
heat removal mechanisms and the CACs are not credited in the containment i
l analysis for 45 seconds, any reduction in initial effectiveness is
]
inconsequential. Therefore, the DBNPS CACs are not susceptible to two-i phase flow concerns.
-l l
l 3.
Basis for Continued Operability of Affected Systems:
Fauske and Associates, Inc., has performed small-scale modeling experiments to study the water-hammer effects which may occur for geometries similar to that which exist at the DBNPS. A qualitative evaluation of these water hammer effects has been performed utilizing these experiments and DBNPS-specific parameters. The results of these experiments confirm the peak water hammer pressures will be low. Although qualitative analysis has been performed only for CAC Train No.1, the similarity between trains supports l
the judgment that loads within all trains will be acceptable pending j
completion ofcalculations.
Although analysis has been performed only for CAC Train No.1, the similarity between trains supports the judgment that loads within all trains will not adversely affect system operability. Based upon the previously discussed
(
results of the preliminary stress analysis, the affected components remain l
I
i n
i i
Docket Number 50-346 License Number NPF-3 Serial Number 2442 Attachment Page11 The CAC units are seismically qualified and mounted. Piping water hammer loads transmitted to the CAC units are expected to be within the CAC design limits. Dynamic loadings within the CAC units would be related to the magnitude and duration of resultant pressures, diameter and lengths of associated tubing, and support location and rigidity. The CAC cooler coil tubes are small diameter tubes, approximately 72"long. They are supported throughout their length by the cooler frame and cooling fins. Water hammer inertial forces within the tube bundles are judged to be low based on the diameter, tubing length, and developed pressures.
1 1
Following restart, after interruption of power, the SW pumps will be capable i
i-of suppressing further boiling and will sweep any voids out of the CACs.
I Because the CAC effectiveness is low early in the transient compared to other l
l heat removal mechanisms and the CACs are not credited in the containment analysis for 45 seconds, any reduction in initial effectiveness is inconsequential.
It is concluded that peak pressures and loadings due to water hammer effects will be low and the CAC system at the DBNPS is not susceptible to damage -
by two phase flow. Therefore, the CACs remain operable.
4.
Corrective Actions-1 Final detailed calculations to confirm the above preliminary results for all CAC trains will be completed and a letter submitted to the NRC with the final j
conclusion by July 30,1997, if any design modifications are required to i
restore full qualification of any components, they will be identified at that time and a proposed schedule for performing these activities will also be included in the letter submitted to the NRC.
11 OVERPRESSURI7 ATION OF PIPING THAT PENETRATES THE CONTAINMENT BACKGROUND A.
Backoround The piping penetrations at the DBNPS were designed under the ASME Code i
1971 Section III, Class 2 requirements. Piping penetrating CTMT was reviewed for thermal pressurization potential and for potential effects on accid..
i mitigating systems. Following main steam line breaks or LOCAs which pressurize CTMT, the Safety Features Actuation System (SFAS) provides signals to start the appropriate Emergency Core Cooling Systems (ECCS) and
)
)
- i. "
Docket Number 50-346
]
i License Number NPF-3 Serial Number 2442 j'
. Attachment Page 12 I
provides isolation signals to CTMT isolation valves which are not immediately l
needed to mitigate the accident. Several containment isolation valves need to be open to remove core heat or containment heat. All valves required for short-i term accident mitigation are either opened by SFAS or are maintained open at i
all times while in Modes 1 through 3. Valves associated with the ECCS i
injection pumps are located outside containment and would not be subject to 1
i thermal pressurization (even if mispositioned) since the downstream flow path i
is connected to the reactor coolant system via check valves. CTMT spray has no
{
potential for thermal pressurization since it has no isolation valves inside CTMT. CACs are operated with their inlet isolation valves maintained open.
