B11913, Forwards Listed Encls,Including Current Emergency Operating Procedures,In Response to 851017 Request Re Use of Feed & Bleed in Environ Qualification of Equipment Outside Containment
| ML20136D506 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 12/13/1985 |
| From: | Opeka J CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | Charemagne Grimes Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20136D511 | List: |
| References | |
| B11913, NUDOCS 8601060192 | |
| Download: ML20136D506 (34) | |
Text
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CONNECTICUT YANKEE ATOMIC POWER COMPANY BERLIN. CONNECTICUT P o Box 270 HARTFORD CONNECTICUT 06141-0270 TELEPHONE 203-ess "
December 13,1985 Docket No. 50-213 Bl1913 Office of Nuclear Reactor Regulation Attn:
Mr. C. I. Grimes, Director Integrated Safety Assessment Project Directorate U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Grimes:
Haddam Neck Plant Environmental Qualification of Equipment Outside Containment By letter of October 17, 1985,(1) the NRC Staff requested that Connecticut Yankee Atomic Power Company (CYAPCO) provide additional information regarding use of " feed-and-bleed" at the Haddam Neck Plant. In response to the Staff's request, CYAPCO provides as attachments the following information:
1.
The results of an analysis of the primary system response for those transients and accidents that rely on " feed-and-bleed" as a means of decay heat removal.
A description of the analytical assumptions is included. (Attachment A).
2.
A summary of the present emergency operating procedures utilizing
" feed-and-bleed" and a description of the circumstances under which alternative equipment is used.
A copy of the current emergency operating procedures is also included. (Attachment B) 3.
The heat removal systems and components outside containment and the specific high energy lines whose rupture would expose them to a harsh environment (Attachment C). It should be noted that portions of this information were reviously provided to the Staff by CYAPCO letter of October 31,1980. )
(1)
H. L. Thompson, Jr. letter to 3. F. Opeka, dated October 17, 1985,
Subject:
Environmental Qualification of Equipment Outside Containment.
hOqq (2)
W. G. Counsil letter to D. G. Eisenhut, dated October 31,1980 at Section Nq B.7 of the Attachment.
J 8601060192 851213 PDR ADOCK 05000213 A - D. CR M HM M h only)
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EICSB (PARR)
FOB (W. RECAN)
o In its September 30,1985(3) and October 17 letters, the Staff also noted that it was reviewing the licensing records and related design information to determine "the extent to which ' feed-and-bleed' constitutes the primary design basis, as opposed to a supporting heat removal method, and the extent to which the Staff has reviewed it in that context....." To assist the Staff, attached is a partial summary of CYAPCO's correspondence and Staff actions regarding this issue (Attcchment D). Attachment D also serves to provide reference to previous submb als as requested by the Staff.
t CYAPCO reaf firms its position that the Staff has previously accepted " feed-and-bleed" at Haddam Neck as a primt. y method of heat removal (see Attachment D). More importantly, and independent of the past, this method of decay heat removal has been demonstrated to be viable and therefore represents an acceptable element of our demonstration of compliance with 10 CFR 50.49.
Nothing in 10 CFR 50.49 or any other regulation prohibits its use for this purpose. To the extent that the Staff now alters its acceptance of " feed-and-bleed" so as to call into question CYAPCO's compliance with 10 CFR 50.49, CYAPCO maintains that such action would constitute a backfit under 10 CFR 50.109.
It is important to note that operating p ocedures rely on " feed-and-bleed" as a least desirable alternative to provide decay heat removal. The operators are trained to follow procedures and will not be reluctant to use " feed-and-bleed" if this method is called for by the procedures. To provide some perspective on the implications of implementing this procedure, CYAPCO expects the cleanup efforts involved with the use of " feed-and-bleed" to be not substantially greater than the effort involved in cleaning up the containment af ter the reactor cavity seal failure in August 1984. The containment was decontaminated and routine
- access was restored four days af ter the event. This is not to suggest that such an evolution would be trivial, but to point out that it would not be comparable to the accident at TMI-2.
