05000255/LER-1980-021-01, /01X-1:on 800725,during Valve Inservice Insp, Operator Opened Containment Sump Valve CV-3030.Valve Stayed Open 36-h.Caused by Operator Error.Valve Closed,Operator Counseled & Shift Turnover Checklist Revised
| ML18045A534 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 08/20/1980 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| Shared Package | |
| ML18045A533 | List: |
| References | |
| LER-80-021-01X, LER-80-21-1X, NUDOCS 8008260493 | |
| Download: ML18045A534 (65) | |
| Event date: | |
|---|---|
| Report date: | |
| 2551980021R01 - NRC Website | |
text
,_
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UPDATE REPORT PREVIOUS REPORT DATE August 6, 1980 NRC FORM366 (7.771 LICENSEE EVENT REPORT U.S. NUCLEAR REGULATORY COMMISSION PALISADES PLANT CONTROL BLOCK: I IG) 1 6
(PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION!
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7 8
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.LICENSEE COOE 14
,5 LICENSE NUMBER 25 26 LICENSE TYPE 30 57 CAT 5B CON'T l:rril 7
8
~~~~~~ LlJ©lo I 5 I o Io Io I 215. I 5 (?)I o I 112 I 518 Io @lo I 8 12 Io I 8 I ol@
60 61 DOCKET NUMBER 68 69 EVENT DATE 741 75 REPORT DATE 80 EVENT DESCRIPTION ANO PROBABLE CONSEQUENCES@
[)):I] jOn July 25, 1980, during the performance.of a guarterly surveillance test
@I!] I (valve* ISI program) an operator inadvertently opened containment sump valve I
[]JI] ICV-3030.
This valve on the suction side of the containment spray HPSI and
((((] ILfSI pumps is normally closed during plant operation.
CV-3030 was open 36
[)))) !hours.
This condition is reportable under Technical Specifications IIl2J 6.. 2.A.2.
[))))
7 B 9 SYSTEM
CAUSE
CAUSE COMP.
. VALVE BO rIIJJ CODE CODE SUBCODE COMPONENT CODE SUBCODE SUBCODE Is IF I@ LAJ@ LAJ@ Iv IA IL Iv IE Ix I@ ill..J@ llLl@
9 10 11 12 13 18 19 20 7
B r,:;-.
LE A/AO CVENT YEAR
\\!:,) REPORT 18 I 0 I NUMBER
~
22 SEQUENTIAL OCCURRENCE REPORT REPORT NO.
COOE TYPE I
I Io I 2111 l.:::::J lo 11 I l1_J 1=J 23 24 26*
27 28 29
. 30 31 ACTION FUTURE EFFECT SHUTDOWN f.'.:::\\
ATTACHMENT NPRD-4 PRIME COMP.
TAKEN ACTION ON PLANT METHOD HOURS ~ SUBMITTED FORM SUS, SUPPLIER.
lx.J@Lx.I@
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33 34 35 36 37 40 41 42 43 CAUSE DESCRIPTION ANO CORREC"rlVE ACTIONS @
REVISION NO. w 32 COMPONENT MANUFACTURER I w 10 13 IOI@
44 47 IO~erator error resulted in the opening of CV-3030 during a routine quarter-I o:DJ !ly valve ISI surveillance activity.
Valve CV-3001 (containment spray valve)
!IILl 1was the intended valve to be cycled.
- The handswitches for CV-3030 and ITIIJ I CV-3001 a.re sj de by side on the main control console.
Corrective action
o:JI] land additional details are provided on the attachment.
7 B
9 FACILITY ljQ\\.
STATUS
% POWER OTHER STATUS \\.:::;:I [ill] W@ I ol 9 I o I@... ! N_A ____ __
BO METHOD OF DISCOVERY DISCOVERY DE.SCRIPTION lAJ@!Licensed Operator 8
9 10 12 13
. ACTIVITY CONTENT
~
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7 B
9 10 11 45 46 44 BO 44 45 NA LOCATION OF RELEASE @
BO PERSONNEL EXPOSURES
~
NUMBER
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DESCRIPTION
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7 B
9 11 12 13 PERSONNEL INJURIES
~
80 NUMBER DESCRIPTION~
[2JI] lol ol ci@)~NA _________________________
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7 8
9 11 12 LOSS O* OR DAMAGE TO FACILITY t4j\\
TYPE
DESCRIPTION
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10 68 69 80
CONSUMERS POWER COMPANY Attachment to LER 80-021, Rev 1 PALISADES PLANT Docket 50-255 License DPR-20 Description of Occurrence During routine testing of valves at approximately 1930 on July 25, 1980, an oper-ator inadvertently opened the isolation valve between the containment sump and engineered-safeguards pumps P-54B and P-54C (containment spray), P-67B (LPSI) and P-66B and P-66c (HPSI).
The valve (CV-3030) is normally closed, remains closed during the injection phase of a safety injection sequence, and is opened auto-matically on a signal of low level in the Safety Injection and Refueling Water (SIRW) tank.
The corresponding valve (CV-3029) in the second line from the contain-ment sump to the redundant set of engineered-safeguards pumps was closed and remain-ed so.
CV-3030 remained open until approximately 0730 on July 27, 1980.
Determina-tion that this occurrence was reportable occurred at 1215 on July 28, 1980.
Significance/Probable Consequences The normal suction valve lineup for the engineered-safeguards pumps is shown in Figure 1.
Normally both safety injection refueling water (SIRW) tank discharge valves are open and the sump recirculation line valves are closed.
Following a LOCA or a main steam line break, the suction valve lineup remains unchanged until a condition of low SIRW tank level is reached.
At that time, the SIRW tank dis-charge valves are automatically closed, the sump recirculation line valves are opened, and the mini-flow recirculation lines between the discharges of the HPSI and LPSI pumps and the SIRW tank are closed.
If a sump recirculation line valve were open during drawdown of the SIRW tank, wa~er would be drawn from the sump rather than from the SIRW tank only if containment pressure exceeded the elevation head between the sump and the SIRW tank by a sufficient amount to open the recircu-lation line check valve and seat the SIRW tank discharge check valve.
This eleva-tion head has been estimated to be about 32 psi during the early stages of the accident.
Three potentially significant safety problems associated with a misaligned sump recirculation line valve have been identified.
They are:
- 1.
The effect on emergency safeguards pump operability, and therefore, on core and containment cooling, if the sump were to dry out.
- 2.
The effect on offsite dose and control room habitability of pumping potential-ly contaminated sump water through the mini-flow recirculation lines and into the SIRW tank.
- 3.
The effect on the main steam line break accident of being able to inject borated water with only one, rather than two HPSI pumps.
The question of emergency safeguards pump operability is addressed in Appendix A.
Based on the analyses presented, it has been concluded that the pumps on the affect-j ed train would always be pumping water, and therefore, would be capable of fulfilling!
their intended function; ie, Primary Coolant System inventory makeup and containment I spray.
Also in Appendix A, the length of time that containment pressure would exceed
)
L Attachment to LER 80-021, Rev 1 August 20, 1980 2
32 psig is estimated at 150 seconds.
This assumes all safeguards pumps are o:per.able.... This _interv!;,l iF>_ t):lus established,_ a,.~ the m~;Lmum lep.gt.b,_ of time that water would be drawn from the sump.
The effect of recirculating contaminated sump water to the SIRW tank is evaluated
.for the case of all emergency safeguards pumps operable in Appendices Band C.
In Appendix B, the total volume of sump liquid recirculated to the SIRW tank is calcu-lated.
Based on the result, the effects on offsite dose and control room habitabil-ity are evaluated in Appendix C.
