ML20039F422
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4 UNITED STATES
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'k NUCLEAR REGULATORY COMMisslON s
WASHINGTON, D. C. 20555
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JUN 12 55 DOCKET N05. STN 50-580 AND STN 50-581 i
APPLICANTS: 0HIO EDIS0N COMPANY, ET AL FACILITY:
ERIE NUCLEAR PLANT, UNITS 1 AND 2 St'3 JECT:
SUMMARY
OF MEETING HELD ON MAY 18, 1978 WITH OHIO EDISON ON THE ERIE NUCLEAR PLANT we met with representatives of Ohio Edison Company, On May 18, 1978, Gilbert /Comonwealth Companies (the applicants architect engineer) and Babcock and Wilcox Company. The purpose of the meeting wa:; to discuss the outstanding items in the review of the Eric Nuclear Plant Preliminary Safety Analysis Report. The meeting attendees are listed in the Enclosure.
The meeting covered the following review areas:
reactor systems, containment systems, structural engineering, auxiliary systems, instrumentation and control systems and power systems. The discussion below sumarizes the items discussed in each area and the resolution reached, if any.
Reactor Systems 1.
Missiles Inside Containment BSAR-205 contains a table that lists the BSAR-205 systems and components that must be protected from missiles inside containment and a table (3.5-3) that lists items for which missile protection must be considered. The applicants' PSAR did not reference this latter table.
In the meeting the applicants committed to consider missile protection for the items in BSAR table 3.5-3.
This commitment was to be documented in a future PSAR amendment.
2.
Leak Detection Systems We needed clarification on three aspects of the leak detection system.
First, we were not sure whether the alarm on the containment sump level rate of increase was alarmed in the control room and whether this alarm was the "one gallon per minute within 60 minutes" alarm. The applicants stated that the alarm was in the control room and the alarm would be set l
l 43 810403 PD j
MADDEN 80-515 PDR
to detect a sump level change equivalent to 1 gallon per minute within 60 minutes. Second, we wanted to know if the power to the airborne particulate radiation monitor would be Class IE. The applicants said that it was, but this was not documented in the PSAR. They committed to correct the PSAR in this area.
Third, the applicants had previously stated that since most of the identified leakage was being piped to i
the reactor drain tank, airborne radioactive particulate monitors would be calibrated by a check source and sampling of the containment atmosphere. Calibration of the airborne detector via the leakage to the sumps was not considered feasible. We asked the applicants if they would commit to consider calibrating, with the sump flow, if the sump flow during operation was great enough to allow calibration. The applicants stated that they would make that comitment.
3.
Boron Precipitation Post-LOCA l
Analysis of the method used to prevent baron precipitation post-LOCA was requested in the reactor systems Q-2's.
The applicants provided a draft of the analysis for Erie (a copy of this draft is attached to this summary). We stated that we would look at it and get back with them if we had any problems.
Subsequent to the meeting, we informed the applicants that our only problem with their proposed design was the lack of flow indication on the dump to sump lines.
4.
Single Failure Analysis
,g In our review of the single failure analysis of the ECCS, we
'4 noted several apparent errors. B&W stated that there were f [60 7
Z/g indeed some errors and that t:.. errors would be corrected as follows:
(1) the valve identified as MU-61 in the drawings d4 '
would be changed to MU-62 and (2) two entries in Table 6.3-2 would be reversed.
5.
Periodic Testing of Isolation Valves In our Q-2's, we stated our position that the closure time for the turbine stop valves, the main steam isolation valves, and the feedwater isolation valves assumed in accident analyses should be verified periodically during plant outages.
The intent of applicants response was not clear. The applicants stated that the PSAR would be amended to state that this periodic testing of the closure times for these valves would be performed.
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6.
Check Valves in the Make-up/Hi@ Dressure Injection (HPI) System We stated that several check valves in the makeup and HPI system that should be leak tested because they formed the boundary between the reactor coolant system and low pressure portions of the HPI system. We stated that it was our position that two check valves in series be leak tested. The applicants were not i
prepared to address our concerns.
7.
Cold shutdown The applicants did not wish to discuss this issue.
8.