In any event, pursuant to the Integrated Safety Assessment Program (ISAP),
CYAPCO and the NRC Staff are working together to determine if any areas of the plant present public health and safety risks that should be reduced. Among the tools which will be used to arrive at the appropriate corrective actions is a plant specific probabilistic safety study, which remains scheduled for submittal in March,1986. Use of " feed-and-bleed" along wit,h other related issues are being considered in the context of the ISAP analysis. In that the issue of " feed-and-bleed" is closely related to other issues currently under review (e.g.,
auxiliary feedwater reliability), CYAPCO believes the issue can best be resolved in this manner. We have every intention of identifying the presence of any significant safety issues, and if any are identified, resolution will be pursued aggressively.
(3)
H. I.. Thompson, Jr. letter to 3. F. Opeka, dated September 30, 1985,
Subject:
i.nvironmental Qualification of Equipment Outside Containment. This letter also requested an immediate submittal of a justification for continued operation. It is our current understanding that this request was withdrawn during an October 2,1985 meeting between the NRC Staff and Northeast Utilities.
. If you have any questions regarding the attached information or if further information is needed, my Staff is available to provide the necessary assistance.
Very truly yours, y
CONNECTICUT YANKEE ATOMIC POWER COMPANY l
h, i 2ll3 M'
' J. F. Opekal U
Senior Vice President Attachments l-1 4
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Docket No. 50-213 B11913 Attachment A Haddam Neck Plant Results of Analysis of CorG Cooling IJsing "T aed-and-Bleed" December,1985
{
Results of Analysis of Core Cooling Using " Feed-and-Bleed" The results of " feed-and-bleed" analyses using the RETRAN computer code are included in Appendices I, II, and III.
Appendix I contains the results of analyses which include explicit modeling of the reactor vessel upper head region and have been QA verified. Two cases are i
presented in Appendix I.
The first case demonstrates that " feed-and-bleed" is an effective means of core cooling when one PORV is held open and only one HPSI pump, one charging pump, one RHR pump, one RHR heat exchanger and their associated support systems are available for use. The results show tha+, although some system voiding occurs, subcooled fluid exits the core throughout the transient.
The second case presents the results of an analysis which resulted in the limiting containment response. It was determined from a sensitivity study that holding open both PORVs with the combination of one HPSI pump operating with no charging pumps operating simultaneously resulted in the most severe containment temperature and pressure profiles of all the " feed-and-bleed" cases analyzed.
A discussion of the resulting containment response using the CONTEMPT computer code is presented in Appendix II.
The transient thermal hydraulic analysis has recently been recalculat9,(utilizing a more accurate modeling of the charging pump performance curveW and the reactor coolant system.
The results of this analysis show that following a complete loss of main and auxiliary feedwater,30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> are required to cool the RCS to RHR initiation conditions. These results bound the less limiting transient c
following a small break LOCA. Appendix III contains the results of the analysis and further discussion.
(1)
The impact of the change of the charging pump curve on feed-and-bleed analysis was identified as a deficiency in the Connecticut Yankee Plant Design Change Task Group Final Report. CYAPCO's response by letter, dated November 6,1985 stated that this evaluation would be completed by January 31, 1986.
The evaluation was completed ahead of schedule to support the Staff's request.
A-2 4
r.
i Docket No. 50-213 B11913 Appendix I Core Cooling Using PORV's at the Haddam Neck Plant December,1985 A-3
CORE COOLING USING PORV'S AT THE HADDAM NECK PLANT Feed-and-Bleed Transient Feed-and-bleed is a means of cooling the core down to the residual heat removal (RHR) system operating conditions in the event that the steam generators are unavailable to remove core decay heat. Feed-and-bleed is needed if a main steam line break outside containment were to compromise all capability to inject feedwater and auxiliary feedwater into the steam generators.
Such an incident would result in a full scram followed by a gradual reactor coolant system (RCS) temperature and pressure rise as the secondary side of the steam generators boils dry.
If all attempts to restore feedwater are unsuccessful, the operator begins feed-and-bleed when the RCS pressure approaches the power operated relief valve (PORV) setpoint of 2285 psia, or the core exit temperature approaches 575 degrees F, whichever comes first.
The operator begins feed-and-bleed by starting at least one charging pump or at least one high pressure safety injection (HPSI) pump and then holding open just one PORV and its associated block valve.
Pump suction is taken from the refueling water storage tank (RWST).
The PORV flow decreases pressurizer pressure and it Teases level until water enters the PORV and two-phase flow is discharged.
Me discharge flow is piped to the pressu-izer relief tank (PRT).
Eventually, the tank rupture disc passes water and steam into the containment.
