ML20210R294
| ML20210R294 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/29/1986 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Keller R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20210R261 | List: |
| References | |
| NUDOCS 8610070230 | |
| Download: ML20210R294 (47) | |
Text
11RCr 191 C/?
&Vi T#
Telephone (412) 393-6000 Nucteer Group P.O. Box 4 Shippingport. PA 15077 0004 July 29, 1986 Mr. Robert M. Keller, Chief Section 1C (Operator I.icensing)
Division of Project and Residen: Programs U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406
REFERENCE:
Eeaver Valley Power Station, Unit #1 Docket No. 50-334, License DPR-66 Written Examination Coments
Dear Mr. Keller:
4 Please find enclosed comments and re.ference material associated with our Training Section's review of the written examination administered by Region I exaciners on July 22, 1986. These comments have been discussed with your personnel during the exam review conducted July 28, 1986.
4 i
Very truly yours, J
ce President uclear Group cc: S. Barber (NRC)
T. Burns Central File (2) 8610070230 860923 PDR ADOCK 05000334 V
t I
COMMENTS ON NRC RO EXAM 7/22/86 Ouestion Comment 1.04.d The answer should be: ACP the same as the ECP.
During a reactor startup there is no steam flow through the turbine and therefore, no extraction steam flow to the feedwater heaters.
Taking a string of heaters out of service at this time would have no effect on feedwater temperature.
1.05.a The question asked for a description of the reactivity changes that must be compensated for, but did not ask h2E this would be done.
However, the key requires an explanation of how these changes would be compensated.
This should not be required for full credit. Also, the time to equilibrium samarium that we teach is approx. 14 days or 335 hours0.00388 days <br />0.0931 hours <br />5.539021e-4 weeks <br />1.274675e-4 months <br />, not 400 - 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />.
1.06.b The answer is not correct.
The flux spectrum is shifted from the thermal to epithermal range as boron, a pure thermal absorber, increases. This will cause more competition for the cadmium but less for the silver and indium. Which shift is more pronounced is not known. The concept of rod worth vs. boron is not presented in the reactor theory manual as referenced.
It is covered in a supplemental handout.
This handout states rod worth decreases as boron concentration increases due to increased competition.
I 1.06.c Although may factors affect rod worth (boron, temperature, FPP's), the predominant factor is always the relative flux seen by the control rod. Over core life, this relative flux will change as it shifts away or toward rodded assemblies. There is no single corriet answer for all cycles.
Each cycle must be looked at separately. For cycle V rod worth goes down over core life, but for cycle VI it goes up (Table 6.4 cycle V and VI).
The correct answer for this exam is it goes down as flux shifts away from rodded assemblies.
l BVPS Reactor Theory Manual, Chapter 8, pages 14-16, BVPS Core Design Reports Cycle V and VII.
I 1.08.b The ratio of core flow to total loop flow will either decrease i
or remain the same after tripping 1 RCP.
The direction depends on the amount of reverse flow, and the increase in operating loop flows.
If the increase in operating loop flow is enough to compensate for the reverse flow, the ratio will remain the same.
If it isn't, the ratio will decrease.
Based on our simulator, the operating loops will in-crease to approximately 115% and the idle loop will read approximately 30% reverse flow. These instrements and their scales make it very difficult to get an accurate reading.
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Question G.9Eugent 1.08.b (cont) Without accurate values, it is not possible to tell if the ratio will decrease or remain the same.
If it does decrease, it will be a small decrease but no determination can be made unless specific values are given, which the question does not provide.
1.08.d There is an error in the key.
It should say "less flow - AAan heat removal".
2.01.a Excess letdown comes from the three loop drain valves on the cold legs. None of the choices in column B is correct.
2.02.a The question asked to explain yhy RCP seal flowrate varies, but not yhigh direction it changes as pressure changes. The second sentence in.the key should not be required for full credit.
2.06.b The answer key is incorrect in two places. The loy,1c for low-low levels in the SG's is 2/3, not 2/4. The second condition should be " Trip of all running Main Feed pumps".
2.07.a Another source of H in containment post-LOCA, is the H 2
2 that was in solution in the RCS.
This should be acceptable as one of the three sources asked for.
2.08.b The Main Feed bypass valves get. no direct signal from Safety Injection; however, a safety injection initiation will result in a Feedwater Isolation signal which closes these valves.
Depending on how the question is interpreted, either NO or CLOSE would be a correct answer.
2.09.b This question is the same as question 6.03.b.
The two answers are not the same. Either wording should be acceptable for full credit on either question.
3.03.a Immediately after the trip, the SUR meters will be pegged low.
3.03.b The exact correlation between SR counts and IR amps is not required to be known from memory and shouldn't be required for full credit.
3.04.b The candidate would have to be able to calculate whether or not this ramp would produce an output equivalent to a 154 step load change to answer this question. This is not required knowledge, and we ask that the question be deleted. The answer key is in-correct. A 7.5t/ min. ramp would take 10 minutes to arm the Steam Dumps.
3.04.d The key is incorrect. There is no deadband on the Reactor Trip controller. Therefore, the answer should be " arm and actuate".
Question Comment 3.05.b Since the question does not specify whether or not the urgent alarm is from the Logic Cabinet or a Power Cabinet, automatic actions resulting from either should be acceptable.
The key only addresses an urgent alarm in a Power Cabinet.
3.08.c There is no one correct answer for this question.
For a AI between +11 and -234, the answer is REMAIN THE SAME.
For a AI more negative than -23%, the answer is DECREASE.
3.09.a There is only one signal other than High-High S/G water level that causes a feedwater isolation signal.
That signal is Safety Inj ection. This should be acceptable for full credit.
3.09.b Turbine trip should be added to the key.
3.10 The key is incorrect.
Letdown isolates at a pressurizer level of 144, not 174.
4.01.a A low-low rod insertion limit alarm and an ATWS are conditions that require emergency boration and should be added to the key.
4.02.b There is no condition in which you can violate a procedure, Tech Spec, or license condition without SRO approval.
The procedure referenced in the key is written for a condition for which no procedure exiscs, not the conditions set forth in the question.
The question, as written, is invalid and we ask that it be deleted.
4.03.b There is a typo on the key.
The first answer should read, " Main Generator breaker - OPEN".
4.04.c One of the symptoms on the key is incorrect.
It should read,
" primary side temperature >550*F", not "Tave >550*".
4.07.c An increase in the amount of krypton would be indicative of j
failed fuel. This should also be an acceptable answer.
4.08.c The key is f.ncorrect.
The maximum SUR is 1 dpm.
4.08.e Thekey{gincorrect.
The SR Hi 9 trip is blocked above P-6 (1 x 10~
amps).
4.09.a The attachment number should not be required from memory.
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i COMMENTS ON NRC SRO EXAM 7/22/86 Ouestion Comment 5.10.a The key mentions power defect when it should say doppler defect.
5.10.b There is a math error in the key.
The answer should be 575.4*.
6.02.a The reference for the answer was misinterpretted.