Feedwater Isolation Valves The BSAR-205 had an interface requiring two feedwater isolation valves in each main feedwater line. The Erie design has only one isolation valve in each line and relies on closure of the feedwater flow control and startup valves as backups. We had requested that the applicants provide (1) justification for the apparent decrease in the reliability of their design and (2) modify the appropriate accident analyses where the isolation of feedwater was required. The applicants stated that they were in the process of evaluating whether or not to add the second isolation valve. We stated that such a change to be in conformance with the BSAR-205 would be acceptable.
9.
ECCS Automatic Piggyback The BSAR-205 design includes provisions for piggyback operation of the ECCS during the recirculation phase of a postulated small break LOCA.
In the Erie design the realignment is done automatically instead of manually as proposed by BSAR-205.
The applicants stated that they do it automatically because of NPSH restrictions on the C HPI pump and the short amount of time available for manual action. We asied them if there was a need to shut the HPI suction valves from the BWST during this mode of operation. The applicants stated that they would check 4
on it and let us know.
- 10. Turbine Bypass The Erie design uses signals from the RPS and ICS to align valves in the air supply to the turbine bypass valves to preclude steam flow in excess of 15% of design.
We asked the applicants if the diverse signals that interrupted the valves' air supply would be separated to prevent a single event from opening the valves. The applicants stated that they would take a look at this.
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. Auxiliary Systems 1.
Quality Grcup Classification of Component Cooling Water to the Seal Water Return Coolers b
The seal water return coolers are cooled by the non-essential component cooling water system.
This system is Quality Group D and is not seismic Category I.
Pripr to the meeting it was our position that these cooling lines be Quality Group C and I
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seismic Category I because loss of the cooling would lead to injection of seal water at greater than design temperatures and may result in reactor coolant pump seal damage. The applicants justification for not putting this heat load on the essential system was that if cooling were interrupted, the subsequent temperature increase in the makeup tank (1) would be slow, (2) would be detected by the temperature alarm on the makeup tank in time for an operator to take corrective action, assuming a 30 minute time lapse after the alann, and (3) can be terminated by several alternatives, the easiest of which is to increase letdown flow. We asked them if the temperature alarm on the makeup tank was safety-grade; it was not. We stated that in order to assume credit for the alarm it would have to be safety grade.
Based on our discussions, the applicants were going to reevaluate their position.
2.
Quality Group Classification of the Feedwater Startup and Flow Control Valves We told the applicants that we were reevaluating our position on the quality group and seismic classification of the feedwater startup and flow control valves.
Citing NUREG-0138 the applicants take credit for these non-safety valves in certain accident analyses. We stated that we may require that these valves be Quality Group B and not just C as previously required.
Accident Analysis 1.
Toxic Gas Monitors The Erie site is adjacent to two railways; several toxic chemicals are shipped past the site on these railways. To protect to control room personnel for selected chemical spills at the railways, the applicants proposed a single monitor for each chemical in each of the control room air intakes. We told them that they must justify not having redundant monitors in each air intake since they had not demonstrated that a single plume would reach both air intakes and a single failure may
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- i prevent isolation of one air intake. Alternately, we said they might be able to show that the probability of a failure'of the 1
monitorsconcurrentwitpalethaltoxicgasplumecoveringthe plant was less than 10- per year. The applicants were going to relock at this accident and the protection required for the control room air intakes.
2.
Tornado Missile Protection The main steam lines and the borated water storage tank (BWST) for the Erie design are not protected from tornado missiles. Since the BWST is required to mitigate the consequences of a main steam line break, one or the other must be protected from tornado missiles for the event to be covered by previously performed accident analyses. We stated that those systems and structures that are not protected must be assumed to be hit by tornado missiles.
Neither the steam lines or BWST can withstand a hit by missiles from the tornado missile spectrum. We stated that if they could demonstrate that such an event would not result in consequences exceeding an appropriate fraction of Part 100, tornado protection would not be required.
Containment Systems 1.
Typa C Tests of Isolation Valves in Closed Systems We stated that for those closed systems where the isolation valve would not be pneumatically tested (Type C), the applicants must demonstrate that (1) a water seal exists over the isolation valve and (2) that there is enough water to maintain the water seal for 30 days following a LOCA. The water seal must be maintained without relying on non-safety grade systems. The applicants stated that this could not be done for the main steam isolation valves (MSIV).