If the charging pump (s) were stopped and low pressure safety injection (LPSI) pump (s) were started by a safety injection signal, the operator would shut off the LPSI pump (s) and restarts the charging pump (s). After about 100,000 gallons have been pumped from the RWST, the operator would align the containment sump valves to the RHR system and the RHR pumps would be aligned to the suction of the charging pumps to establish high head recirculation. When the RCS conditions fall below 375 degrees F and 340 psig, the operator would switch to the long term cooling mode.
RETR AN Analyses The feed-and-bleed transient was analyzed using the RETRAN02 MOD 002 and MOD 003 computer codes.
The scenario investigated was a total loss of feedwater with coincident loss of normal power. Since feed-and-bleed is a last resort method of core cooling, initiation is delayed as long as possible to permit restoration of main or auxiliary feedwater. As a result, the scenario can be divided into three parts - 1) before opening the PORV, 2) af ter opening the PORV (before switching to sump recirculation), and 3) af ter opening the PORV (after switching to sump recirculation).
Justification for the use of RETRAN for these analyses is provided below.
A-4
The first 30 to 40 minutes of the transient is characterized by the reactor coolant system heatup and pressurization typical of a loss of feedwater transient. During this period of time, RETRAN models the control of pressuriz.r level and the degradation of the steam generator heat removal capability. Since the loss of feedwater accident is a typical use of RETRAN, use of RETRAN for this portion of the feed-and-bleed event is justified.
Guidance is given to the operator to initiate feed-and-bleed prior to the pressure reaching the PORV open setpoint (2285 psia) or the core exit temperature exceeding 575 degrees F.
At this time, the operator would start one high pressure safety injection pump (one charging pump already running) and then hold open one PORV. Analyses using RETRAN have shown that, for these conditions, the injection flow exceeds the PORV flow and the energy removed by the fluid exiting the PORV is well in excess of the core deca / heat. As a result, the pressurizer fills up and the RCS cools down during this phase of the event. The only predicted system voiding af ter opening the PORV occurs in the upper head.
From a system voiding point of view, this result is similar to the results of a large main steam line break or a steam generator tube rupture event for which RETRAN is also qualified.
Af ter depletion of 100,000 gallons from the refueling water storage tank (RWST),
injection to the RCS is accomplished via a residual heat removal (RHR) pump delivering water from the containment sump to the suction of a charging pump.
Since the high pressure safety injection pump is turned off, the injection flow rate decreases during this phase of the feed-and-bleed transient, causing the system pressure to also decrease. RETRAN predicts additional voiding to occur in the more stagnant regions of the RCS when switching to sump recirculation.
However, the water injected into the cold leg flows through the core and the pressurizer as a single phase liquid. RETRAN predicts that the fluid exiting the core will be sub-cooled during this phase of the transient. Since the regions through which the injected water flows remain sub-cooled during this phase of the transient, the use of RETRAN for this phase of the feed-and-bleed transient is also justified.
The analyses performed show that feed-and-bleed adequately removes core decay heat. The fluid in the core region remains sub-cooled throughout the event and core uncovery would not occur. Two RETRAN cases are discussed below.
Case 1 is a minimum injection case to show adequate decay heat removal. Case 2 is the RETRAN case that results in the highest containment temperature and pressure profile.
Case 1 - Minimum Injection Flow This feed-and-bleed case showed adequate core cooling with minimum injection flow. The assumptions used for this case included:
One HPSI and one charging pump used for feed (assumes failure of one diesel generator).
One PORV held open.
A-5 l
'ncludes primary metal mass.
One charging and one RHR pump and RHR heat exchanger used for high-head recirculation.
The assumptions minimize the injection flow for the feed-and-bleed scenario both during the initial injection phase and for the high-head recirculation phase.
In addition it was assumed that feed-and-bleed was initiated when both core exit thermocouples exceed 5750F and RCS pressure reaches the PORV setpoint. This delays feed-and-bleed initiation since the procedures instruct the operators to initiate feed-and-bleed if either condition is met.
Table 1 provides a sequence of events and Figure 1 provides the RCS pressure for this scenario. There is a slight pressure rise at the beginning of the transient followed by relatively steady pressure.
As the steam generator inventory depletes, heat transfe-degradation occurs resulting in a large increase in pressure (%2000 seconds).
Feed-and-bleed was initiated when the PORV pressure setpoint was reached (%2400 seconds). After the PORV is opened, the injection flow exceeds the PORV flow (Figure 2) as the RCS pressure decreases.