The sample system also has delay coils to allow decay of N-16, therefore, this is not an advantage. There is no correct answer to this question and we request that it be deleted.
6.02.b RM-DA-100 would indicate activity in the drains from the Aux.
Feed pumps and should also be an acceptable answer.
6.03.a The range of the CCR flow indicators in the Control Room are 0-1400, 0-2500 and 0-5000 gpm.
It would be almost impossible to detect a 10 gpm leak on these meters. The best way to locate the leak from the control room would be by using component temperature alarms and sump alarms.
6.04.c Another acceptable answer for when Alternate Dilute is used would be when a more rapid p change is desired.
6.05 The instrument failure presented in the question would not lead to a reactor trip. A high failure of PT446 would fail Tref to 577*.
Rods will withdraw to raise Tavg to 577*.
Reactor power will not go much higher than steam demand.
Since the question presents an incorrect situation, we request that it be deleted.
6.06.a The valve number on the key should be PCV-CH-145.
6.09.d The answer key is incorrect. When containment pressure drops below the CIB setpoint, the CIB reset signal is cleared.
If i
pressure goes back above tho CIB setpoint, another automatic start of the quench spray pumps will result.
6.10.c The key is incorrect. Answer 1, 2, and 3 occur simultaneously.
7.01.b The second sentence in the key is not asked for the question and should not be required for full credit.
I 7.02.a The Heat Sink red path should be < 5% in all SG's.
7.02.c The question asked for parameters, not setpoints. Numbers
(
should not be required for full credit.
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7 d
Question Comment 7.03.a The key is not complete.
"Run back the turbine" and "close the main steamline bypass valves" should be added to the key.
7.03.c Same comment as 7.02.c.
7.06.a This question asks that a step in a normal operacing procedure be repeated from memory.
This is not required knowledge and we ask that the question be deleted.
7.07 This question requires memorization of a procedure.
This is not required knowledge and we ask that the question be deleted.
7.09.c Time limits were not asked for in the question and are not required knowledge and should not be required for full credit.
.7.10.c Same comment as 4.07.c.
7.10.d The referenced Tech Spec has been amended.
The key should read
" Operation may continue, provided the activity does not exceed the limit on the Tech Spec curve."
8.03 The question asks'what action should be taken, not why.
The key requires both for full credit.
The first sentence should be deleted from the key.
8.10.b Since five (5) minutes have passed already, a candidate could answer 10 minutes instead of 15 minutes for the time limitation.
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COMMENTS ON NRC SRO EXAM 7/22/86 Ouestion Comment 5.10.a The key mentions power defect when it should say doppler defect.
5.10.b There is a math error in the key. The answer should be 575.4%.
4 6.02.a The reference for the answer was misinterpretted. The sample system also has delay coils to allow decay of N-16, therefore, this is not an advantage. There is no correct answer to this question and we request that it be deleted.
6.02.b RM-DA-100 would indicate activity in the drains from the Aux, Feed pumps and should also be an acceptable answer.
6.03.a The range of the CCR flow indicators in the Control Room are 0-1400, 0-2500 and 0-5000 gpe.
It would be almost impossible to detect a 10 gpa leak on these meters. The best way to locate the leak from the control room would be by using component temperature alarms and sump alarms.
6.04.c Another acceptable answer for when Alternate Dilute is used would be when a more rapid p change is desired.
6.05 The instrument failure presented in the question would not lead to a reactor trip. A high failure of PT446 would fail Tref to 577*.
Rods will withdraw to raise Tavg to 577*.
Reactor power will not go much higher than steam demand.
Since the question presents an incorrect situation, we request that it be deleted.
i 6.06.a The valve number on the key should be PCV-CH-145.
6.09.d The answer key is incorrect.
When containment pressure drops below the CIB setpoint, the CIB reset signal is cleared.
If pressure goes back above the CIB setpoint, another automatic start of the quench spray pumps will result.
6.10.c The key is incorrect. Answer 1, 2, and 3 occur simultaneously.
l 7.01.b The second sentence in the key is not asked for the question and j
should not be required for full credit.
l 7.02.a The Heat Sink red path should be < St in all SG's.
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7.02.c The question asked for parameters, not setpoints. Numbers i
should not be ret;uired for full credit.
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J.O C. a V.'
Xe'non (Fig. 26 & 27)
A.
Values - pem 100% equilibrium
-2750 100% peak
-4900 50% equilibrium
-2250 50% peak
-3050 B.
T4mes - hours 100% equilibrium 40-50 hours 100% peak 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after trip return to 100% equilibrium 24-26 hours after trip return to 0% concentration 80-90 hours after trip VI.
Samarium - (Cycle 5 data unavailable at this time but values should be imilar)
(Fig. 28)
A.
Values (pem)
Cycle 4 Cycle 5 i
100% equilibrium N - 610 pcm same values may 100% peak N -1010 pcm be used for Cycle 5 50% equilibrium N-610 pcm i
~
B.
Times (days) y 100% equilibrium
% 14 days
}
1, 100% peak
% 14 days VII. Secondary Source - Remain in Locations H-3 & E13.
(Fig. 4)
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VIII. Burnable, Poison - 72 BPRA's - 12 with 20 rods (Fig. 4) 24 with 16 rods l
8 with 12 rods 12 with 8 rods 16 with 4 rods total 880 rods IX.
Delayed Neutron Data BOL EOL (See Table 5)
S
.00629
.00507 I
.97
.97 1
(S x I)
S,gg
.0061
.0049 l
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- l. C&. C
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WESTINGHCUSE PROPRII"ARY C* ASS :
i TABLE 6.4 CALCULATED RCD WORTH AND SHUTDCWN MARGIN - CTC*.,Z 5 Wrth Design Shutdown Wrth I,ess 10 Percent Reg 2irements Margin Condition h)
( 4a- )
(%a-)
(%i-)
BCL,H2P 48 Rods In 8.53 BCL, H2P 47 Rods In 7.56 6.81 2.59 4.22 ECL,H2P 48 Rods In 8.19
)
EOL,H2P 7.32 6.53 3.51 3.Of of Rods In,
c s
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~...
wasnmancuss momsTAny class 2
.J TABLE 6.4 C.M.CULATED ROD WORTH AND SHUTDOWN MARGIN - CYCLE 6 Design Worth Reactivity Shutdown Worth less 10 Percent Requirements Margin Condition
(%)
(%)
(%4o)
(74 )
BOL, HIP 8.35 48 Rods In BOL, HIP 7.42 6.68 2.64 4.04 47 Rods In EOL, HIP 8.52 48 Rods In EOL, HIP 7.56 6.80 3.54 3.26 47 Rods In e
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14S A S-ce6421 6-10
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\\c/pubh1 (thiH1-vas finally able to be started to sweep the gas from these regions. As a s
g direct result, considerable study has been conducted on the formation and effects 6f non-condensable gases in a Westinghouse.oressurized water reactor.
As discussed in previous sections of this program, the steam generators are w
needed to remove decay heat for a significant amount of time for the small break loss of coolant accidents.