For these valves we suggested showing that a loop seal of water could be created in the main steam piping and that this seal would stop leakage through the steam generator to the MSIV's.
In addition, we informed the applicants that an analysis submittal by Davis Besse 2/3, which is currently under review, may be used if found acceptable by the staff.
2.
Hydrogen Generation We questioned the applicants' assumed zinc corrosion rates for the hydrogen production analyses. We gave the applicants a copy of the zinc corrosion rate curve we would use in our analysis (Enclosure 3).
We stated that we had found that hydrogen
generation rates were very dependent on containment temperature
.l and the applicants analysis did not appear to account for this.
1 We stated that in order to complete our review of the hydrogen recombiner system we needed a curve showing the long term i
temperature response of the containment to a LOCA. The applicants stated that they would provide a curve of the containment temperature for our confirmatory analyses.
Structural Engineering 1.
Weighted Model Damping The applicants described their analysis method for damping.
I Although their method deviated from the method in SRP 3.7.2, the applicants stated that it was conservative. We agreed that their method was acceptable and requested that they provide additional discussion in the PSAR on this. They agreed to do so.
2.
Use of ACI-349 We discussed whether or not it would be useful for the applicants to reference Regulatory Guide 1.142, " Safety-Related Concrete Structures for Nuclear Power Plants Other Than Reactor Vessels and Containments" in lieu c f 1 Jentifying the exceptions taken from ACI-349. The applicants stated that they would rather not do this. We had no problem with this.
The applicants stated they would update their response to reflect exceptions to items in Appendix A to ACI-349.
They discussed their exceptions and we concluded they were acceptable. The exceptions dealt with testing of piping buried in c.oncrete members and the adherence to Appendix A to ACI-318-71.
3.
Ductility in Concrete Beams The PSAR states that the ductility for concrete members where compartment pressurization controls design would be 3.0 or less.
We stated that our position was that the structures response to compartment pressurization should be maintained elastic and therefore ductility should be 1.0.
The applicants accepted this posi tion.
4.
Equation for Overall Response to Missile Impact Equation (G) in PSAR page 3.5-13 was not identical to that in a previously accepted reference.
The applicants stated that the change was intentional, but that they would modify it to be identical with the reference.
i 5.
Ninety-Day Concrete l
The applicants will design several structures based on the specified strength of concrete at ninety days after pouring.
The applicants will conform to our position on this with the exception of the average strength for core samples taken from the structure. The applicants stated that they would follow I
the guidance in ACI-359 which states that the core samples should average 85% of the specified strength with the worst sample no less than 75% of specified strength.
We stated that unless sufficient justification was provided, the average of the core samples should be 100% of the specified compressive strength. We consider this to be an open item.
6.
Fill Concrete The applicants stated that fill concrete underneath the foundation slabs would include a quality assurance / control program similar to that used for concrete used in seismic Category I structures. We find this acceptable.
Electrical Instrumentation 1.
Logic for Main Steam Isolation Signal PSAR figures 32.26-2 indicate that both the A and B ESFAS signals go to the main steam line isolation, feedwater isolation and turbine stop valves. We requested that they clarify what the "and" gate in the logic diagrams represented; our concern was that single electrical failures should not prevent actuation of the valves.
2.
Routing of ESFAS Signals Components in the Turbine Building The PSAR stated that the ESFAS signals to the feedwater flow control valves and startup valves, and the turbine stop valves would be carried to the valve b.v cabling routed in separate independent non-class IE raceways.
In the meeting, we requested that these cabling be routed in conduit as well as separated from the redundant cable. The applicants agreed to provide this.
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3.
Testing of Engineered Safety Features During Normal Operation I
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j Excluding the testing of the turbine stop valves, MSIV's, and feedwater isolation valves, we stated that we need additional information on the engineered safety features that will not be j
tested during normal operation. We stated that they should provide justification for not testing the actuated equipment in the component cooling water system, the reactor building isolation system and the essential service water system.
i Power Systems 1.
Design Criteria for the Offsite Power Systems The SRP lists several regulatory guides and IEEE Std's as acceptance critaria for the offsite power system.