In the unlikely event that a Maximum Hypothetical Accident had occurred with the recirculation sump valve open, thyroid dose at the site boundary would not have exceeded 35.6 Rem in the first two hours.
Appendix D contains an analysis of the containment pressure response to a large LOCA assuming one train of emergency safeguards equipment (pumps and air cooling fans) is out of service and the sump recirculation line valve in the other train is open.
This analysis was requested by the NRC staff at a meeting held at Region III offices on August 13, 1980.
This analysis takes into account that the tempera-ture of the containment spray water will rise to that of the sump once the cold water in the spray lines and headers has been flushed and replaced with hot sump water.
With two containment spray pumps operable, the time required to flush the lines has been estimated to be approximately 100 seconds.
Assuming this to.be the case, and also that only one air cooling fan is operating, the analysis still shows that containment pressure will exceed 32 psig for no more than 150 seconds.
A total of 490 gallons of sump water is calculated to be recirculated to the SIRW tank in this case.
Appendix E contains an evaluation of offsite dose and control room habitability based on the results in Appendix D.
Appendix F contains an evaluation of the core reactivity consequences for a main steam line break early in Cycle 4 (approximately 1,400 Mwd/Mtu core average exposure).
This evaluation shows that, at this point in the cycle, the reactor could be cooled to 600F and remain subcritical (until xenon decay) assuming no boron injection and the highest worth rod stuck out of the core.
The effects of having only one HPSI pump aligned for injection of borated SIRW tank water after a main steam line break at the worst point in the cycle (end-of-cycle) are evaluated in Appendix G.
This evaluation shows that the consequences of such an event would be slightly more severe than presented in the FSAR and in the thermal power uprating report, wherein two HPSI pumps were assumed to be aligned for borated water injection.
Summary:
The analyses provided herein show that in the unlikely event of a loss of coolant accident or a main steam line break accident with CV-3030 open, the conse-quences would have remained within applicable regulations and guidelines.
Corrective Actions
After CV-3030 was discovered open, the valve was immediately closed.
The operator involved was relieved of watchstanding duties, provided additional counseling as to his responsibilities and attentiveness while on duty, and after resumption of duties has been closely observed by the Shift Supervisor.
The valve's position is displayed on the control panel and changes of position of the valve are recorded on a data logger.
"Benchmarks" indicating normal system alignments for power operation have been placed on control room panels to aid
Attachment August 20, e
to LER 80-021, Rev 1 1980 3
operators in determining off-normal conditions.
The shift turnover checklist has been revised to include verification of system alignment of all engineered safety features by comparison of switch positions and indicating lights with their bench-marks.
CV-3029, CV-3030 and other valves essential for engineered safety features performance have been added to the checklist.
The checklist also includes a require-ment to document any keys inserted into keylocks and to explain their presence.
Periodic review of data logger printouts are being performed to detect system align-ment changes.
This occurrence has been discussed with Shift Supervisors and other licensed oper-ators.
Their responsibilities in maintaining correct system alignments and in recognizing the significance of off-normal alignments was re-emphasized.
To provide continuity during shift change periods, Shift Technical Advisor (STA) relief times have been staggered by two hours from the remainder of the shift.
The STA's responsibility in providing engineering evaluation for varying plant operating conditions has been re-emphasized.
.A PALISADES 7 /80 ECCS TRAIN A ECCS TRAIN B SAFETY INJECTION REFUELING WATER TANK CV-3057 CV-3031 NORMAL CONFIGURATION SAFETY INJECTION & CONTAINMENT SPRAY FIGURE 1 CONTAINMENT SUMP I CV-3030 I
APPENDIX A EFFECTS ON ENGINEERED SAFEGUARDS OPERABILITY
~\\
PROBLEM:
To determine the affect of CV-3030 being left open during plant)operation.
CV-3030 is the isolation valve on the recirculation line from the containment sump to ECCS train B (See Fig. 1).
Prior to RAS, following a LOCA, the containment sump will be isolated from the ECCS train due to the SIRWT head maintaining check valve 3166 closed; even though CV-3030 is open.
However, as the containment pressure increases due to the LOCA, there may be sufficient pressure on the back side of the check valve to counteract the SIRWT head.
When this happens a flow path is established from the sump to the ECCS train B~
Note that when 3166 opens 3239 closes, thus isolating the SIRWT.
Thus, the affect that this flow path has on the LOCA will be analyzed.
Once the flow path from the sump is established, two major concerns must be addressed:
(1)
Will the ECCS pumps on train B draw steam from the sump, thus possibly de-grading their operability.
(2)
At the discharge of the LPSI and HPSI pumps is a recirculation line to the SIRWT.
This is to protect the pumps from deadheading if the reactor pressure is greater than the pump shutoff head.
This recirculation line is valved closed after RAS but is open before.
Thus, if pump flow is coming from the sump, reactor water will be added to the SIRWT presenting a possible exposure problem.
In order to determine the extent of the problem, the amount of reactor water added to the SIRWT needs to be determined.
Effects on Engineered Safeguards Operability The Question is will the train B ECCS pumps draw steam from the sump and thus become steam bound or in some other way degraded.
In order to make this determination the following must be determined:
(1)
For what period of time is train B drawing from the containment sump?
( 2)
What is the flow rate out of the sump?
( 3)
What is the flow into the sump?
( 4)
Is the sump ever drained thus allowing steam to the pumps?
Appendix A 2
A.
Containment Pressure Needed to Establish Flow As shown in the diagram below if P1 > P2 the sump will remain isolated.
- Thus, the containment pressure necessary such that P2 > P1 must be calculated.
5-11(~*-r' pl = elevation head of SIRWT p2 = elevation head of sump + Pc
~
where Pc is containment pressure.
t~1 u
f 1
{;'
--~--<r'li--~---.~~~
v, i
The same elevation of 3239 and 3166 is the or 583 1 5" (Ref M 107 sh 10-2 150
-;; ~u.,,,p5 Assume that the water in the SIRW tank is at the low level alarm: 665 1 4" (Ref M-398-Sheet 18). pl = elevation of water in SIRWT - elevation of check valve
= 665 1 4" - 583'5"
= 81.9' containment p2 = elevation of water in sump - elevation of check valve + pressure
= (590' 583'5") + Pc Elevation of containment sump:
In order to minimize the containment pressure needed to open the check valve the sump is assumed to be filled up to the 590' floor:
Top of sump 588 1 4" floor 590' elev.
Bottom of sump 585' Ref: M-74 Rev 7 The pressure dp across the check valve is zero when:
pl = p2 81.9' = 6.58' + Pc 6? lb ft2 ft~ x 144 in2 Pc = 75.3' x
Pc = 32.4 psig
- 1)
\\(
Appendix A 3
Therefore, for containment pressures in excess of 32 PSIG it is assumed that a flow path is established from the containment sump to ECCS train B.
B.
Period of Time Containment Pressure Above 32 PSIG
- - To be conservative, the maximum time that the containment pressure is above 32 PSIG is used.
The values listed below are from the FSAR chapter 14.
These are higher than expected values since the heat sinks used were very conservative.
Break Size
- Time Pressure above 32 PSIG
~ t 42" DE Fig 14.18-2 1.6 sec - 140 sec 138.4 42" SE Fig 14.18-3 3.0 sec - 115 sec 112 30" SE Fig 14.18-4 6.4 sec - 150 sec 143.6 18" SE Fig 14.18-6 22 sec - 170 sec 148 24" SE Fig 14.18-5 12 sec - 150 sec 138 To bound the above numbers 150 sec will be used as the time that the containment is above 32 PSIG.