When the RCS pressure decreases to the saturation pressure of the fluid in the upper head, some upper head voiding occurs (Figure 3), the pressurizer fills and the RCS pressure reaches an equilibrium value.
The RCS pressure stays effectively constant until the switchover to high-head recirculation (s-7000 seconds). At this time the HPSI pump is shut off and the only injection flow is from one charging pump taking suction from the RHR system. This results in a forther drop in RCS pressure and upper head level. Also, the decreased injection flow results in a heatup of the primary system (Figure 4). The pressure then stabilizes with sub-cooled flow through the PORV.
The figures show that this feed-and-bleed scenario provides adequate decay heat removal. At the end of the analysis the RCS pressure and the upper head inventory are stable. Also, Figure 4 shows that sub-cooled fluid exists the core outlet plenum throughout the transient, except for a brief saturated period when the PORV first opens.
The heatup following switchover to high-head recirculation results in a maximum upper plenum temperature, which remains sub-cooled, and a slow cooldown follows.
Case 2 - Worst Containment Response A sensitivity study of the number of operable pumps (HPSI and charging pumps) was performed to decermine the combination of operable pumps resulting in the 4
limiting containment response. The combination of one HPSI pump operating with no charging pumps operating simultaneously during the injection phase resulted in the most severe containment temperature and pressure profiles (See Appendix II) of all feed-and-bleed cases analyzed.
The assumptions used for the limiting case include the following:
Two PORVs held open when hot leg temperature exceeds 5750F or pressurizer pressure exceeds 2285 psia.
A-6 s
Charging pumps isolate when the safety injection (SI) signal is actuated and remain off until switching to sump recirculation.
One HPSI pump remains running from the time of SI actuation to sump recirculation.
One charging pump, one RHR pump and one RHR heat exchanger are used for high-head sump recirculation.
Table 2 provides a sequence of events for the scenario. Figure 5 provides the RCS pressure during the trar:sient. Feed-and-bleed was initiated when the hot leg temperature reached 5750F, approximately 1890 seconds into the transient.
Note that unlike Case 1,'it is conservative for this analysis to initiate feed-and-bleed early.
The RCS pressure remains relatively stable during the time the HPSI pump is injecting coolant into the system (Figure 5). This is because the injection flow nearly matches the PORV mass flow during this period of time (Figure 6).
As shown in Figure 6, the RCS injection flow rate decreases to approximately 83 lb/sec upon termination of flow from the HPSI pump and initiation of high-head sump recirculation (M240 seconds). Af ter termination of flow from the' HPSI pump, the system pressure decreases, resulting in some system voiding, and an increase in the enthalpy of the fluid being discharged through the PORV (see Figure 7).
However, the injection flow rate and PORV discharge flow rate eventually becomes equal and the discharge enthalpy reaches a near constant value at 5890 seconds.
7 A-7 i
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Table i HNP Feed-and-Bleed: Case 1 Sequence of Events Event Time (seconds)
J-Loss of Feedwater
.01 Reactor Trip 3.0 t
Hot Leg Temperature Reaches $750F 2030.
2 Pressurizer Pressure Reaches PORV Setpoint 2414.
One PORV Latched Open - Start of Feed-and-Bleed 2414.
- Switch to High-Head Recirculation 6997.
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Table 2 HNP Feed'and-Bleed: Case 2
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Event Time (seconds)
Loss of Feedwater 0.01 Reactor Trip 3.0 Hot Leg Temperature Reaches 5750F 1890.
Two PORV's Latched Open - Start of 1890.
Feed-and-Bleed Peak Pressurizer Pressure (2163 psia) Reached 1890.
1
. Switch to High-Head Recirculation 52te0.
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Docket No. 50-213 Bil913 Appendix II Containment Response for Feed-and-Bleed Method of Decay Heat Removal Haddam Neck December,1985 A-17
Containment Response for Feed-and-Bleed Method of Decay Heat Removal at Haddam Neck Containment Response This attachment describes the Haddam Neck containment response to a worst case feed-and-bleed transient. The worst case pump combination operating in conjunction with PORV operation was selected. The combination consisted of one HPSI pump operating with no charging pumps operating simultaneously. This resulted in the maximum total energy addition to containment before switching to the recirculation mode. The assumptions governing the mass and energy release are discussed in Case 2 of Appendix 1.