For example, if the break size is about 2 inches in diameter, the steam generators are needed to remove decay heat for about one hour. After that time, the combination of the break itself and
~
safety injectici will remove all of the decay heat.
In the case of a one inch break, the steam generators are needed to remove decay heat for about one day.
Various computer runs have indicated that the most limiting case for the amount of non-condensables produced is a break size of two inches.
Specifically, we will take a look at this case, which also uses the following assumptions:
1.
3 loop plant
{
2.
2775 MWt 3.
17 x 17 fuel 4.
2 inch cold leg break l
S.
Feedwater available 6.
Minimum safeguards Sources of Non-Condensable Gases There are many possible sources of non-condensable gases, some much more significant than others. Most involve the production and/or release of hydrogen. Assuming no forced convection, if the gas is released in the core reg' ion, a bubble will collect in the reactor vessel head.
If the bubble grows to the point where it reaches the nozzles, it will leave the vessel and seek system high points. Sources of non-condensable gases include:
1.
Dissolution of H : As mass is lost throui,:h the break, pressure is 2
[
loweret, thus releasing dissolv'ed H
- ~*h' **
I'"* *N"*"9" "* th' 2
system.
For this analysis, all the non-condensables are assumed to be I
(
20 07845:4
released when coolant pressure reaches sa'turation, and initial H 2 g
concentration is assumed to be 50 cc/kg.
2.
Radiolysis: As you may recall, if the surrounding H C""C'"*#'*I 2
is low, a' water molecule in the coolant disassociates into its component parts of H and 02 ""d'" th' I"#I"*""' "#
- 9'""" ""
2 neutron flux.
r,n
~
2H O 2H2+O2 2
For this analysis, this process was assumed to start when the H 2 concentration rdached 5 cc/kg. Gas production due to radiolysis of water in the containment sump was assumed to be negligible.
3.
Pressurizer Vapor Space: Approximately '600 scf of H is e ntained 2
in the pressurizer, and it is assumed to expand into the reactor coolant system per the ideal gas law after the pressurizer empties.
4.
Zircaloy-Water Reaction: From a previous section, you know that the
[
amount of H liberated from this reaction is dopendent upon the 2
magnitude of temperatures in the core.
Computer generated core temperatures are used to determine the amount of gas generated.
5.
Accumulator Nitrogen: The N that provides the motive force for 2
l accumulator injected water is assumed to be released to the system iti the same way as the gas in the pressurizer would be released.
6.
Fission Gases: Those fission product gases generated within the ft.el
(
are assumed to be released to the system only if the fuel rods burst.
(
That prediction is included in the computer program.
7.
Helium: Each of the fuel rods is propressurized to a specified amount h
with He. As with the fission product gases, He is assumed to be h
released to the system only if the fuel rods burst.
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21 07845:4 9
,__-.-r___.-
,_.__-_m_
_.,__m-__
m.
8.
Dissolution of Gases from Safety Injection Flow:
For this analysis, all of the safety injection water injected into the primary system was
. assumed to be air saturated (18 cc/kg of various non-condensable gases).
As with many of the computer programs designed for accident analysis, certain conservative assumptions were used which simplify the analysis, and assure us that the results tend to be worst case.
For this analysis the following con-servative assumptions were used:
1.
Perfect mixing of safety injection and reactor coolant system (assures maximum amount of nort-condensables reach core).
2.
No reabsorption of released gases.
3.
All non-condensables dissolved in coolant are released as soon as the reactor coolant system pressure drops to saturation.
A breakdown of the results of the computer analysis are as follows:
k' Time Event 90 see
. Dissolution process begins as coolant flashes 144 see Radiolysis starts (H2 = 5 cc/kg) 190 see Pressurtzer. empty 1350 sec Zr-H O reaction begins 2
2100 sec Minimum pressure in RCS reached (pressurizar releases no more gas) 4050 see Break removing all decay heat At no time did the accumulators inject, nar was there any fuel red bursting.
The actual calculation of the volume of non-condensable gases produced yields some interesting results. During the time that the steam generators are needed to remove decay heat (4050 sec),1648.2 cubic feet-(at FTP) of non-condensables are produced. Corrected for actual plant pressure and temperature occurring during the accident, this results in about. 5G ic. feet. Each
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e TABLE 2-2 STEAM DUMP CONTROL SYSTI2i SETPOLVIS Load Rejection Controller 120 sec. h,
r, 10 sec.
r7 2 sec.
78 Deadband 5'T Proportional Gain in percent of total dump capacity per *T*
5 /
- F'
- 10. 6, $,20. 6,25* 7 Trip-Open Temperatures Sudden Lead-Loss Satpoints (permissives 15: 50% of full C TA and C-IB) oad s
d )d-Plant Trip Controllar Proportional Gain in percent of total dump capacity per
'7*
2.5 /*F i
Trip-Open Temperatur3s 11.1*F.
20*F Hesder Pressure Controller 1005 psig Set Pressure Proportional Gain in percent of total dump capacity per psi
- 0.25 / psi Raset Time Constant 180 sec.
Atmospheric Relief Valve Controllers Set Pressure 1035 psig Proportional Band (valve full stroke) 50 psi Reset Time Constant 180 cec.
[
- The total dump capacity consists of all thi condenser dump valves.
t 2-16 l
uln +,, L a 9lc0 % bf #!< a., /
- uses /o % 6 c' "M, AaA.,/
1+
+ n /S9., 7 %
<.A /
Eout = G(e"!T(Vp - V )3 t
when t = 0, Eout = G(V -V)
F I
when t = 3r, Eout = 0.
The plot on. Figure 3-4 uses a gain of 2, a step of 5 volts, and a time constant of 10 seconds. !!otice that the initial and final values of the step input do not affect the steady state output:
it is always zero.
The response to a ramp input is interesting.
Eout = G r(1 - e /') Rin t
when t = 0, Eout = G r(1 - 1) Rin = 0 when t - Sr, Eout = G r(1 - 0) Rin = Gr Rin Since r has units of seconds and Rin has units of volts /second, our steady state output has units of volts; that is, it is a constant DC voltage level.
The voltage depends on the card gain, the cimuit time constant, and the slope of the input ramp.
A There art several protection and control circuits which take advantage of the rate unit's response to a ramp input. One is the C-7 steam dump aming signal. This circuit interlocks the steam dump off until turbine power drops by 10 e rt:ent.
It has a econd time constant. We can translate the step Tc
./5 %
into an equivalent ramp.
c-M assurae 10 VDC = 120 percent turbine pressure 15 g percent = 10 ()tf/120) =.%VDC, the C-7 bistable setpoint taen IN
/. 9 6 d u c Let's find what ramp rate, in percent per minute, produces the same output as a y pert:ent step:
/$
spement = 120 sec (1 - e /120) R t
in
/3 Assume steady state conditions; that is,-pore than five time constants f
have passed:
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/$ gpercent = 120 sec (1 - 0) R i
in l
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LAHShm mt 9/0C Scalmy /t1dnue l h
y percent /120 sec = (percent /2 minutes R
in = /6
/3 or R, = f percent / minute 9
- 7. 3 Thus, if a f percent / minute ramp lasts for five time constants, or ten minutes, the C-7 bistable trips. Ramps of higher magnitude take shorter times to reach the trip setpoint.