- Normal operation - 4 air coolers/3 spray pumps C.
Flow From Sump As shown in Figure 1 the flow from the sump will be through 2 containment spray pumps, 2 HPSI's and 1 LPSI.
Thus, the flow rates of these pumps as a function of reactor pressure must be known in order to determine the total flow rate from the sump.
Containment Spray:
2 spray pumps will be connected to the sump and a third pump (on train A) is still connected to the SIRWT.
With 3 pumps flowing into 2 headers the flow rate per pump is 1100 gpm.
Once the initiating signal for containment spray (CHP) is reached there is a delay in getting a start signal to the pump and bringing the pump up to speed.
For this analysis a total delay of 4 sec is used.
Also, once the spray pumps are on it takes an additional 22 sec to fill the spray headers (assume no flow from nozzles until all headers filled).
After the sprays flow into the containment it is
\\f Appendix A 4
just added to the sump.
Thus, the spray pumps are draining the sump from 4 sec -
26 sec after the break (actually, more water will be added to the sump than is taken away since the s~ray pump on train A will be adding SIRWT water.
However, no credit will be taken for this water).
Therefore, the following will be the flow from the sump due to the containment sprays:
Time 0-4 sec 4-26 sec
> 26 sec Flow from Sump 0
2200 gpm 0
Low pressure Safety Injection (LPSI) pump:
The LPSI is rated at 3,000 gpm.
- However, it has a shutoff head of 410 ft (-177 psi) so that for higher pressures it just main-tains recirculation flow.
From Tech Spec surveillance test data (See page 2, Appen-dix B) the LPSI recirculation flow is 230 gpm for head pressures > 410 ft.
Since the reactor pressure is much higher than the LPSI shutoff head, a break is chosen which most rapidly depressurizes the reactor and thus initiates LPSI injection sooner.
The* break chosen is the 0.6DEG/PD LOCA which is the limiting break for Palisades.
Table 1 lists the LPSI flow as a function of time after the LOCA.
It is based upon the reactor pressure transient as reported in XN-NF-77-24, the pump head as a function of reactor pressure and flow as a function of pump head as calculated in "Palisades ECCS Analysis -
Core 2".
High Pressure Safety Injection (HPSI) pumps:
Each HPSI is rated at 300 gpm.
The HPSI has a shutoff head of 2800 ft (-1200 psi) and thus provides core injection seconds after the LOCA.
The flow from each HPSI pump as a function of time from the break is listed in Table 1.
The total HPSI flow would be twice the value listed.
Total Flow from Sump:
The total flow rate from the sump is listed in Table 2.
No flow from the sump is assumed for 4 sec to account for SIS setpoint to be reached, start signal to reach pump, and pump to come up to speed.
The yalues in Table 2 do not include containment spray after 26 sec since this water is just being added back to the sump at the same rate.
Appendix A 5
D.
Flow into Sump The water being added to the sump is that coming from the break.
For this calcu-lation to be bounding the break flow should be minimized so as to minimize the flow into the sump.
Therefore, for determining flows into the sump a 1.0 ft 2 break will be assumed.
The peak pressure for the 1.0 ft 2 break is 41 PSIG which is above the 32 PSIG needed to establish flow path from sump to train B.
However, for breaks much smaller not.only will the peak pressure drop to below 32 PSIG but also the reactor pressure transient will be less severe-delaying LPSI injection to beyond the 150 sec time period.
2 CONTEMPT run Using the 1.0 ft blowdown data from {
BILLOl4
} the amount of water added to the sump (compartment 3 pool region) was calculated using CONTEMPT-26.
This calculation was done using the Uchida heat transfer correlation between the atmosphere in the containment and the heat sinks rather than the usual Tagami correlation.
The net result is to have greater containment pressures and a decrease in the water added to the sump.
Table 2 lists the water added to the sump from the break as a function of time since the break.
Note that the value listed is the amount of water that would be in the sump at that time if none were being removed by the pumps on train B.
E.
Water in Sump At the time of CV-3030 misalignment the sump was filled 8%.
Thus, the line from the sump to the ECCS pumps was full -
no credit is taken for the water in the sump itself.
The length of 24" pipe from the sump to the SIRWT tie-in is 60 ft (Ref M 107-10-4).
Thus, the volume of water iri the lines is:
60 ft fr
=
=
188 *. 5 ft 3 1410 gal Thus, not taking credit for water in the sump itself, there is initially 1410 gal in the sump system (includes 24" line).
Table 2 lists the amount of water in the sump for a 1.0 ft 2 break as a function of time since the break.
It is equal to the water in the sump from the flow less that removed by the ECCS pumps.
The information is also plotted in Figure 4.
Appendix A E.
Conclusions As shown in fugure 4 water is. always present in the sump.
Note that figure 4 is very conservative since the flows from the sump were maximized assuming a large LOCA but the flow into the sump was minimized assuming a smaller break.
6 In actuality for a large break the flows into the sump would be much larger and for a small break the flow from the sump would be much less (since LPSI injection is delayed).
Appendix A 7
Water in Sump As stated earlier in this report, the sump was at 8% level during the misalignment of CV-3030.
The analysis took no credit for water in the sump but assumed that the lines from the sump to the pumps were filled with water.
The volume of water used was just that in the 24" line which equaled 1410 gallons.
A question was raised as to the validity of assuming the 24"* line was filled with an 8% level in the sump.
The concern was due to the fact that the 24" line comes out the side of the sump and not the bottom.
As shown in the drawing on the following page, the bottom of the pipe is 4 3/4" from the bottom of the sump and slopes downward 2°57'.
The level sensor has a span of 120" (100% scale) with 0% being l" from the bottom of the sump.
- Thus,
..[ 1a1' ::100°1~
"1 *'.-:7 skpt?..~.;'-
~ - -
f
?,.4
f
/
Co r.-to.11\\m12.'i\\-f 5vl\\'\\ p V')./
8% is 120" x. 08 = 9. 6" from the 0% level or 10. 6" from the bottom of the sump.
In order to calculate the point at which the pipe is completely filled, an exaggerated diagram of the above figure is drawn below:
l
LI 7c:-")
I~' i 5,, 0:: ~ 4 /I -
( i D * (,.
- -1 *
.J
-~~*------ ~
t
--'~- _._- -
---~;
I ~.15
- z..
- .~*
~-
- - i,.* -
'::?. c- '=' r,/,, -::: *1 "'1 1..i,.
r---1~!.-**
(/'--
.l
(~
l -:.Q /,,
~*
Appendix A 8
Thus, for the first 29.4' from the sump the pipe is only partially filled.
If it is very conservatively assumed that the pipe is empty unless it is fully filled, then the volume of water in the 24" pipe is decreased by:
29.4 ft. ~
(24) 2 4
12 2
= 92.4 ft.
= 691 gallons Thus, the amount of water in the line drops from 1410 gal. to 1410 - 691 or 719 gal.
The 1410 gal. was only the water in the 24" line.
If the two 14" lines to the
- containment sprays were included, the volume would increase considerably.
From "Bechtel Mechanical Group Calculations, Vol. I and II", the length of 14" pipe for the two containment spray pumps is 104'.
Thus, the volume increase is 104 ft. x + ~i~) 2
= 111 ft. 2 = 830 gal.
J'his more than compensates for the 691 gallons assumed lost above.
Therefore, it is concluded that the previous results are still valid and conservative.
Appendix A TABLE 1 Time Since (l)Reactor (2)LPSI Pump (3)LPSI Break, Sec Press., PSIA Head, Ft.