These guides and standards should be used as guidance in the design of the offsite power system. We recognize that the regulatory guidance was written specifically for Class IE power systems and that the offsite power system is not Class IE; however, we requested that the applicants describe how the intent of this guidance will be used in the Erie design.
2.
f.vailability of Offsite Power Sources
'le stated that a single failure should not preclude the cvailability of at least one preferred offsite power source.
In the Erie design, the single failure of one non-class IE Sattery would cause the loss of all offsite power sources.
We stated that this was unacceptable.
3.
Power for HPI Pumps A&C The Erie design does not use the swing bus concept used in BSAR-205 for the C liPI pump.
Instead the C pump is permanently aligned to the A train. We asked the applicants to describe the mechanism that would prevent loading of both pumps onto the diesels. The applicants described this mechanism.
At the conclusion of the meeting we stated that our review of two items was incomplete and we might need further information from the applicants.
_9-l The first item is the classification of the reactor coolant bleed degasifier
'j package, the deborating demineralizers ' and the reactor coolant bleed holdup tanks.
We needed the radioactive inventories for these components.
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The second item was our calculation of the doses from a LOCA simultaneous with containment purging.
)ml.17//4Vi Dean L. Tibbitts Light Water Reactors Branch No. 2 Division of Project Munagement
Enclosures:
1.
Attendance list 2.
Boron precipitation analyses 3.
Corrosion rate curve cc w/encls:
See page 10 l
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Ohio Edison Company
- 10 JUN 1 2 EGS N
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a cc: Mr. B. M. Miller
~i Ohio Edison Company 76 South Main Street i
Akron, Ohio 44308 Mr. William Kessler i
Commonwealth Associates, Inc.
209 East Washington Jackson, Michigan 49201 Gerald Charnoff, Esquire Shaw, Pittman, Potts & Trowbridge 1800 M Street, N. W.
Washington, D. C.
20036 Thomas A. Kayuha, Esquire Ohio Edison Company 76 South Main Street Akron, Ohio 44308 Mr. A. H. Lazar Babcock and Wilcox Power Generation Group P. O. Box 1260
'ynchburg, Virginia 24505
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Mr. Robert W. Tuf ts 352 West College Street Oberlin, Ohio 44074 Ms. Evelyn Stebbins 705 Elmwood Road Rocky River, Ohio 44116 Mr. Richard E. Webb 2858 One Hundred Eleventh Street Toledo, Ohio 43611 t
Ohio Edison Company ATTN: Mr. Lynn Firestone Vice President 76 South Main Street Akron, Ohio 44308 l-I 1
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ENCLOSURE NO. 1 ai LIST OF PERSONS ATTENDING THE D}
MAY 18, 1978 MEETING WITH OHIO EDIS0N ON THE ERIE NUCLEAR PLANT OH10 EDIS0N COMPANY NRC - STAFF l
R. Walter S. Newberry J. Hultz G. Mazetis K. Campa D. Pickett GJBERT/COMMONWEALTHCOMPANIES J. Shapaker J. Halapatz W. Kessler B. Turov11n i
R. Long R. Lipinski L. Gundrum F. Rinaldi R. Thomas V. Benaroya I. Szigethy R. Kirkwood G. Clyde H. Li J. Letki J. Knox K. Kniel D. Tibbitts B_A'IC0CK & WILC0X F. McPhatter G. Glei AC.lS - STAFF R. Muller
M* M v\\
ENCIDSURE NO.2 211.78 The response to question 211.33 concerning prevention of (6.3) excessive boric acid buildup during the post LOCA period-is incomplete. Reference to the flow path using relief valves DH-3 RV8A and B in conjunction with valves V12A and V12B does not appear to be correct.
Provide or reference a complete dis-cussion justifying the acceptability of the " dump to sump" mode with respect to our criteria.
This discussion must at least include the following:
1.
Calculations to show that the dump to sump mode of di-
, lution meets the requirements of our position considering boil off due to decay heat (4 weight percent ma gin, etc.).
2.
Show that 40 gallons per minute is sufficient.
3.
Discussion to show that the system meets the single failure criterion.
4.
Discuss how the operator will insure adequate flow; describe the indications he will have in the control room. Our position is that the operator must be capable of confirming that sufficient flow exists.
5.
Show how the system is testable.