Therefore, the thermal pressurization issue does not affect any systems needed for short-term accident mitigation.
l In the long term, manual actions may be required to provide a post LOCA boron dilution flow path to prevent concentration and precipitation of boric acid in the i
reactor core. The primary active flow path provided for this purpose is the decay heat removal drop line, which uses two motor operated gate valves _
located inside CTMT. The piping between these valves is already protected against thermal pressurization by provision of a small check valve which vents to the RCS. The isolation valves were modified to prevent thermal pressurization of the bonnets during the Tenth Refueling Outage (10RFO) as g
part of the GL 89-10 program. A secondary flow path which is maintained for
{
long term post-LOCA boron dilution is the pressurizer auxiliary spray line. Due to a small flow capacity, this line is effective only for LOCAs occurring from j
low initial power levels or for very long term use. However, one of the two j
isolation valves on this line (Penetration 74c) is subject to potential binding as a i
result of thermal pressurization of the line. Action (as described below) has been taken to assure that these valves operate. No other systems which are required to mitigate accidents are affected.
)
i A screening criteria identified that thirteen containment penetrations contain fluid which could be isolated and pressurized due to thermal expansion during accident conditions. Cold fluid lines entering CTMT from outside the auxiliary building 1
j-were assumed to have been isolated at 50 to 60 F. Fluid temperatures of lines internal to the auxiliary building were assumed to be approximately 95 to 100 F d
1 when isolated. The bounding, effective LBLOCA temperature is approximately 250 F, with a longer term secondary peak of approximately 242 F. Therefore, l
250 F was applied as an upper temperature for the section ofisolated piping inside CTMT for most of the penetrations. (242 F was used in some of the larger penetrations, where thermal lag could be easily credited.) Isolated sections of 4
piping outside CTMT were maintained at, or increased to 95 F. Potential
)
4 4
d m _.
i i T l
L V*
Docket Number 50-346 License Number NPF-3 i
l Serial Number 2442 l
l Attachment Page 13 l
convection currents of fluid in the penetration piping was accounted for by assigning CTMT temperature to the unheated piping in the 4.5 foot annular region l
l between the CTMT vessel and the shield building. The stress analysis credited thermal growth and clastic expansion of the piping, and utilized ASME steam table i
data for water expansion and compressibility.
B.
Resnonses to NRC Items 1.
Actions Taken:
Piping systems that penetrate the DBNPS containment were evaluated for susceptibility to thermal expansion of fluid that may cause overpressurization of the piping during postulated accident conditions. This evaluation was i
performed by: 1) identifying the potentially affected piping and the service i
conditions,2) determining the resultant pressure from a LOCA, 3) determining the resultant functionality of the piping and its components, and
- 4) determining compliance with the ASME Code stress allowables.
The DBNPS review of thermal overpressurization was not limited to just containment penetrations. An additional review of valves located inside l
CTMT that provide double isolation and are required to open to mitigate the consequences of a design basis accident was conducted. This review was used to determine if any safety significant piping sections could be subjected to thermal overpressurization. None were identified. The thermal pressurization of non-safety related piping inside containment and the effects of this pressurization on containment isolation were also considered. In addition, i
consideration was given as to whether a potential failure would divert flow from safety equipment and prevent the safety equipment from performing its required safety function. Based on these considerations, it was concluded that the overpressurization of these systems would not adversely affect any required system functions or containment integrity.
i 2.
Conclusions Relative to Overnressurintion of Piping:
Thirteen containment piping penetrations were identified as potentially susceptible to post-LOCA thermal overpressurization. They are listed in the following table.
i
4 s-Docket Number 50-346 License Number NPF-3 Serial Number 2442 I
Attachment Page 14 j
l i
Penetration Name Results 1
Pressurizer Sample Line Meets ASME Code Faulted Stress Allowables 3-Component Cooling Meets ASME Code Faulted Stress Allowables Water to Containment l
14 RCS Letdown Meets ASME Code Faulted Stress Allowables 56 Reactor Coolant Pump Meets ASME Code Faulted Stress Allowables i
Seal Retum i
13 Containment Normal Meets DBNPS Interim Stress Allowables l
L Sump l
47s Core Flood Tank Meets DBNPS Interim Stress Allowables Sample Line 74C Pressurizer Auxiliary Meets DBNPS Interim Stress Allowables.
l Spray Drained to ensure availability post accident.