1 + rNs Lead / Lag Unit:
1, 3
l{ere we have a circuit with two time constants: r g, the numerator or
" lead" time constant and r D the denominator or " lag" time constant. The circuit's risponse depends on the ratio of and 2 This can be seen on TN 0
Figure 3-5.
In both cases, the circuit's response to a step input is:
t/c D Eout = G((1 +
(*,.N,
, y) ;
3gyp,y3,y3 t
I D
This is quite an equation; let's analyze it. The ters:
h g]
t T N
[1 + ( T
- 1) e D
... is time dependent. When t = 0, it reduces to:
[1 + (
N,
, g) 3, T N T
TD D
When we substitute this back into the overall equation we get:
Ecut-G(N,(yp,y),y3 I
I D
(
This means the initial risponse is a step whose peak is dependent on the card gain times the ratio of the time constants times the size of the step.
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1.21.2 SETPOINTS_
510 PSIG Low Steam Line Pressure 50 see Lead time 5 sec Lag time
-99 PSIG Eigh Steam Pressure Rate 50 sec Lag time P-11 2000 PSIG 543F Lo-Lo Tavg e
Intermediate High Eigh Containment Pressure 5 PSIG (Steam Line Isolation)
Steam Dump Control Sudden Loss of Load arm first and second banks 15% of full load arm third and fourth banks 50% of full load 3.57% DUMP /I Gain, load rejection controller 2.5% DUXP/F Gain, reactor trip controller (50% dump is maximum) e Load Rejection Controller Deadband Load ejection gn avg-Tref) 10.6F first bank trip open 15F second bank trip open 20.6F third bank trip open 25 7 fourth bank trip open Reactor Trip High (Tavg-54TF) 11.17 first bank trip open 20F second bank trip open Header Steam Pressure Controller 1005 PSIG Set Pressure 400 PSI Proportis. Sal Band 180 SEC Reset Steam Generator Atmospheric Cump Valves 1035 PSIG Controller Setpoint Controller Proportional Band 50 PSI 180 SEC Controller Reset 1060 PSIG Pressure Switch Trip Open 20 IN. HG.
Low Condensor Vacuum Steam Dump Lockout 90 PSIG PCV-GN-102 PCV-GN-101 3-5 PSIG 110 PSIG RV-GN-101
, ISSUE I REVISION 1
i J.u i. b B.V.P.S. - 0.M.
1.1.1 INSTRUMENTATION AND CONTROLS (continued)
J Power Cabinet ICAB-RC/PWR-F/L1 Figure 1-46 shows the basic circuit of one Power Cabinet. Five, three-phase, half-wave, phase controlled thyristor bridges are used to control and regulate current to the control rod mechanisms. The five bridges are connected through three fused disconnect switches to an enclosed bus duct which runs along the top of the power cabinet.
Three bridges supply the four stationary gripper coils in each of the three groups. The fourth bridge supplies all three groups of four novable gripper coils through three multiplexing thyristors. The fifth bridge supplies all three groups of four lift coils and each of these lift coil circuits includes a sep,arate disconnect thyristor.
Each stationary gripper coil is connected in series with a diode (to prevent the flow of current from one coil through parallel-connected coils during a reactor. trip) and with a series sampling resistor which feeds a seasure of the coil current back to the phase controlled thyristor bridge.
l The novable coils are similarly connected except that each group of four coils is connected to a multiplexing thyristor used as a disconnect switch and the three groups of four coils are parallel-connected through four current sampling resistors.
The lift coils are connected.in the same manner as the movable coils except that each coil has a series thyristor which serves both for multiglexing and for disconnect.
i Multiplexinz Control Multiplexing permits rod motion on one of the three mechanism groups while the other two are held in position.
Multiplexing control is obtained by turning on the proper set of thyristors in the lift coil circuit (four fire simultaneously as shown in Figure l
1-44) and by turning on the proper thyristor in the movable l
gripper coil circuit. Stationary coil c,ircuit multiplexing is obtained by means of an additional input to the control circuits of the stationary phase controlled bridge thyristors.
The different method of multiplexing the stationary gripper coils is required because these coils provide holding power for the rods.
The two groups that have not been selected for motion must be profiled for reduced current to hold them in position.
As I
previously stated, each mechanism in the lift group is provided with a disconnect thyristor even though they are normally 4
sultiplexed in groups of four.
This.is done to allow anual disconnect of individual rods for test or retrieval of a d 'gyed rod.
3 4
i i
i ISSUE 2 REVISION 4
B.V.P.S. - 0.M.
1.1.1 INSTtUMENTATION AND ColfrROLS (continuedl I
Alarm Circuits provided for both " URGENT" and "NON-URGENT" failures.
Alarus are Any failure that affects the ability of the system to move rods is considered
" URGENT".
"URGENr" alarms are detected by the regulation failure detector, phase failure detector, logic error detector or the multiplexing error detector. This alarm performs the following functions:
1.
Automatically de-energizes the lift coil and energizes both I
the stationary gripper coils and the movable gripper coils at reduced current.
I l
2.
Energizes' a red ' light, URGENT FAILURE, on the front panel of the Power Cabinet.
3.
Energizes a plant annunciator window, ROD CONTROL URGENT l
FAILURE.
j 4.
Stops all automatic rody" "NON-URGENr" alarms are caused.by failure of either of the two paralleled +24 VDC power' supplies.
The "NON-URGENT" alarm energizes an amber light P.S. FAILURE on the front panel of the Power Cabinet and energizes a plant annunciator window RCD CONTROL NON-URGENT FAILURE.
A green light GROUP CYCLING on the Power Cabinet front panel is energized while the cabinet is sequencing a group of mechanisms through one step of 780 milliseconds.
3 Failure Detectors Four typhs of failure detectors have been provided to monitor the system and operate the urgent alarm in the event of power or i
circuit failure. These detectors are:
~
l 1.
Regulation Failure Detector - Senses that the output current j
to the coils does not match the current command signal within a preset time or that the full current command signal is on j
longer than a preset time.
This detector serves a dual purpose.
It prevents dropping of rods should the system be unable to regulate the required coil current and protects the coil from overheating should a logic cabinet slave cycler failure call for continuous full current.
A separate detector is provided for each of the five thyristor bridges.
i i
/
^
. ISSUE 2 REVISION 4
)
B.V.P.S. - 0.M.
1.1.1 INSTRUMENTATION AND COEROLl3 (continued)
/
new command, the control circuits set the FPINHX signal to ZERO to release a fast pulse control circuit in the master cycler. As a consequence, high speed pulses derived from the oscillator are used to count out the first GO pulse. The first GO pulse then disables,the fast pulse circuit so that subsequeat GO pulses are produced in the normal way by counting the slow speed pulser clocks.