Flow, gpm 0
1425
>Shutoff 230 2
1175 230 4
1000 230 8
825 230 12 600 230 16 300 230 20 100 320 3750 40 (S) SS. 7 277 4375 80 49.1 27S 4375 120 44.9 270 4375 200
- 39. 3 265 437S (1) Ref. XN-NF-77-24 for 0.6 DEG/PD (2) Ref. Figure 2 this report (3) Ref. Notebook:
"Palisades ECCS Analysis - Core 2" (4) Ref. Figure 3 this report (S) From CONTEMPT BILLOED C4)HPSI Pump Head, Ft.
>Shutoff 2700 2550 2320 2055 1670 1420 Run out 9
(3)HPSI Flow, gpm 40 39.5 175 420 460 540 600 600 600 600 600
_J
Appendix A 10 TABLE 2 (l)Total Flow (2)Water Removed C3)Water In Volume of Time Since From Sump In Sump From Water In Sump Break, Sec From Sum;e, gpm Interval, gal Break, gal Actual Gallons 0
1410 2
1410 4
2780 0
3,421 4831 8
3270 218 7,913 9105 12 3350 223 11,433 12,402 16 3510 234 14,562 15,297 20 7150 477 17,419 17,677 40 5575 1858 26,414 24,814 80 5575 3716 31,938 26,622 120 5575 3716 36,829 27,797 200 5575 7433 44,408 27,943 (l) From Table 1:
1 LPSI + 2 HPSI + 2200 for containment spray (26 sec. only)
(2) Equals the water removed between t and t + At = W (t +..1 t) (t + tit - t)
(3) Ref. CONTEMPT RUN #BILL07Y
.- -~*
PALISADES 7 /80 ECCS TRAIN A ECCS TRAIN B LPSI PUMP P-6 7 A SAFETY INJECTION REFUELING WATER TANK CV-3057 CV-3031 NORMAL CONFIGURATION SAFETY INJECTION & CONTAINMENT SPRAY FIGURE 1 CONTAINMENT SUMP CV-3029 I CV-3030 I
o' 0
0
'd
[\\)
\\0 rn11r-..
[001fll ~1~:11H
...... It c+
ro*
I IF{"~
10 X 10 TO THE CENTI METER KEUFFEL & ESSER CO. MAO[ IN U.S A.
l 18 X 25 CM.
46 1510 PUMP HEAD, FT
- .1=1l-*****
li:J"
.fl'!"*
- r
- t:1:**
jc[iFFHfFffFHHFH!i JAt.
1 llHI II
[
. l ct (J)
(J)
I I I I
10 X 10 TO THE CENTI METER KEUFFEL & ESSER CO. MADE IN U.S.A.
... ~H*-H H-H-H-tit
.. i-H*IJ 1
Cf?
. - J -. -
l-ttH+Hiftttttii*rH+l+ift1tHi 1U X 25 CM.
46 1510 PUMP HEAD jJ+/-l: :l+/-l::Hl:l:f+/-U Jii:t
!J l/**.
I!
10 X 10 TO THE CENTI METER l<EUFFEL Be ESSER CO. MADE rn US.A 18 X 25 CM.
Water in Sump, 1000 gallons IH+~H+++HH t+++r+H+tti*tt+t+t+t+rtt+HH++H+HH+iHtHt
... j 46 1510
APPENDIX B RECIRCULATION FLOW HYDRAULICS -
Appendix B 1
RECIRCULATION FLOW HYDRAULICS -
LOCA The discharge of the LPSI/HPSI pumps has a recirculation line to the SIRW tank.
This recirc line is to protect the pump from "dead-heading".
In each line is a flow orifice, the characteristics of which are not known.
However, the loss co-efficient of the recirc line can be calculated based on HPSI/LPSI surveillance test data.
Knowing the loss coefficient, the recirc line flow can be calculated knowing the pump head.
Table 1 lists measured recirc line flow as a function of pump head for each pump operating individually.
This information is from the June 1980 Tech Spec sur-veillance test.
Ap'.flendix B 2
H P S I L P S I PUMP P66A P66B P66c P67A P67B TABLE 1 HEAD, PSIG 1209 1209 1219 184 184 FLOW'* GPM 40 30 30 230 220 Using the above data the recirc line loss coefficient can be calculated from:
~p = K p v 2
+
g 2g For the HPSI pumps, the data for P66A will be used for conservatism.
w = 40 gal x
_,m_i_n __
min 60 sec x
ft3 7.48 gal
=.09 ft 3/sec
'{
Appendix B 3
W = vA v = W =.09 ft 3/sec A
.02 f't 2 v = 4. 5 f't/sec K = (~P -
g) (2g) p v-2 v is for 2 11 line A = 3
(_g_)z 4
(12)
A =.02 f't 2 g = 80 1 see page 5
34.7 psi K
(1209 lb - 34.7) (144 in2 ) (64 ft/sec 2 )
(
in2
) (
ft 2
)
(62 lb/f't3) (4.5 ft/sec) 2 K =
8620 For the LPSI's the data for P67A will be used.
W = 230 gal x min x ft3 min 60 sec. 4,....,8,,__g_al_
W = 0.5 ft 3/sec W = vA v = W = 0.5 ft 3/sec A
0.95ft2 v = 10 ft/sec K = (~P -
g) (2g)
P v2 v is for 3" line A =.!. ( _]_) 2 4 (12)
A= 0.05 ft 2 K = (184 lb
- - 34.7) (144 in2 ) (64 f't/sec 2 )
(
in2
) (
ft 2 )
(62 lb/ft3) (10 ft/sec)Z K = 222
Appendix B 4
SUMMARY OF RECIRC LINE LOSS COEFFICIENTS PUMP K
HPSI 8620 LPSI 222 Knowing the loss coefficients, the recirc flow can be calculated as a function of pump head as shown below:
Rx r.-------.t~~....
p P = l?i + k v 2 2g where P is in gauge pressure The distance l?i is from the recirc line tee to the water level in the SIRW tank.
recirc line tee evalation = 581' 3" (M107, sheet 22-5) bottom of SIRWT = 643' (M398, sheet 18) elevation from tee to SIRWT = 61.75 feet The elevation of the water in the SIRW tank is decreasing with time.
For conser-vatism the elevation should be minimized in order to maximize velocity and hence recirc flow.
From the FSAR Section 14.18 RAS for the 42" DEG/PD break occurs at about 1000 seconds.
But for this analysis only the recirc flow for no more than the first 200 seconds is of interest.
Thus, the level of the SIRW tank used is that level it would be at in 200 seconds.
AIJpendix B 5
- 1.
Low Level Alarm. on SIRW
- 2.
RAS 6654" }
6453" Height of water in tank assumed at t=O 2 = 20' Height of water after 200 sec Ii-200 l 20' = 16'
[_ loo[J Elevation to top of water in tank used:
645'3" + 16 1 = 661 13" Distance between water level and bottom of tank:
661 13" - 643 1 = 18'3" g = 61.75 1 + 18.25'
= 80 1 M398, Sheet 18 The pressure Pis essentially the pump head.
This pump.head is equal to the reactor head, plus the elevation head to the reactor plus the line losses.
Tables of pump head versus reactor pressure have been calculated for Palisades ECCS analysis.
Therefore, knowing P, the flow velocity v can be calculated from:
v = /(P-g) 2g k
and the mass flow is W = vA
Appendix B 6'
RECIRCULATION FLOW PUMP FLOW VELOCITY FT/SEC
- MASS FLOW GAL/MIN.
.54./ P-80
.086./ P-80 22.44 v 8.98 v Note that in the above equations P is the pump head in feet and is a function of reactor vessel pressure.