4 Component Cooling Exceeds DBNPS Interim Stress Allowables.
Water Retum Soft seat butterfly valve. LLRT documented 1
leakage will mitigate pressure rise.
12 Component Cooling Exceeds DBNPS Interim Stress Allowables.
Water to the Control Limited Plastic Deformation will occur.
Rod Drive Mechanisms Modification Number 97-009, Supplement 0 i
being developed to prevent deformation.
'21 Demineralized Water to Exceeds DBNPS Interim Stress Allowables.
Containment AOV Globe Valve--Provides inherent relief.
32 Reactor Coolant Drain Exceeds DBNPS Interim Stress Allowables.
to Reactor Coolant AOV Diaphragm--Provides inherent relief.
Drain Tank '
Partially drained.
48 Pressurizer Quench Exceeds DBNPS Interim Stress Allowables.
Tank Outlet '
AOV Globe Valve--Provides inherent relief.
49 Refueling Canal Fill Exceeds DBNPS Interim Stress Allowables.
I Used only during refueling outages. Partially drained.
l 1
~
W Docket Number 50-346 License Number NPF-3 Serial Number 2442 Attaclunent Page 15 Four (1,3,14 and 56) of these thirteen piping penetrations have been determined to meet the American Society of Mechanical Engineers (ASME)
Boiler & Pressure Vessel Code allowable stress values and no further action is required to reduce stress levels at these penetrations.
Three (13,47a and 74c) of the nine remaining piping penetrations meet the DBNPS interim allowable stress values. This is acceptable in accordance with Generic Letter 91-18 until the ne.xt refueling outage when action will be required to ensure these piping penetrations meet ASME Code stress allowables.
The remaining six containment piping penetrations could exceed DBNPS interim allowable stress values under conservative assumptions, with no credit taken for pressure mitigating characteristics (e.g., seat or packing leaks that would be expected to develop at increased pressures). None of these piping penetrations support post-accident mitigating functions and each of the piping penetrations is operable as discussed below.
One of these six piping penetrations (Penetration Number 4 -
Component Cooling Water Outlet from Containment) has pre-existing, i
acceptable, measured leakage as determined from local leak rate tests performed during the Tenth Refueling Outage (10RFO) in accordance with 10 CFR 50, Appendix J. Since this valve is a soft-seated butterfly l
valve, the quantified leakage would reliably increase to even greater values during this event. This leakage will limit pressurization of the piping penetration to an acceptable level. The need for any further plant or procedure modifications will be evaluated.
Two of the six piping penetrations (Penetration Number 21-Demineralized Water and Penetration Number 48 - Pressurizer Quench Tank Outlet) have air-operated globe valves which are oriented such that pressurization of the penetration piping will be relieved by causing the disk to be lifted against the actuator spring force without causing valve or piping damage. One of the six penetrations (Penetration Number 32 - Reactor Coolant Drain to Reactor Coolant Drain Tank) is an air-actuated diaphragm valve which would be slightly displaced by the pressure increase, allowing adequate leakage to prevent overpressurization. The inherent relief i
l
~..
i Docket Number 50-346 License Number NPF-3
, SerialNumber2442 Attachment Page 16 l
capability of these valves is reliable and provides acceptable I
assurance of protection.