RESP 0fl5E: The response to question 211.33 has been revised to show the proper valve disignation for the boron dilution path.
1, Calculations have been performed for the dump to sump dilution mode to demonstrate its effectiveness in pre-venting excessive boric acid buildup during the post LOCA period.
Due to the internals vent valves, a period of natural circulation flow will exist within the reactor vessel.
This natural circulation flow path exists from the downcomer, through the core and upper internals, then through the vent valves and back to the downcomer.
In order to calculate the natural circulation flows, the JOAM2 code is utilized.
The output from the F0AM2 code, vent valve overflow and flow into the core as a function of relative core power, is utilized in calculating the boric acid concentration (C/Co) versus time.
The concentration, C, in'th reactor vessel core and downcomer is calculated by the equation:
C = mass s lute
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mass solvent C
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W C
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C o
ti core o where:
M
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core
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= concentration of flow into core, core Cj = concentration at time 1, i
tg = time of initiation of run, C, = reference concentration, W
= fl w rate out of core, out C
= concentration of flow out of core.
out The assumptions used to determine the boron concentration in the core region are as follows:
I a.
A constant core mixing mass is assumed.
The mixing mass is equal to the sum of mass of water in the core plus the mass of water in the upper plenum equal to the elevation of the water in the down-comer.
b.
A constant LPI injection rate is assumed.
c.
A constant value is set for the enthalpy needed to l
produce steam (1000 Btu /bm).
d.
K-factors, used in the determination of core equiv-alent water level during the decay heat drop line dilution mo'de, are based on full-flow values.
e.
The initial boron concentration of the. core, down-comer, and injection is assumed constant and equal.
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The injection water is assumed to mix with the vent
$$j valve overflow before entering the downcomer or going out the break.
g.
Since the reactor is presumed to undergo a slow transient, the process occurring in the core may be considered to be quasi-steady state.
Therefore, the reactor vessel is assumed to be in a quasi-steady state condition for each time step.
,h.
An initial power level of 3800 MWt plus 2% uncer-tainty was utilized.
i.
Decay heat is based on 1.2 times the ANS standard.
J.
A 40 gpm dump to sump flow was initiated at 24 hones.
Figure 211.78-1 shows the results of the analysis.
A maximum concentration (C/C ) of 1.54 was reached at 766.7 days, which iswellbelowtheSolubilitylimitof32:1.
Thus, the dump to sump dilution mode satisfies the HRC requirements.
2.
The calculations discussed above demonstrate that a dump to sump flow of 40 gpm is more than adequate to prevent excessive boric acid buildup during the post LOCA period.
3.
PSAR table 6.3-2 is revised to add the following entries:
Component Component No.
Mal function Comment 011 suction DH-V12A or V12B Fails to open Other string provides line valves 100% at design flow Dump to sump DH-V16A or V16B Fails to open Other string provides Sump valves 100% at design flow.
4 Preoperational testing of this system will be conducted to verify that each train nects design requirements.
After the dump to sump valves are opened and the system is placed in operation, there is no need to throttle the valves, and the operator will not perform any operations based on the dump flow rates.
Therefore it is the Rtw position that finw rate measur6.ent is not required fnr tne dump to sump lines.
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t 5.
Preoperational testing mil include verification of the-dump to sump.
The preoperational test procedures will e'
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-(2) one set of dump valves be opened, and (3) flowrate be l
calculated based on time to change the sump level.
This will be repeated for the second dump to sump string.
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W 100 1000 J0,000 TIME (HOURS):
BORON CONCENTRATION VS. TIME: 40gpm DUMP-TO-SUMP AT 24 HRS.
.j l OHIO EDISON COMPANY Enit nucle An PL ANI
. uNeis i n 7 F. I.G_U_.R E PS AR 211.78 1 AM 1716 017R) l l
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%el 211.33 Describe the procedures and system design for the ENP (6.3) which will prevent excessive boric acid 'uildup in the o
reactor vessel during the post loss-of-coolant accident long-term cooling period.
The following requirements must be satisfied:
E (1) The boron dilution function shall not be vulnerable to a single failure. A single active failure postu-lated to occur during the long term cooling period can be assumed. However this failure would then be in lieu of a single active failure during the short term cooling period.