Data Lozzina-As the control or Shutdown Banks (A and B) move in or out, the data logging circuits provide step pulses that indicate when a bank is starting or finishing a step (depending on direction) and when each group of the bank is finishing a step. The step pulses result from the interaction of the slave cycler sovable decoder strobes (MBIAC-MB3BD) and the bank selection signals (CID-C4D, SA-SB).
The group step pulses actuate the step counters on the Control Board (via output relay drivers). The bank step pulses are sent to the Computer and the Rod Position Indiestion System.
When an urgent alarm occurs in a Power Cabinet, the group step pulses are inhibited.
A1 arm The al'ars circuits process failure detection received from circuits within the Logic Cabinet and also process failure detections received from the Power Cabinets. Two types of alarms are recognized: non-urgent and urgent.
If a non-urgent alarm condition is detected in the Logic Cabinet, the NON-URGENT ALARM lamp on the Logic Cabine,t lights and the IQ:UAX signal goes to ZERO, thereby signaling the Annunciator Panel (via a relay); if a non-urgent alarm condition is detected in a Power Cabinet, KNUAX i
goes to ZERO but the lamp does not light.
If an urgent ala u
\\ condition is detected in the Logic Cabinet, the URGENT ALARM lamp on the Logic Cabinet lights, and the KUGAX signal goes to ZERO, l
thereby signaling the Annunciator Panel (via a ' relay);
if an i
i urgent alars condition is detected in a Power Cabinet, KUGAX goes I
to ZERO but the lamp does not light.
Within the Logic Cabinet, only redundant power supply failures are classified as non-urgent. Slave cycler failures or oscillator failures (Al or A2) are classified as urgent.
In addition, if the printed card interlock system is broken by removal of a card, an urgent alarm is generated.
The detection of an urgent alarm causes the general alarm signal (AGEN) to go to ONE. As a result, the INKBTX signal goes to ZERO I
and overlapped rod motion, manual or automatic, is prohibited until the condition is investigated and the circuits reset.
M utput
. ISSUE 2 REVISION 4 i
r
%. r-
) >
w TARLE 2,2-1 (Continued) lg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 1
l fj
\\
NOTATION I'
F (T-T')*K(P-P')-f(AI)]
f NOTE 1:
Overtemperatura AT i AT,[K -K2 7
3 y
l c}
wliare:
AT, Indicated AT at RATED TilEltHAL POWER
=
Average tamparatura. *F T
=
T',= 676.3*F (indicated T at RATED TiltilHAL POWER) 4 w
ave Pressurizar Frassure, psig P
=
l I
2235 psig (Indicated RCS nominal operating pressura)
P'
=
l 7
Tlia function generated by the lead-lag c5ntro11er for.T,,, dynamic
=
compensation.
i Time constants uttilzad in the lead-lag controllar for T,,,T
M sacs, Tg & T2
i T2 = 4 sacs.
i Laplace transform operator.
S
=
4 9
bbO O.o..,
4+ 4 O
1 A * *
.b 1
v l
I TABLE 2.2-1 (Continued)
L l
h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPolNTS j
- 5
~
NOTATION (Continued) 1 N
$l-Operation with 3 Loops Operation with 2 Loops Operation with 2 Loops Q
(noloopsisolated)
(1 loop Isolated)
I e
K
= 1.18 K
= 0.99 K
= 1.1 j
g y
y y
f U
K
= 0.01655 K.
= 0.01655 K
= 0.01655 2
2 2
i K
= 0.000001 K
= 0. M 1 K
= 0. M 1 3
3 3
and f (AI) is a function of the indicated difference between top and bottom detectors 1
y l
a of the power-range nuclear lon chambers; with gains to be selected based on measured instrument response during plant startup tests such that:.
i between -23 percent and + 11. percent, f (AI) = 0 l
(1) forq$q-qIndq are percent RATED TilERMAL POWER in the top and bottom (wher halvesofthecNrerespectively,andgg
- gb is total TilERMAL POWER in
. percent of RATED TilERMAL POWER).
~ _ _ _... _.. _. _ _
for each perce'nt that the magnitude of (qIy reb)ced by 1.54 percent of
-q exceeds -23 percent, 1
(11) l the AT trip setpoint shall be automatical Its value at RATED TilERHAL POWER.
l g
the AI trip setpoint shall be automaticalky rebu)ced by 1.91 percent of l
- a j
for each percent that the magnitude of (q
-q exce.eds + 11 percent, (111) l 3
l Its value at RATED TilERHAL POWER.
2 I
?
u:
J.o 9. -
1 n
B.V.P.S. - 0.M.
1.24.1 INSTRUMENTATION AND CCNTROLS (continued) to the motors are lifted or the motor operator removed. The valves will be operated by hand only and are normally locked open. The valve position indicating lights are located on Benchboard - Section A at the non-functioning control switches.
~
Feedvater Isolation l
See Figure 24-10 for feedwater isolation logic (lower right hand corner). Main feedwater regulating valves and bypass flow control valves receive diaphram operating air from their valve positioners through two, in-line, three way solenoid operated vent valves.
Normally, the air valves are aligned to pass operating air frcm the valve positioners to the actuator. On loss of solenoid power the air valves reposition to vent air from the actuators causing rapid valve closures (less than 5 seconds).
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~-
Steam Generator Water I.evel Control (SGWI4) IFCV-lW-478, 488, 4981 Each stieam generator is equipped eith a three-element feedwater control system (feed flow, steam flow and water level) which maintains a programmed water level in the secondary side of the steam generator during normal plant operation. The controller continuously compares measured feedwater flow with steam flow and steam generator water level signal with a water level setpoint to regulate the main feedwater I.
control valve opening.
Manual control of the main feedwater control' valve is provided by the system using the benchboard mounted auto-I l
manual station. A bypass line with a bypass feedwater control valve is provided around each main feedwater control valve. The bypass valve is used to manually or automatically control feedwater flow at loads below approximately 15 percent, which are outside the load range where automatic control of the main feedwater control valve gives optimum response. For the following discussion, see Figure 24-10 which shows the control system for steam generator No. 1.
Steam generator water level control (SGWLC) system setpoints are listed in Section 2 of this chapter.
Each steam generator has three (3) narrow range level instrument trains. Only the instrument train of protection channel III sends a level signal to SGWLC and Low Power (LP) SGWLC. Each feedwater header and steam header is equipped with two flow detectors.
The flow instruments selected by the operator to record on the vertical board mounted flow recorder are also the inputs to SGWLC.