CONTEMPT Analysis CONTEMPT El/26A, a QA-qualified computer code for evaluating containment temperature and pressure, was used for the analysis.
The initial conditions / assumptions are as follows.
Initial Temperature 1400F Initial Pressure 17.5 psia Initial Relative Humidity 100%
Pressurizer Relief Tank Rupture Disk Failed at Time =
1890. sec.*
These assumptions are conservative and thus the calculation results of Case 2 in Appendix 1 will envelope all conditions which were considered for this analysis.
Case 2 of Appendix 1 predicts a maximum containment temperature of less than 2000F and 27.46 psia at time equal to 3725 seconds.
These values are conservative and are substantially less than the containment design parameters.
Note that the sump water attained a temperature of 2450F and a volume of 3
69,288 f t. This would raise the sump water level to approximately 2.5 f t.
Table I summarizes the maximum values inside containment for the analysis, all of which are enveloped by the LOCA profiles to qualify equipment. Figure I provides the containment pressure profile for this scenario. Figure 2 provides the containment liquid and atmospheric temperature profiles for this scenario.
- No credit is taken for the time delay between the opening of the PORVs and the rupture of the Pressurizer Relief Tank Disk.
A-18
TABLE 1 Feed-and-Bleed Containment Response Maximum Value Summary for Compartment 3 Variables Value Time (sec.)
Total Pressure (PSIA) 27.46 3,725.
Steam Pressure (PSIA) 11.42 3,726.
Atmos Temperature (Deg F) 199.7 3,725.
Pool Temperature (Deg F) 245.3 3,725.
Atmosphere Energy (BTU) 8.7289 x 107 3,731.
Pool Energy (BTU) 7.7860 x 108 43,190.
Total Energy (BTU) 8.1641 x 108 43,200.
Air Mass (LBM) 1.4694 x 105 0,0 Water Mass, Atmos-Vapour (LBM) 6.5877 x 104 3,726.
Water Mass, Atmos-Liquid (LBM) 157.83 3,001.
. Water Mass, Pool (LBM) 4.0830 x 106 43,200.
Water Mass, Total (LBM) 4.1044 x 106 43,200, Humidity (-)
1.0 0.0 Condensation Rate (LBM/SEC) 1.4699 x 106 4,986.
Mass Transfer Coeff (LBMOL/S-FT2) 1.4939 x 106 2,045.
Heat Transfer Coeff (BTU /S-FT2-R) 4.6390 x 106 2,036.
Fan Cooler-Cooling Rate (BTU /HR)
-7.4167 x 107 3,729.
A-19
FIGURE 1 CY FEED AND B'.EED CONTRIN.".ENT RESPONSE INIT: 140 D:G Hwo 17.5 Pn '
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Docket No. 50-213 B11913 Appendix III Long Term Coolability Following Feed-and-Bleed Haddam Neck Plant December,1985 A-22
FEED-AND-BLEED LONG TERM COOLABILITY Introduction This section provides a transient thermal hydraulic analysis of a complete loss of main and auxiliary feedwater event at Haddam Neck during which feed-and-bleed is initiated to cool the reactor coolant system (RCS) to residual heat removal (RHR) initiation conditions. When RHR entry conditions are achieved, the feed-and-bleed mode of cooling can be discontinued for this event. The less limiting small break LOCA event is bounded by the results of the loss of main and auxiliary feedwater event.
Based on the analysis contained herein,30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> are required to cool the RCS to RHR entry conditions using the feed-and-bleed method of cooling. A discussion of the method of analysis along with the analysis results are presented in the following sections.
Method of Analysis i
The analysis was performed with the NULAPS transient non-equilibrium thermal l
hydraulic blowdown code. NULAP5 is a modified version of RELAP5/ MODI which was developed by EG&G of Idaho Falls, Idaho. Northeast Utilities has made extensive modificatiors to RELAPS/ MODI in order to enable the code to perform small break LOCA ECCS licensing analyses for the Haddam Neck Plant.
A detailed description of the NULAPS code along with the code benchmarking qualification, and application to the small break LOCA is given in References 1 through 3 which were previously provided to the Staff.
Figure i presents the NULAP5 model of the Haddam Neck Plant. A single loop containing the pressurizer was explicitly modeled while the remaining three loops were combined to form the two loop representation of the system. Also note that the PORV discharge piping was modeled to include the additional resistance effects on the cooldown time. Additional assumptions and inputs to this analysis are presented below.