Using Figures 2 and 3 of Appendix A (plots of reactor pressure vs pump head) the LPSI and HPSI recirculation flow can be cal-culated as a function of reactor pressure.
The results of these calculations are listed in Tables 2 and 3.
Knowing the LPSI/HPSI recirculation flow as. a function of reactor pressure, the recirculation flows for a 0.6 DEG/PD LOCA can be calculated.
This is done in Table 4.
Note that while the reactor pressure is above the pump shutoff head the flow is just the recirculation flow as determined by tests.
The total SIRW recirculation flow for Train B is the sum of the recirculation flows of one LPSI and two HPSI pumps as listed in Table 4.
This calculated sump recirculation flow is listed in Table 5 as a function of time since the large LOCA.
Also listed is the volume of water added to the SIRWT between two time intervals.
These values are added to give the total volume of water added to the SIRWT as a function of time since the large LOCA.
J
.)
Appendix B 7'
As stated in Appendix A recirculation from the sump to the SIRW tank is assumed to occur for the first 150 seconds following the break.
From Table 5, at 150 seconds the total recirculation flow is "'.'220 gpm.
Upon interpolating, then, the total water added to the SIRW tank in 150 seconds is 601.6 gallons.
However, of interest is not the total water added to the SIRW tank but rather the total water from the LOCA that is added.
As shown in Appendix A the suction lines of Train B contained a minimum of 1410 gallons of water.
From Table 2 of Appendix A the total volume through the Train B pumps is 1152 gallons in 20 seconds (1152 is sum of values in column 3 up to 20 seconds).
Thus, it is con-servatively assumed that the 1410 gallons in the suction lines of Train B takes 20 seconds to be pumped out.
Therefore, for the first 20 seconds of recircula-tion to the SIRW tank uncontaminated (non-LOCA) water will be added.
From Table 5, in 20 seconds recirculation has added 111.1 gallons to the SIRW tank.
- Thus, the a.mount of water from the LOCA that is added in 150 seconds is:
601.6 gallons
- - 111.1 gallons 490.5 gallons Therefore, 490 gallons of water from the reactor following a large LOCA is added to the SIRW tank.
Appendix B 8'
REACTOR PRESSURE PSIA 203 198 189 178 165 147 128 95 59 23 15 LPSI RECIRCULATION FLOW TABLE 2 PUMP HEAD FLOW VELOCITY MASS FLOW P, FT v 2 FT/SEC W2 GPM 410 9.81 220.1 400 9.66 216.8 398 9.63 216.1 380.
9.35 209.9 370 9.2 206.4 360 9.03 202.8 350 8.87 199.1 315 8.28 185.8 280 7.64 171.4 250 7.04 158
Appendix B 9,.
REACTOR PRESSURE PSIA.
1238 1174 1069 923 562 160 15 HPSI RECIRCULATION FLOW TABLE 3 PUMP HEAD FLOW VELOCITY MASS FLOW P, FT FT/SEC GPM 2800 4.48 40.3 2700 4.40 39.5 2600 4.32 38.8 2500 4.23
- 38.
2000 3.77 33,8 1500 3.24 29.1 1325 3,03 27.2
Appendix B 10 TIME SEC 0
2 4
8 12 16 20
- 40.
80 120 200 REACTOR PRESSURE PSIA 1425 1175 1000 825 600 300 100 55.7 49.1 44.9 39.3 TABLE 4 LPSI FLOW HPSI FLOW GPM GPM 230 40 230 39.5 230 38.4 230 36.9 230 34.2 230 31.1 187.8 28.3 110.2 27.7 167.7 27.6 166.2 27.6 164.1 27.5
- These numbers are values of containment pressures at the given times. It is taken from CONTEMPT for a 0-6 DEG/PD, RUN BILLOEB.
f l
Appendix B 11 TABLE 5 To calculate total recirc flow from sump add flows of one LPSI and two HPSI's.
SUMP*
INTERVAL TIME RECIRC FLOW INTEGRATED TOTAL ADDED TO SEC GPM FLOW, GAL SIRW, GAL 0
310 10.3
.10.3 2
309 10.3 20.6 4
307 10.3 30.9 8
304 20.5 51.4 12 298 20.3
- 71. 7 16 292 19.9 91.6 20 244 19.5 111.1 40 226 81.3 192.4 80 223 150.5 342.9 120 221 148.7 491.6 200 219 294.7 786.3 Two HPSI and LPSI from Page 10.
APPENDIX C CALCULATIONS OF RADIOLOGICAL EFFECTS AS RESULT OF HYPOTHETICAL ACCIDENT WITH CV-3030 OPEN
Introduction
Since the opening of CV-3030 would no~ in itself prohibit ECCS operation, it is.assumed that only primary coolant activity at the Technical Specification limit of 1 microcurie per gram dose-equivalent I-131 is involved in recir-culation to the SIRW following a postulated DBA.
Two aspects of this event are considered: 1) Release of iodines to the environment via the SIRW tank vent following addition of 490 gallons of sump water to the ta.Iik via the recirculation line; and 2) Dose to control room personnel from the 6" recir-culation line during the 130-second period it is filled with a mixture of 48% SIRW and 52% containment sump water.
For the sake of completeness, doses from MHA fluids circulation also has been determined.
MHA dose parameters are readily available for control room dose calculations, so the MHA calculations were utilized in determination of DBA doses by appropriate scaling of the nuclide inventory. Offsite doses were calculated independently by use of iodine inventories for each_ case.
Total body doses from noble gas were not calculated because coolant from the sump would be lacking in noble gasses due to degassing upon release from the primary system.
Results - DBA Maximum offsite dose of.52 millirem to thyroid is calculated from release of 1.86 mCi dose equivalent I-131. This release resulted from a transfer of 490 gallons_of undiluted primary coolant to the SIRW, 100% of which mixes with SIRW water. The tank empties to a 2-foot level in a maximum of 20 minutes.
No iodine escapes during pump down, since airflow through the SIRW vent is inward at that time.
One percent of the available iodine inventory escapes to the atmosphere within the next two hours (similar to a fuel pool accident described in Regula-tory Guide 1..25).
Dose was calculated in accord.ance with Regulatory Guide 1.25.
Dose at the controi room console is calculated to be less than 0.01 millirem integrated over the two-minute_ period during which primary coolant is flowing through the 6" recirculation pipe outside the control room.
The dose is lov primarily because two feet of concrete is present between the piping and control room interior. Dose from the SIRW tank is negligible because concentration is very low once diluted in the SIRW volume.
Also, a minimum of 4 feet of concrete separates the tank from the control room.
2 Results -
MHA Offsite thyroid dose of 35,6 rem at the site boundary is calculated with 25% of core iodine inventory diluted by l.2xl05 gallons of fluid available to the sump within the first few seconds of the DBA.
The q_uantity of iodine released is 126 Curies.
All assumptions are similar to those described for the DBA case.
Dose at the control room console is calculated to be 0.23 rem due to liq_uids from the sump present in the recircul.ation line outside the control room.
E:h."J)csure from the SIRW itself is negligible since dilution in SIRW water is la.i:-e;e, only 10% of the activity remains after 20 minutes. and the control room is shielded by a minimum of 4 feet of concrete in that direction.
Conclusion Neither control room habitability nor offsi te doses are seriously affected by opening of CV-3030, since in all cases represent small fractions of 10CFR50 and lOCFFJ.00 limits.
It must be emphasized th~t doses have been calculated in a conservative ma..~ner.
In particular, the MHA fission product inventory based on TID 14844 greatly exceeC.s t!-1e inventory actually expected in the first few minutes of an accident.