Each steam flow instrument receives pressure compensation from its own sssociat,d steam e
( ISSUE 2 REVISION 14
-_g-
.-~%,,m..-.
m
STAGE IDESSURE S TEAM NR I I VE L lil COMPUTEll 01447 CH446 FLOW i)SEL SWITCH fl 474 t T 476 SIG COVD Tl*RBINE FIRST FR 478 C5 4- -~*Fot%ER MISMATOI 1, TEST SWITCH STACE PRESSLWif
~~~ 1 ~EST SWITCH L
yg jT 44*yo IP~
LO4M i
If
- Tp F
47
[V
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SKQM
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.. [V LC 476 f m2/3 HI LV
$] TEAM PRESS SIG Sth COMPARATOR 1 Y
S k
Il2 FILTEINFROGRAM LEVEL V
Tp g,u COMP IVS 476~l TRIP SRK f5RNA-F AIAA W
L L
60Aue SIGCcNO
- y TER LOGIC VV
' h478A}g'& N C
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s 4768 3{d
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.m I OGICL I f L Q Cd C I & H M Y 2/3 LOW-LOW FC 478 ERROR u
A yV I,t LEVEL A/M I
y e
STAllO4 REED FLOW irr476}
Trnl w ANY SG --
2/3 HI*HI LE fL TEST SWITCH LOOPS RX To TRIP l
Tp- --
FQ 476 3
FLOW ERRORgyppygg_ 473
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g
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P
-g (c47ggmyQ j
7_
g~
SIG COVO"CO6FUTER Fs ~ Ff LV TOTAL r
i LR
'/y gyy SEL SWITOi cRROR TRIP BYTnSS CHANNEL 47i_If1 }
kr 50GNAL VALVES ^
n f#
~
fu s?l5A
$ b FEE 0 FLOW 1>FR 478 4 "~*
EAS.Tl@fN?f m
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-s
-15"-l VALVE FW futPS 4MOVS l
fM 476
~
SEL
~
497
-sSIG Cost' y 7,g
- P WSITM LOG'C
_ _,fsNT cow Kfso a/S 4
~
Ett X-47,o AIR SUPPLY Air Ia open 9
$=1 q[fgg Spring so close F
%,e 7e De-energize.
I
\\
8" STEAM GENERATOR WATER LEVEL CON THOL FIGUlE 24-10 E L'W'"' Y l
~
FIGURE NO. 24-10 l
SG WATER LEVEL CONTROLLER
- 4. 02. la Nuclear Gsoap - Station ad=4=fatrative Procedures.
Chapter i
and p...i m answers in accordamse to NGD-22,
" Notification and Nous Release Policy."
d.
The following is a list which is to be used as 4 gnide in daaiding if an incident is reportable.
1)
Losses believed to be near or above dednatible.
2)
Insidents where fixed fire protection systems have operated.
3) f ar id-ta where prompt assistance could help prevent further loss or expense or where assistese is otherwise desirable.
4)
Tamid==ts where
'a--adia-ise or malicious mischief is suspected.
5)
Imargency
'T * --
ta to fire protectio l
equipsest such as underground fire sai:
breaks.
4)
A serious failure occurs involving key electrical,
==ah ma i ami or pressure systems
~
components such as the main turbine generator.
(
or Station transfotmars.
Y.
m ar e a to Coerstinz Procedures Procedures are intended to provide instructions to cover a reasonable number of foreseeable situations which could be conceived to happen at BVPS Unit #1.
One of the basic reasons for extensive training and qualificatioa requirements which are es.ablished for operating personnel is to ensure that these personnel have the technical knowledge ane.
Judgement capabilities required to understand and utilize the proper procedures for the various plant operations and cc take the necessary actions when procedures are not provided.
1.
In the event a procedure cannot be followed as writter.
or a procedure is not available, the activity should not be coedneted um,less required by an emergency or casualt--
situation until the procedure has been revised or prepared, approved and issued, or a decision has bee:
made by the Station Superintendent that a new or revised procedure is not required.
2.
Operating personnel may & reasonable action that departs from a license condition or a
Technics *.
Specification la an emergency 4.s per 10 CFR 50.St. (x' when this action is immediately needed to protect the public health and. safety and no action consistent with Page 40 of $2 Revision ;*
Chapter 4 -
Muslaar Group - Statica Ad=faintrative Procedures license conditions and Technical Specifications that can l
provide adequase or equivalent protec. ion is immediately apparums. The gafdelines of SAP-4 VI, Procedure AE, shoald be followed when taking the action. This action a licensed Senior l
shall be approved, as a ainimum, by Operssor prior to taking the action.
3*.
The immediate actions in the emergency procedures Off and performed Chapter 534, most be committed to memory in the spesified sequence. In the event of an emergency or casusity not covered by an approved procedure, operattag pers,onnel have the responsibility and sathority to taka whatever action considered necessary to prevent injury to personnel or damage to plant equipment and to place the plant in a safe condition.
4.
Initial conditions shall be followed at all times.
Notaal System Arrangecent as stated in-Inicial Conditions is seant as determined by present Control Boon Logs, Status Boards and/or Station-controlled flow diagrams.
Any deviations to this mast be evaluated by Operst$cus supervision.
La exception of strict coupliance of initial conditions is when plant equipment is -out of service and procedure portions are used for gnidamaa to d-t (e.g.,
valve
- strokes or other astians) os re& as a piece of equipment to service.
1 i
In those cases where procedure intent changes are act 3
k involved and related systems are not affeued, a one-time-only OtlCN say be suhaitted in accordance with esisting ad=4aintrative procedures to cover those cases.
i An example is main. feedwater regulation valve strokes which are required in Mode 5 or 6, but can be performed in Mode 1 while the valve is out of service.
5.
Vhes estensive operations, infrequent operations or any operations requiring documentation are to be performed, j
the operating p'rocedure gg, be pr,esent and followed.
l In Chapters 50, 51, and $1, those procedural steps not l
plus (+) sign may be omitted or performed marked by a out of sequence at the discretion of the Nuclear Shift Supervisor, provided a Nuclear. Operations supervisor initials each omitted step affected with a brief explanation of justification for the change in the
" Remarks" section. The steps marksd by a
(+)
sign cannet be omitted but say be started out of. sequence.
6.
Routine procedural actions that are frequently repeated do not require the procedure to be pressat once the operator has eastered the coeration, but ue shall review the procedure and supervision will sonitor actions assinat the procedures to assure that improper habits are not being formed.
Examples are boration and
(,
i Page 41 of 52 Revision 20 i
~
l e
9
4 o -f. c 3VPS - E0P
- .32A.1 1
(
NUMBER TITLE l
FR-H.1 Response To Loss of Secondary Heat Sink i
4STzr H AcrI0s/zxrscTzo azStowSt ;
- azSeowSt nor OsTArazo ;
f y
- .**************s,
,*]
CAITTION l
- l
- IF RVST level decreases to less than 20 IT., the SI system should be aligned for cold les recirculation using ES-1.3, " Transfer To Cold Leg Recirculation," Seop l'.
- If SGs are classified as hot / dry SGs, (< 10% wide range level and pri.cary side temperature > 550F), feed flow should only be established to one SG.
When primary temperature is reduced below 550F, then feed flow can be
__ y initiated to the remaining intact SGs.
e*****************************************.,
i NOTE i
Steps 19 through 21 are a continuous loop until feed to SGs is established.
r
- 19. Continue attesets To Establish Secondarar Heat Sink In At Least One SG
- AW flow
- Main W flow k
- Condensate flow
- 0ther low ' pressure flow
(
f PAGE 13 CF 19
- S St.7.