One of the six penetrations (Penetration Number 49-Refueling Canal Fill) has been partially drained. This piping is only used during refueling outages.
l The remaining penetration, (Penetration Number 12 - Component Cooling Water to Control Rod Drive Mechanisms) would likely develop packing leaks or valve seat leakage to limit the pressurization. However, without crediting j
this relief mechanism, the thermal expansion ofliquid in this penetration is projected to not result in more than approximately 2 percent plastic deformation under worst case conditions, and therefore, the affected i
penetration will remain operable. Modification Number 97-009 Supplement 0 will provide a design change to ensure that this penetration remains within ASME Code stress allowables.
3.
Basis for Continued Operability of Affected Systems:
i Worst case LOCA conditions were used in the operability evaluation.
Specifically, it was assumed that there would be no leakage from any valves in the isclated piping to mitigate the pressure rise. Calculated in this manner, for the six penetrationsjust discussed the pressure increases were determined
)
to be sufficient to cause pipe stresses beyond DBNPS interim stress I
allowables. The volumetric expansion of the water would result in piping deformation, but not piping failure. Piping strain only on the order of 2 j
percent is predicted, while strain on the order of 20 percent would be j
necessary to cause piping failure. This led to a determination that no required system function or containment integrity would be lost, even without actions i
(such as draining or consideration of valve leakage) to mitigate the pressure j
increase. It is concluded that these piping penetrations will remain operable and, therefore, no system or containment integrity will be lost.
4.
Corrective Actions:
Thirteen containment piping penetrations were identified as potentially susceptible to post-LOCA thermal overpressurizaticn. Four of these thirteen penetrations have been determined to meet the ASME Code faalted allowable stress values and no further action is required to reduce these stresses. The remaining nine penetrations have been dispositioned as discussed below, i
e j
Docket Number 50-346 i
License Number NPF-3 Serial Number 2442
)
Attachment Page 17 The Reactor Coolant System Drain to the Reactor Coolant Drain Tank
-J (Penetration 32) was isolated and the piping was verified to be partially l
drained on January 17,1997, to ensure the piping is protected from 1
overpressure concerns. This penetration is isolated by air operated l
valves, which preliminary evaluations have shown to provide acceptable inherent pressure relief. Modifications or procedure changes will be made if determined necessary.
l 4-On February 1,1997, approximately two gallons of water were drained j
from the Refueling Canal Fill (Penetration 49) piping. Analysis showed that its as-left LLRT leakage would have partially mitigated the i
3 overpressurization. Since this piping is only used during refueling j
outages, it was partially drained to provide an extra measure of r
l overpressurization protection.
1 j
A modification is being developed (Modification Number 97-009 I
j Supplement 0) for the Component Cooling Water Supply to the Control Rod Drive Mechanisms (Penetration 12). This modification will ensure protection of the penetration from overpressurization coacerns ad will be
)
installed during the next plant shutdown.of sufficient duration i
implementation, provided the engineering design work is complete and the j
necessary materials are available. This modification will be installed no l
later than the next refueling outage,11RFO, which is currently scheduled j
for April,1998.
On January 23,1997, water was drained from the Pressurizer Auxiliary Spray (Penetration 74C) piping. Although this penetration meets the DBNPS interim stress allowables, the penetration was partially drained
)
to prevent pressure binding of the penetration isolation valve. The
}
Pressurizer Auxiliary Spray line may be used for long term boron
[
dilution in certain LOCA scenarios. Therefore, action was taken to ensure that pressure binding of the isolation valves cannot occur. The
~
interim action taken was to ensure that valves DH2735 and DH2736 4
were closed and approximately one half gallon of water was drained a
from between them. This will eliminate any potential for significant j
pressure increases and challenge to the operability of this flowpath.
j The presence of this small void does not result in other concerns. The small amount of air in the system will not cause a corrosion problem 4
because the line is stainless steel. When the system is unisolated for return to service, there is no concern related to water hammer as a result of the void because of the small size of the void and the relatively low I
i 1
l e
Docket Number 50-346 License Number NPF-3 Serial Number 2442 Attachment '
l Page 18 pressures and velocities that would exist in this line when it is put into.
service. Non-condensable gases injected into the reactor coolant system l
are insignificant. A change being made to sppropriate procedures will i
ensure this configuration is maintained.
l The three penetrations (13,47A, and 74C) which meet interim L
allowable pipe stress values, will either be modified or will be protected from overpressure concerns by procedural changes. The two penetrations (21 and 48) with air-operated globe valves will have their relief pressures confirmed through examination of manufacturing data.
l Modifications or procedure changes will be made if determined l
necessary. These actions will be completed by the end of the next t
refueling outage (11RFO) which is scheduled to commence in April 1998.