1.
The initial core power is 1861.5 Mwt (102% of the full power value of 1825 Mwt) 4 2.
The decay heat curve utilized the 1971 ANS Standard increased by 20%
3.
Complete loss of offsite power assumed.
4.
One diesel generator is credited resulting in minimum safety injection which includes one HPSI pump operating during the injection phase with a switch to one charging pump during the recirculation phase.
I A-23 i
5.
Only one PORV is activated. No other operator actions were assumed.
6.
Feed-and-bleed is initiated once the hot leg temperature reaches $750F following reactor trip.
Discussion of Results A complete loss of offsite power initiates the event which results in a reactor trip and termination of the main feedwater. No auxiliary feedwater is credited.
As a consequence of the loss of offsite power, the reactor coolant pumps are also tripped followed by a coastdown which results in a decrease in the primary system flow rate. Primary pressure is reduced early in the event as noted in Figure 2 as the PORV is opened within the first 300 seconds when the hot leg temperature reaches 5750F. However, it should be noted that the PORV would have been opened at any time between 0 and about 30 minutes into the event since steam generator heat removal capability is not lost until af ter this time period. As such, the analysis results would remain valid for all PORV opening times within 30 minutes of the event. Actuation of the HPSI pump results in a small increase in primary pressure as noted in Figure 2 early in the event as the RCS refills with injected water. A pressure plateau results and persists for approximately the first three hours of the event during which RCS pressure is controlled by the HPSI pump. At about three hours, recirculation is initiated and the charging pump is placed in service. Due to the lower capacity of the charging pump, RCS pressure cannot be maintained and the RCS depressurizes since the PORV discharge rate exceeds the charging pump flow as shown in Figure 3 at approximately three hours into the event.
After this depressurization period, as shown in Figure 2, the RCS again repressurizes as the charging pump increases RCS pressure until the PORV flow matches the injection flow at about six hours into the transient. For the duration of the event, the RCS slowly depressurizes as core decay heat decreases and the charging pump and PORV cool the core until RHR conditions of 350 psia and 3750F are reached between 25 and 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> into the transient (see Figures 2 and 4). At this time, the RHR system can be placed in service and feed-and-bleed can be terminated.
During the recirculation portion of the event a steam bubble formed in the upper head of the reactor vessel which acted to slow depressurization during the latter portion of the transient (see Figure 4). Although the upper head region contained steam, the remainder of the FCS remained filled with liquid. Because the HPSI and charging pumps each can provide more than sufficient flow to match the core and system boil off, should boiling occur, there is no potential for draining the RCS loops the oughout this transient.
REFERENCES 1.
W. G. Counsil letter to D. M. Crutchfield, dated April 14, 1983,
Subject:
NULAP3, A Fortran IV Digital Computer Program for Nuclear Steam Supply System Blowdown and Fuel Rod Heat Up Analyses, April,1983.
A-24
. -. =.
2.
W. G. Counsil letter to W. A. Paulson, dated August 23, 1984,
Subject:
Supplement 1, (same title as above), July,1984.
3.
W. G. Counsil letter to_ W. A. Paulson, dated August 23, 1984,
Subject:
Calculative Methods for the Northeast Utilities Small Break LOCA ECCS
~ Evaluation Model, Volume I and II, July,1984.
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6 A-25
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Docket No. 50-213 B11913 Attachment B Haddam Neck Plant Emergency Operating Procedures For Feed-and-Bleed December,1985
Emergency Operating Procedures for Feed-and-Bleed Haddam Neck Plant Emergency Operating Procedure EOP 3.1-44 " Abnormal Feed Flow" utilizes
" feed-and-bleed" as a least desirable alternative during a complete loss of feedwater and a loss of heat sink.
Section 3.8 of the EOP describes the circumstances under which alternative equipment would be used (e.g. PORV's and safety injection or charging flow). A copy of this procedure is attached.
" Feed-and-bleed" is also incorporated into Revision 1 of the Westinghouse
- Emergency Response Guidelines. These guidelines are utilized in draft Haddam Neck EOP. HP-Rev.1, " Response to Loss of Secondary Heat Sink."
It is CYAPCO's intention to continue to use " feed-and-bleed". as a least desirable alternative to remove decay heat under the new EOP format.
B-2
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