- REV!S:0N 0
4.0 f <-
B.V.P.S. - 0.5.
1.30.2 PRECatTTIONS AND LIMITATIONS (continued) c 7.
Criticality sust be anticipated any time the control rods are being withdrawn or when boron dilution operations are in progress.
8.
If criticality is to occur at conditions (i.e., temperature, xenon, and boron concentrations) different from those for which previous criticality data is availabla and the differences are such as to cause an increase of 0.5% Delta K/K (500 pen) or more in core reactivity (as determined by procedure F,
" Estimated Critical Rod Position Calculation" of Chapter 50),
the approach to criticality sust be guided by plotting an inverse count rate versus control rod position.
9.
Avoid any operation which' produces changes in reactor coolant temperature of the order of i 10 F or in boron concentration of the order of
-10 ppe while the reactor is critical below the power range or when approaching criticality.
10 Do not exceed a startup rate (SUR) of one decade per minute (1 DPM) unless authorized for special casts.
11.
Loop stop valves shall not be closed in more than one loop unless the reactor coolant system is connected to the residual heat removal system and the residual heat removal system is operable.
4 to the reactor coolant system (RCS) must be at the 12.
The sakeup water same baron concentration or greater as determined by sampling, except when diluting.
13. The
=mwi-um reactor coolant system (RCS) heat-up of 100F in any one l
hour period sust not be exceeded. The administrative maximum primary system heacup rate is 60F per hour.
The maximum heatup of the pressurizar saat not exceed 100F in any one hour and a maximum spray water temperature differential of 320F sust not be exceeded.
The administrative
w4
pressurizar heatup is 100F per hour.
(T.S.
3.4.9.1 and 3.4.9.2) 14.
The temperatures of both the primary and secondary coolants in the steam generators shall be greater than 70F when the pressure of either coolant.in the steam generator is greater than 200 psig (T.S.
3.7.2.1).
15.
Initially, steam must be withdrawn from the steam generators slowly and feedwater added to the steam generators slowly to minimize transients in the primary system and thermal stress cycle temperature for the feedwater nossles. During hot standby operation, do not feed the steam generators at greater than 200 spe each.
16.
Observe all applicable Reactor Coolant chemistry specifications contained in the Beaver Valley Chemistry Manual.
17.
The pressurizer auxiliary spray must be initiated slowly by gradually increasing the charging line flow to minimize thermal. shock to the
( ISSUI 2 REVISION 0
h.0).a B.V.P.S. - 0.M.
- 1. lea.1
SUMMARY
DESCRIPTION (continued) valves outside containment are operated from a second common control air header which also has a common actuation valve.
These valves, normslly open, close automatically only on e CIA signal or manually by switches located on the Benchboard - Section A.
The blowdown sample valves are also closed upon receipt of a start signal from the auxiliary feed pumps.
High temperature and pressure reactor coolant samples can be obtained from the hot legs, cold legs, pressurizer, and RH system.
The RCS hot legs have sample caps on the loop side of the loop isolation valves at each hot leg drain connection.
Loop 1B also has a sample tap on the reactor side of the loop isolation valve.
Each tap i
has a manually operated root isolation valve near the point where it :
leaves the RCS.
These four sample lines then pass through individual {
=**
c 1
?2 The 3 RCS hot legs have sample caps downstream of the motor operatedd drain valves for each loop. These sample line legs exit containment in a manner identical to the hot leg sample.
The RHS heat exchanger inlet and outlet samples pass through their individual, manually operated root and air-operated isolation and telection valves, and then leave the containment separately.
After passing through individual manual valves, they join the hot leg sample header just upstream of sample cooler (ISS-E-5].
The pressurizer liquid space sample is drawn from containment in 3 manner similar to the RMS sample.
At the sample panel it passes through a manual isolation valve before entering its own sample cooler (ISS-E-7].
A high temperature sample consisting of steam and non-condensaole I
gasses is taken from the pressurizer vapor space.
It passes through a I
manually operated root isolation valve and an air operated isolation valve and leaves the containment.
At the sample panel it passes through a manual isolation valve before entering its own sample cocler (IS3-E-6). A capillary tube consisting of 90 feet of 1/8 inch CD tubing wound in a 3-inch diameter coil is installed in the sample line downstream of cooler [lSS-E-6].
This sample line is also used for degassification of the pressurizer whe'n the reactor plant is being shutdown and opened for repair or refueling. At the sample panel high temperature samples are cooled by passing them through the tube side of sample coolers cooled by reactor plant component cooling water.
Five sample coolers and associated manual valver are located at the sample panel.
I ISSUE 2 l
REVISION 5 l
i l
l l
,y n.
5.04.b 1.43.1 y1*.i B.V.P.S. - 0.M.
m
,OT 1pu$v_
MAJOR COMPONENTS (continued)
Desizn Data (Re~ actor Coolant High-Range Monitor) m Off line liquid sampler e,
Vater Medium
. iM, fig Co60, Cs137 Limiting Isotopes
' h' k Sensitivity, Microcurie /cc Cc60 1 E-1 2.5 Max. Background, Mr/Er Medium Temperature, F 120 200 Max. Medium Press, PSIG 50-120 Ambient Temperature, F 7
0 Max. Ambient Press, PSIG Qg Design Data (Rea'ctor Coolant Low-Ratge Monitor)
In line liquip sampler Type i
Vater Medium Co60, Cs137 Limiting Isotope
[j Sensitivity, Microcurie /cc Cc60 1 E-04 2.5 Max. Background, Mr/Hr
. '..gqi Medium Temperature, F 120 200 Max. Medium Press, PSIG
[ g 'q*xt Ambient Temperature, F 50-120 1,#ff
'O
/Weh Max. Ambient Press, PSIG
~ ~ ~ -
,,\\b:/,5' I - Auxiliarv Feedwater Area Drain Tank Monitor iRM-1DA-1001 rg. g-
[
Auxiliary Feedwater Area Drain Tank is analyzed by a Skid mounted no continous sample from the liquid radiation monitor which obtains a
[
- %C-l In the event Tank's drain connection and, returns it to the Tank vent.
r j
to the yard oil separator is of a high alarm condition discharge b
automatically redirected to the tunnel sump.
Y.Mk Design Data gaeQ c
- Offline liquid sampler Type
- Water Medium
- CS-137 e
.ms i
Limiting Isotope Sensitivity Microcurie /cc Cs137
- 1 E-06
- 25 Max Background
- 120 Medium Temperature
- 15 0 Max Medium Pressure Vaste Gas Decay Teks Monitor (RM-1GV-1011
.%g'__
The waste gas decay tanks monitor can analyze a sample pumped frem any selected decay tank. The sample is returned to the decay tank free which it was withdrawn.
A valving arrangement permits individual The detector is inserted sampling of each tank with the same monitor.
in a well in an 'oiT-line sampler.