Plans for final corrective actions for the piping penetrations will be j
provided in the letter submitted to the NRC by July 30,1997.
l l
l i-e l
+
-t@l 1
ll
\\
l LI I
q l
1 si
^ i Il I !
I N
l[- l.
I l l l
l ll l
II d p I
8 l s i elll1
- i @;
gl
~
1 l'l l
o l!!!
H.
i s;
'9, e +>l
,g j' f
q$g
,1 6
+
j
'9 q' +T ~ M %h'
~
- 2 f@,i h
1
_'#-I
.9,*, q91 W 9
a
[.
,4 g
J o
+
g
~
g a",
,,u:
-~? "
>a-LJg>
i h9i i '
e-4 j'
~..
l il ea-T1"]
g o
u.
+
i
)
p ip E
1 O-ba
- i, f
M i+
Q
- [>
_~j'D
~'
['
_ s '.
,),
+
j ouJg) g
<b
- 3 I'
9-
- 1
- '. t og
, g. -' I_ _ _ _ i As i,
Q ySj
'i g f; Ls,.
l-t yl $-t a
E ii!
i
- i i E2 ?u j
- t. - (::'
3
- , / l :_'*"w i a,
N 4'
- I L',
4 o
k1_g)
- W
]
='
qz
- g
+
3 I
Y I
h' Ir T
I i
sT I
ne H
g: l P
n
+:en:
I i
I j;
I l
il 4
l i
tll as as i [
j [
'8
+ ]:
l e
i
. I, i
i i
$7 f
U
- I
- [
.ai ' ai u
d'i#'i
- 'i
- l i t'
- i t
- '=
L u
6
'8 vn
=
~.
4 e
e-.- e-e_ ' e_
9,
- n g_
,y i
"' b i l
e_
j i
g ij i g4 l l[1
[@ [}
[@
[@-
~
i
.m u
tat i
4 l
1 i
ist.
i
- i ; w !
.4. n.
!*a5, J
i : i i
E2 n i
b, ii m
i, i
.g i s
g q
$$$i II T"h d
j et e43 gg;3g
.4 11;
,H 5
n.1 ni(
us.
ni, n
e3a#s t_______9 I
y.
yo Docket Number 50-346 4
License Number NPF-3 DAVIS-BESSE CAC ELEVATION Serial Number 2442 Attachment 7
Page 20 620'---~~-----------~~~---------------------------------------------------------------------------------------------------~~--------------------------------------~~~-------.e---~~~-------
614' 614*-6*
612'-6*
sir.........................................................................................
...................................u..t.........
,- s-7s-ps-pa-60s.....................................................................g.,.
597'-3
,,ar.- -
7 s-53 s-
=
s-3,0 m,.
CACl
-+'
f, =
ar 3-sis (BOTTOM OF CAC) 5-c80* ----osv,
r s
,,e 0-
~0
,,0-rc ilui.9.
g I,
o 5""5 ro tCCs Ctas.
rac.
- 3. M
$ 70' ---- -
6s'---- -~~r----------------~~-------------~~~---~~~---~~--------------------------------~~~~~-------------------------------
5 24
- TO CCW
_' 570' 562 -
t.
560' ----- -
~~----~~~~-----------------------~~------------~~----------------------~~--------------------- -------------------~~~-~~~---~~--------------------------...---
INFORMATION ONLY
.