For exalli:ple, WASH-1400 worst-case accident descriptions indicate that final gap activity begins to escape the core only after one minute, and core melt occurs only a~er 16 minutes.*
e TABLE I (RADIOLOGICAL)
FLUID VOLUMES DI~UTING FISSION PRODUCT INVENTORY PRIMARY COOLANT VOLUME 7,, 800 FT3 1/2 CLEAN WASTE RECEIVER TANK
. 4_.065 FT3 SAFETY INJECTION BOTTLES (ll) 4,000 FT3 PRE-EXISTING SUMP VOLUME 304 FT3 TOTAL DILUTION 16_.169 FT3 MHA IODINE - 131 CONCENTRATION CORE INVENTORY AT 2650 M~~T = 6,65 X 107 CI 5.835 X 104 GAL 3.041 X 104 GAL 2.992 X 104 GAL 2.275 X 103 GAL 1.210 x ios GAL r
25% CORE INTO COOLANT AT T = 0 ~ 1.66 x 107 CI I-131 -) 137 Ct/GAL DOSE EQUIVALENT 1-131 ~ 258 CI/GAL
\\
(*
.,/
FIRST 20 SECONDS - DosE EQUIVALENT 1-131 EVALUATION
~. ------------~ ------- rn f'--_
. 1;". I\\
~ I FL.UID ACTIVITY IN= (QnOOJ. C!
71CLEAN 11 PIPING CONTENT FLUID Acr1v1rv Our = AssuMED ZERO i-UIID INVENTORY = <0.1 CI RELEASE TO ATMOSPHERE*- NEGLIGIBLE
<A1 R FLm'1I NG hno TANK) 1
2 20 SECONDS TO 150 SECONDS DOSE EQUIVALENT I-131 EVALUATION.
...--~~----~-----
lA f:;=_
~; \\
- I I
~
- o tt 0
~
o?
u-2 0
<---. 22'
_,---- -~
a
~
v-I...
t
- )~ '. *
- .. *.... :.:*.'--:::: ~
20' 'NATER LEVEL
--~
r*LuID.~cr1v1rY IN = l.26xlo5 Cr. --.:.:. 0.50 C1/GJ-\\L l __ _
~---l>
F~.uxn Plcr1vrrv *our = AssuMED ZERo 5
...... ~. --*==<--
Fi.1~1D irivENTORY = l.26xl0 CI--~ 0,50 CI/GAL RELEASE TO ATMOSPHERE ~ NEGLIGIBLE (AIR FLOWING INTO TANK)
/\\CT IV I TY
,~CTIVITY IrJVENTORY 3
150 SECONDS ro 20 MINUTES*-.. DosE EQUIVALENT l-131 EVALLiATI°ON
.., 'e
~
t '...
q IN ;:; ZERO Our = l,.13x105
= l.26xl04 C1 I
f *
~
\\
\\
CI @ 0.5 CI/GAL
@ 0.5 CI/GAL I
0
~20'
\\..
0
'\\
~
RELEASE TO ATMOSPHERE = ZERO CArR FLOWING INTO TANK)
WATER LEVEL
. e r1 ;: In A."
if**
~-UJID 4
20 MINUTES To 2 HouRs PLus 20 MINUTES ~ DosE EQUIVALENT 1-131 EVALUATION 0
~
~.
d..
ca
/-\\CTIVITY
~
Acr1v1rv Our = ZERO INVENTORY = l.25xlo4 CI
~...
f......
0 I t
___________ tr I 0.
-. r.{.
- . [ i
~
- cc
~
'lo.
ti
- 0
\\ * *
\\,,
Q
- 0 e
I
'
- I 2' 'vV ATER LEVEL l
RELEASE ro ATMOSPHERE = 126 Cx D.E, 1~131
DOSE TO THYROID AT SITE BOUf'JDl\\i~Y -
MH.~
DosE = (Q) CX/Q) CB) Cr) CDCF)
WHERE:
DosE = REM Q
= CJ/SEC RELEASE RATE (126 CI/7200 SEC) = 0.0175 CI/SEC DOSE EQUIVALENi !-131 X/Q
= SEC/M3 DIFFUSION (5.5xlo-4 SEC/M3 PER R.G.1.4 AND AMENDMENT 31)
B
= M3/HR BREATHING ~ATE Cl.25 M3/HR PER R.G.1.4)
T
DCF = DosE CoNVERSION FACTOR Cl.48xl06 RAn/C1 INHALED~ PER R.G.1.4)
RESULT:
35,6 REM DBA RESULT - PRIMARY COOLANT AT laO JJCI/ML DosE EQUIVALENT I-131: 0.52 MILLIREM
CotlTROL RooM HABIT/\\BfLITY 0 6" RECIRCULATION LINE EXPOSURE DETERMINED TO BE ONLY SIGNIFICANT CONCERN 9 2' CONCRETE WALL BETWEEN LINE AND CONTROL R06M Q INTERVENING EQUIPMENT AND DISPLAY PANELS NOT INCLUDED AS SHIEL_pIN~
8 15' TO 20' DISTANCE BETWEEN SOURCE AND OPERATOR AT MAIN CONSOLE 9 LINE FILLED WITH
- 1)
MIXTURE 52% SUMP FLUID; 48% SIRW TANK WATER
- 2)
IODINE CONCENTRATION AS USED WITH OFFSITE DOSE CALCULATIONS
- 3)
PARTICULATE CONCENTRATION BASED ON 1% OF CORE INVENTORY TO SUMP 6 EXPOSURE.DURATION 130 SECONDS RESULT:
<0~01 MREM
('!*..,.
.,,1;
- ~ ~**~ ;, I
~
I '"*
'* "\\'
{'
.\\
f /:.,i
~...
~ : *..
.i.:
APPENDIX D CONTAINMENT PRESSURE RESPONSE ASSUMING LOSS OF TRAIN A SAFEGUARDS PUMPS
A1Jpendix D 1
Problem For the period of time during which a flow path exists between the containment sump and ECCS Train B, the sump will be filled with hot water from the break.
The affect of hot water spray on the previous results is analyzed here.
To be conservative, loss of Train A will be assumed.
Solution A reanalysis of the containment response to a 0.6 DEG/PD LOCA was performed using CONTEMPTLT-26.
This analysis assumed only two (2) containment spray pumps and one (1) air cooler.
Both containment spray pumps are on Train B.
With only two containment spray pumps operable the total spray flow is 2680 gpm.
Thus, the total flow from the sump will be different than that calculated in Appendix A.
The LPSI and HPSI flows as listed in Tabie 1 of Appendix A will remain the same.
In Table 1 of this section the flow from t.he sump as a function of time since the LOCA is listed.
Note that in 20 seconds 1280 gallons have been pumped.
Thus, it can still be conservatively assumed that the 1410 gallons in the sump lines is pumped out in 20 seconds.
Since the LPSI and HPSI recirculation flows do not change, 490 gallons of water from the LOCA is still pumped to the SIRW tank -- as long as the 150 seconds that Train B is connected to the sump is still bounding~
As reported in LER-.80-003 the two containment spray headers are always filled with water up to elevation 735.
As shown later this amounts to 3165 gallons of water in the lines prior to the LOCA.
Thus, "hot" water does not enter the containment from the sprays until the "cold" water has passed.
.Appendix D 2
Time until addition of hot water The volume of containment spray piping above elevation 735' is 1166.8 gallons (actually this is the amount of water required to fill pipes).