(2) The inadvertent operation of any motor operated valve (open or closed) shall not compromise the boron dilution function nor shall it jeopardize the ability to remove decay heat from the primary system.
(3) All components 6f the system which are within con-tainment shall be designed to seismic Category I requirements and classfied Quality Group B.
(4) The primary mode for maintaining acceptable levels V
of boron in the vessel should be established.
should a single failure disable the primary mode, certain manual actions outside the control room would be allowed, depending on the nature of the action and the time available to establish backup mode.
($) The average boric acid concentration in any region of the reactor vessel should not exceed the level of four weight percent below the solubility limits at the temperature of the solution.
(6) During the post-LOCA long term cooling, the ECC system normally operates in two modes:
the initial cold leg injection mode, followed by the dilution mode.
The actual operating time in the cold leg injection mode will depend on plant design and steam binding considerations, but, in general, the switch-over to the dilution mode should be made between 12 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA.
(7) The dilution mode can be accomplished by any of the folicwing means:
s (a) Simultaneous cold leg injection and hot leg suction 211.33-1 AM. 8 (10-14-77) 4
211.33(Cont'd)
(b) 5 "Jitaneoas hot and cold leg injections (6.3)
(c) oli nate hot and cold leg injections.
QUESTION:
(B)
In the alt:'rnate hot and cold leg injection mode, the operathig time at hot and cold leg injection should be su.Ticiently short to prevent excessive boric acid builuop.
(9)
The minimum ECCS flow rate delivered to the vessel during the dilution mode shall be sufficient to accomodate the boil-off due to fission product decay heat and possible liquid entrainment in the steam discharged to the containment and still provide.
sufficient liquid flow through the core to prevent further increases in boric acid concentration.
(10) All dilution modes shall maintain testability com-parable to other ECCS modes of operation (HPI-short term, LPI-short term, etc). The current criteria for levels of ECCS testability shall be used as guidelines (i.e., Regulatory Guides 1.68, 1.79,,
GDC 37).
Describe your design cpnfiguration for precluding boric acid precipitation in auxiliary systems associated with the ECCS.
RESPONSE
The system to prevent excessive boric acid buildup in the reactor coolant system which.will be used for the ENP plant is shown in B-SAR-205, Figure 9.3-5 and consists of redundant 2" lines, and DH valves V16A and V16B. This I
arrangement, in conjunction with DH valves V12A and V128, provides single-failure proof safety-grade flow paths from the reactor hot leg to the containment sump and is referred to as the " dump-to-sump" mode of system opera-tion.
This mode of system operation will satisfy the requirements set forth in the above question.
Each dump-to-sump line is capable of providing 40 gpm flow from the reactor hot leg.
Each dump line provides for a positive flow path through the core by taking suction on the reactor outlet pipe and dumping the core outlet water into the sump.
The elevation of the decay heat drop line valves and the dump-to-sump valves. shall be such that adequate driving head is available to cause saturated water or water steam mixture to flow from the hot leg to and through the dump-to-sump line. Itakeup water to the core is provided via the LPI System which takes suction from the sump after the BWST has been emptied.
J 211.33-2 AM. 17 (06-01-78)
.i4 a
211.33(Cont'd) Calculations for concentration buildup within the reactor (6.3) vessel show that relatively long periods of time are
RESPONSE
available before the dump-to-sump mode is required.
These time periods are a function of power level, the decay heat level, and the ability-of heat generation within the core to support natural circulation flow.
within the reactor vessel. This natural circulation flow path is formed by the downcomer, the core, the upper l
internals,. the vent valves and flowback to the downcomer.
The minimum time at which the dump-to-sump mode of opera-tion must be available to prevent excessive boron concen-l tration is approximately 2 days. However, actual initia-tion of the dump-to-sump mode of operation of the decay heat removal system will be perfonned within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the LOCA.
The electrical bus and operation alignment for each dump-to-sump valve will be the same as those decay heat drop line valves OH-V12A and B, respectively.
System align-ments for the dump-to-sump mode of operations may be controlled from the c'ontrol room.
All components of this system which are within the contain-ment will be designed to seismic Category 1_ requirements and classified Quality Group A or B.
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211.33-3 AM. 8 (10-14-77)
e ENCLOSURE NO. 3
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