Thus, at a flow rate of 2680 gpm it takes to fill this volume:
1166.8 gal 2680 gpm x
60 sec min
= 26.1 sec The time from the break that this vollUile is filled is:
t = time to CHP + time for plllilp full flow + time to fill lines
= 1 sec + 4 sec + 26.1 sec
= 31.1 sec Therefore, spray starts to enter the containment 31.1 seconds after the break.
After this time the volume of cold water in the lines is added to the contain-ment.
The time for all of this water to be sprayed is:
3165 gal 2680 gpm x
60 sec min
= 70.9 sec This is the amount of time that the cold water in the plllilp discharge lines will be spraying.
However, as stated before, the pump suction lines are filled with 1410 gallons of cold water prior to the break and it takes over 20 seconds before this is pumped out.
Thus, the total time that cold water is sprayed is:
70.9 sec time to spray discharge volume 20 sec time to spray suction volume 90.9 sec time cold water sprayed
A,Ppendix D 3
Since it takes 31.1 seconds before spray water enters the containment, hot water spray does not begin for 31.1 + 90.9 = 122 seconds after the break.
The 0.6 DEG break was rerun using CONTEMPT to determine the affect of hot water addition after 122 seconds.
The results are plotted in Figure 2 where the containment pressure response is compared to the case previously analyzed.
Note that the net affect is to increase the time that the pressure is above 32 psig by about 20 seconds.
The deviation between the two curves above 32 psig is due to the reduction in air coolers and pump flow rate rather than the hot water.
The difference between the curves becomes greater once hot water injec-tion begins, but the pressure is below 32 psig at that point.
In any case, by losing Train A the containment pressure is above 32 psig for less than 150 seconds.
Thus, the use of that time remains valid.
Amount of Water in Containment Spray Lines The containment spray lines were modeled by Bechtel (JOB 5935 - 3/4 Mechanical Group Calculations Vol I & II). In Figure 1 is the nodalization of the spray lines from the Train B containment spray pumps.
The volume of water from the pumps to the isolation valves is 660 gallons as calculated in Table 2.
The containment spray water goes through the tube side of the shutdown cooling heat exchanger.
Per drawing Ml-G-D, sheet 167-3 the volume of water on the tube side of each heat exchanger is 660 gallons.
Thus, there are 1320 gallons of water in the. heat exchangers.
The total amount of cold water in the containment spray lines is 1185 gallons.
Thus, the total volume of cold water in the containment spray
Appendix D 4
discharge lines is:
660 1185 1320 3165 gal pumps -- isolation valve gal isolation valve -- el 735' gal Heat exchanger gal
.Aippendix D 5
Time since1 break, sec 0
2 4
8 12 16 20 1
0.6 DEG/PD TABLE 1 Total flow2 from sump, gpm 3260 3750 3830 3990 7630 2
1 LPSI + 2 HPSI (from Table 1, App A) + 2680
.Water removed from sump in interval, gal 0
250 255 266 509 1280 gal
Appendix D 6
Node1 12 13 14 15 16 17 18 19 20 21 22 24 1
see Figure 1 TABLE 2 Diameter, in 8
8 10 10 10 10 10 10 10 10 8
8 Length, ft 1 4
4 15 15 10 7
3 12 15 25
- 61.
25 Total Volume, ft 3 1.4 1.4 8.2
- ,8 *. 2 5.4 3.8 1.6 6.5 8.2 13.6 21.3 8.7 88.3 ft3 gal
e CALCULATION SHEET 0510 (4-681 SIG NA TU RE _
_,,.,.._,_*._.f-_C..:.__ _____ _
DATE. Ir ?.--11
.e.
~t.t-\\"
1~7)
TITLE _________
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JOB NO. f,'] ~;;.
su BJ ECT __
=C'-'-r--'-'-'-
"7'....:.'
,;:..._, '-'-:r*'-'-,
nw.i"'.::::-.....,t:..:...'-l_,__=5~1'"'i.::....
.. ?"""". t-:""""
5'--"'4;-"'"_._. -f.._..._e..,_,_a...._*
1 ____:_. _________
r t.
SHEET NO.
0
,,,7
. C..".',,-
~
4:s::::.c..<,,1C.
~
~
- . 9 25'.
1,
- 6.
..5;.,(111p-,
~
\\.. *
- i
- :;
<:l-=-===;--=------¥,,.
- 7.
24" P-664
?-&7A
ti*
ii
- ~.
10 X 10 TO THE CENTI METER KEUFFEL & ESSER CO.,MADE IN U 5 A.
l-1+11-R1-!.LH 18 X 25 CM.
TIME/SEC 46 1510
. _- )'
Cl>
1::$ p.
I-'*.
t:J
APPENDIX E OFFSITE DOSE AND CONTROL ROOM HABITABILITY ASSUMING LOSS OF TRAIN A SAFEGUARDS PUMPS
Introduction
Calculations have been performed in accordance with data derived from the analysis given in Appendix D; 490 gallons of containment sump water is delivered over the period between 20 seconds and 150 seconds following a LOCA.
During this period, the recirculation line contains 100% sump water as opposed to 52% as in the Appendix_C analysis where both ECCS trains were f'unctional.
A second variation from the Appendix " C calculations arises during the first 150 second9 when the SIRW tank receives a net inflow
- of fluid because ECCS train A is not f'unctioning to remove water.
The net inflow allows airborne radioiodine release during this period as well as during the two hours a:fter the tank empties to its shut-off level of two feet.
Results Site boundary thyroid dose for the DBA is calculated to be 0.57rem due to release of 2.03 mCi dose equivalent I-131. Site boundary thyroid dose for an MHA is 38.8rem, and results from release of 137.4 Ci of dose equivalent I-131.
Both the DBA and ~IBA doses are a factor of 1.09 times the result of the earlier two train analysis. The increase of 9% is due to the release of activity over the 130 second fill period at a release rate of 1% per two hours (0.000139% per-second), given an average concentration of 0.25 curies/gallon over that period.
Dose at the control room console is calculated to be 0.44rem.
This increase over the two-train dose of 0.23rem arises from the increase in recirculation pipe consentrations from 52% to 100% of sump concentration in the single train event.
Conclusion Doses continue to represent small fractions of 10CFR50 and 10CFRlOO limits.
A large degree of conservatism remains in the methods by which.these doses are calculated.
For further disscusion see the conclusion section of Appendix C.
APPENDIX F EFFECT ON MAIN STEAM LINE BREAK AT 1400 MWD/MTU
.A'.ppendix F 1
I.
REFERENCES
- 1.
XN-NF-79-94(p) "Palisades Cycle 4 Startup Predictions and Nuclear Data for Operation".
- 2.
"Palisades Cycle 4 Startup Data, Supplementary Information",
R.G. Grummer to B.D. Webb, November 30, 1979.
II.
DATA
- 1.
Net worth (N-1) at 6o°F Reference 2, Table 2 BOC4
=
3.63% Ap,
EOC4
=
4.09% Ap
- 2.
Power defect at 700 ppmb
- - Reference 1, Figure 6.5
= 1* *. 2% Ap
- 3.
Net.rod worth (N-1) at 532°F
- - Reference 1, Table 6.1
- 4.
- 5.
BOC4 = 4.90% Ap EOC4 = 5.49% Ap Shutdown boron concentration, keff Reference 1, Table 6.2 532°F BOC4 1000 EOC4 150
=.98, No xenon -
N-1 Configuration.
60°F 1050 500 Core conditions 1400 Mwd/MT Cycle burn up 700 p:pmb
- 6.
Reciprocal boron worth Reference 1, Figure 6.6 III.
ANALYSIS 1000 ppm, 6o°F, BOC = 77.0 ppm/% Ap 150 ppm, 6o°F, EOC = 70.5 ppm/% Ap
- 1.
Worth of control rods 6o°F including uncertainty.
BOC = 3.63 x.90 = 3.27% Ap EOC = 4.09 x.90 = 3.68% Ap
. Interpi\\'Hating to cycle burnup of 1400. Mwd/MT worth (N:..l) =
3.27 +
1400 (3.68 - 3.27) = 3.32% Ap l0,4oo
- 2.
Reactivity added by cooldown.
This is derived by extracting the change in net rod worth due to cooldown from the change in shutdown boron concentration from hot to cold conditions.
Appendix F
~
a.
At 1000 ppm l].050 - 1000 - (4.90 - 3.63) x 7f!/77 = 0.52% ~P b.
At 150 ppm
!500 - 150 - (5.49 - 4.09) x 70.iJ/70.5 = 3.56~ ~p c.
Interpolating to 700 ppm
~P700 = - 62 + lOOO - 7oo (3,56 +.62) = 0.86% ~p 1000 - 150
- 3.
Shutdown Margin.- Equals rod worth minus power defect minus reactivity from cooldown.
3,32 - 1.20 - o.86 = 1.26% ~P CONCLUSION The Palisades reactor could have cooled all the way to 6o°F without boron inejction and remained subcritical until xenon decay.
There is adequate margin in the analysis to account for large uncertainty factors.
APPENDIX G EFFECT ON MAIN STEAM LINE BREAK AT END-OF-CYCLE
Appendix G 1
PURPOSE EFFECT OF MAIN STEAM LINE BREAK AT END-OF-CYCLE To determine the effect on the return to power (and MDNBR) following a MSLB with CV-303CLopen.
BACKGROUND The large inside break from full and zero power was reanalyzed for the Palisades
.therinal uprating to 2530 Mwt in Exxon Nuclear Company Report XN-NF-77-18.
This analysis, as the analysis in the FSAR, assumed two HPSI pumps available for in-jecting boron into the primary coolant loop.
With CV-3030 open and the contain-ment pressurized in excess of 32 psig, HPSI pumps P-66B and C would be pumping unborated sump water rather than borated (1720 ppm) SIRWT water.
In this case only P-66A - the HPSI pump on the redundant train - would be available for boron injection.
The injection flow from P-66A as compared to that assumed in the safety analysis is shown in Figure 1.
The safety analysis assumed both pumps to be pumping through the same HPSI header, thus flow from two pumps is not two times that from one pump.
METHOD OF ANALYSIS
- 1.
Estimate difference in integrated HPI flow from one or two pumps up to time of peak return to power.
- 2.
Determine resulting difference in PCS boron concentration.
- 3.
Determine reactivity difference due to boron difference.
- 4.
Determine impact on core power..
- 5.
Determine impact on MDNBR.
. ; *...j 3
0 u..
1000 500 oL-...:...i..~-J.....~..1.---1~~~......L-.--~-..J~-1..--:~00
~-l-__,,-+-_._~--~1soo 0
500 10 Primary System Pressure (psia)
'FIGURE 1 *HIGH PRESSURE SAFETY INJEC.TION FLOW VS PRESSURE x z I :z,,
I
"'-J
"'-J I.
,J j
.. Appendix G 3
FULL POWER CASE Ref:
XN-NF-77-l8, page l20 Time of SIS (.-l600 psia) -
l8 seconds Valve stroke time
,._ lO seconds Peak return to power time""'ll5 seconds (when steam generator dries out) l. Calculate RPI Flow to ll5 Seconds TIME TWO PUMP ONE PUMP SECONDS PCS PRESSURE FLOW FLOW INTEGRATED DIFFERENCE 28 800 psi 700 gpm 430 gpm 0 gallons 40 600 790 5l0 55 60 440 860 565 l5l 80 400 880 580 250 100 380 890 585 351 120 380 890 585
~
Pressure taken from Figure 3.87 of XN-NF-77-l8; two pump flow from Figure 3.83 of XN-NF-77-18; one pump flow from Attachment l, Sheet l of CE letter P-CE-4538; integrated difference was calculated assuming a constant flow rate differ-ential between time points.*
- 2.
Calculate Difference in PCS Boron Concentration Boron concentration in SIRWT = 1720 ppm Density of SIRWT water = 62.3 lbm/ft3 1 ft3 = 7.48 gallons
\\
,, Appendix G 4
Approximate mass of coolant in PCS = 454,ooo lbm (RE: PCS Functional Desc No M-10, page 80)
Assuming uniform mixing of boron in PCS - a reasonable assumption with pumps running as loop transit time is only 10 seconds - the difference in loop boron concentration at 120 seconds is about 1720
((452 gal) (62.3 lbm/ft3) (7.48 gal/ft3)\\
ppm 454,ooo lbm
}
~ 14 ppm
- 3.
Determine Reactivity Difference Due to Loop Boron Concentration Difference Boron Worth =.0001 ~p/ppm (Page 10 of XN-NF-77-18)
- . ~p = (14) (.0001) =.14%
- 4.
Impact on Core Power Doppler reactivity vs core power curve assumed in analysis - Figure 3.82 of XN-NF-77-18.
Slope of curve between 0 and 30% power is roughly 1% ~P/30%.
. 14% ~P is equivalent to 4.2% in core power.
5, Impact on MDNBR In order to evaluate the impact of a 4.2% increase in core power on MDNBR, need to know 8MDNBR for the stuck rod power distribution.
A very rough 3P
- estimate of 8MDNBR can be derived from Exxon Nuclear Company letter dated ap 1/2/80.
Referring* to Table 1 for a.46% increase in core heat flux and a 0.5°F decrease in core inlet temperature, Exxon Nuclear Company predicts a
.7% decrease in MDNBR.
Assuming the inlet temperature change is negligible
- '*Appendix G 5
(at full power a 1% change in power has 3 times the effect on MDNBR than does a 1 °F change in temperature),
3MDNBR.
aP
= ~4~ = -1.5%/%
Thus a 4.2% increase in heat flux is roughly equivalent to a 6% decrease in MDNBR.
Zero Power Case REF: XN-NF-77-18, page 120 Time of SIS = 15 seconds Valve stroke time = 10 seconds Peak return to power time = 95 seconds (when charging pump flow reaches core)
(same procedure as for full power case)
- 1.
Calculate HPI Flow to 100 Seconds PCS*
TWO PUMP ONE PUMP TIME PRESSURE FLOW FLOW INTEGRATED DIFFERENCE 25 sec 940 psi a 580 gpm 370 gpm 0 gallons 40 690 750 480 60 60 570 800 525 151 80 570 800 525 243 100 570 800 525 Gill
- Figure 3,93 of XN-NF-77-18.
'.. Appendix G 6
- 2.. Difference in Boron Concentration 1720 ppm
( ( 334) ( 62. 3) I ( 7. 48) \\
11 ppm 454,ooo
)
- 3.
Reactivity Difference Due to Loop Boron Concentration Difference b.p = ( 11) (. 0001) =.11%
- 4.
Impact on Core Power
(.11%) (30% power/1% ~~) = 3.3% power increase
- 5.
Impact on MDNBR (3.3%) (-1.5
%MDNBR/%Power = -5% MDNBR
.02
<J I
.µ
.µ u
ttl QJ 0::
i:::. 01 QJ t::;;)
i:::
ro
..i:::
u 0
20
@. 30%
40 60 80 100 Power {% of 2530 MWt)
FIGURE 3.82 VARIATION OF REACTIVITY WITH POWER AT CONSTANT CORE AVERAGE TEMPERATURE 120 140 i.......
N w z
I z,,
I
-....J
-....J I
1 I
I e
e