--A high-high alarm alerts cho operator to divert the waste gas feed to another tank. The monitoring ISSW 2 g'_
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SCALE. NO*4 INo 8700-AB-13*JC _
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1.7.4 0.
BLENDER AL'ITRNATE DILirrION OPERATION This procedure describes the method for reducing the reactor coolant f
boron concentration using the makeup system in the alternate dilute mode. ~ 'dils' mode initially provides a more rapid start of the boro $
- concentration reduction than the dilute mode by supplying sakeuo wald
!to boeh - M rf 9: " --f --
a"=a muety it is normally used for small concentration reductions during xenon transients following load reductions.
Conditions conditions exist as in dilution except a more rapid start of The same dilution is required to follow a xenon transient.
Instructions This procedure is the same as for "DILITTE" except that the control is set for " ALTERNATE DILITTE".
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MAJOR COMPONENTS (continued)
I Compressor Motor Type Squirrel Cage Volts / Phase / Frequency 460/3/60 Speed, RPM 1,800 15 Horsepower Quench Soray Pumps f105-P-1A & IB1 i
ne quench spray subsystem, shown in Figure 13-1, are made up of two Each of separate parallel subsystems, each are 100 percent capacity.
these subsystems draws water independently from the RWST and discharge through a separate spray ring inside containment.
Precaustions are taken to prevent clogging the spray nozzels inside the containment.
debris The quench spray pumps have suction strainers which. entraps any, that may clog the spray nozzles, also during test operation by opening the normally closed valves in the recirculation line to the RWST. This l
test operation will detect the pressence of particulate matter in the RWST that could clog the spray nozzles. The quench spray pumps become operable automatically upon receipt of a CIB signal. This CIB signal is generated from a high high containment pressure, manually, or by test.
Once the pump has started, from one of these sources, the only way to stop the pump is to depress the CIB reset pushbutton (1) on the Benchboard Section A or (2) in the solid state protection racks. This will reset the CIB relays. Once the relays are reset the control switch may be placed in the STOP position.
If the CIS signal is still oresent the CIB initiate pushbuttons must be depressed before the pumps will automatically restart.
In order to provide adequate NPSH for the recirculation spray (RS) pumps, cold QS water will be diverted to the RS pump suctions.
Approximately 15 0 spe will be diverted to the inside RS pumps, and approximately 300 spe will be divertsd to the outside RS pumps.
Orifices will be employed to provide the necessary flow split between each set of outside and inside RS pumps.
The quench spray pumps suction valves (HOV-1QS-100 A & Bj are normally If these valves are not in their fully open position, an alarm open.
sounds in the control roce. The valves, though normally in the open position, also receive a CIB signal to open to insure opening if in the closed position.
The quench spray discharge valves (MOV-1QS-101A, B] are normally shut but receive a CIB signal to open.
If one of the discharge valves are in the fully open position or a quench spray pump start signal is at least a minimum generated and the discharge flow does not attain low flow alarm will annunciate in the control room to preset flow, a alert the operator that a low flow condition exists during system l
operation or a discharge valve is open during system standby.
. ISSUE 2 REVISION 6
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~).o).o SYMPTOMATIC RESPONSE / UNEXPECTED CONDITIONS E-0 Series 1.
RCP TRIP CitrTERIA Trip all RCPs if BOTH conditions listed below occur:
Charging /HESI pump (s) - FLOW INDICATED.
a.
b.
RCS/ Highest SG D/P - LESS THAN 145 PSID (S10 PSID ADVERSE CNMT]
2.
SI ACTUATION CRITIRIA and GO TO E-0, " Reactor Trip Or Safety Injection," Step 1, if Eb2R Actuate SI condition listed below occurs:
RCS subcooling based on core exit TCs - LESS THAN SUBC00 LING LISTED CN ATTACHMENT 6 PRZR level - CANNOT BE MAINTAINED GREATER THAN 5% (50% ADVERSE CNMT]
3.
RED PATH
SUMMARY
SUBCRITICALITY - Nuclear power greater than 5%
a.
b.
CORF.C00LI$G
- Core exit TCs greater than 1200F
-OR-Core exit TCs greater than 700F M RVLIS f,211 range less than 40% with no RCPs :nnning, Narrow range level in all SGs less than 5% (45% ADVERSE CNMTj c.
HEAT SINK g total feed flow less than 350 GPM Cold les temperature decrease greater than 100F in last 60 d.
INIIGRITY minutes AND RCS cold leg temperature less than 290F.
CONTAINMENT - Containment pressure greater than 45 PSIG.
e.
4.- CHARGING Pt2fP MINIFLOW CRITERION Close charging pump eini-flow isolation valves after SI actuation prior to RCP trip setpoint.
Mini-flow isolation valves say be re-opened any time after RCS tepressurires -
above 127S PSIG, but must be open at 1925 PSIG (1600 PSIG ADVERSE CNMT!.
4 e
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NUMBER E-0 Reactor Trip Or Safety Injection
.i 4 STEP jd ACTION / EXPECTED RESPONSE l
' RESPONSE NOT OBTAINED j i
l Circled numbers show immediate action steps.
t
!1 h Verify Reactor Trip Manually trip reactor. E reactor will NOT trip, THEN
- Rod bottom lights - LIT GO TO FR-S.1, " Response To Nuclear Power Generation /AWS,"
-AND-Step 1.
- Rod position indicators -
AT ZERO
- Annunciator AS-C., "Turb.
Trip Init. Due To Reactor Trip" -
LIT
- thusron flux - DECREASING t
h Sound The Standby Alarm, Announce Unit #1 Reactor Trio And Evaluate If EPP Should Be Initiated d
h Verify Turbine Trio Manually or locally trip turbine.
a.
Throttle and governor valves -
E turbine will NOT trip, THEN CLOSED manually run back turbine.
IT turbine can NOT be run back,
(
b.
Reheat stops and interceptors -
EEN close main steamline trip and L
CLOSED bypass valves, and non-return valves.
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(
PAGE 3 Of 22
- ssLT 1 RE7:5::N :
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l REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITICN FOR OPERATION
~
3.4.8 The specifio activity of the primary coolant shall
~:e limited to:
1 1.0uci/ gram DOSE EQUIVALENT I-131, and a.
1 b.
1 100/E uci/ gram.
APPLICABILITY:
MODES 1, I, 3, 4 and 5 ACTION MODES 1, 2, and 3*
a.
With the specific activity of the primary coolant >1.0 uCi/ gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figur, 3.4-1, be in HOT STANDBY with Tavg
<500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the speci'fic activity of the pritaary coolant > 100/Y uCi/ gram, be in HOT STANDBY with Tavg < 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
t MODES 1, 2, 3, 4 and 5 a.
With the specific activity of the primary coolant > 1.0 uCi/ gram DOSE EQUIVALENT I-131 or > 100/T uCi/ gram, perform the sampling and analysis requirement of item 4a of Table 4.4-12 until the specific activity of the primary coolant is restored to within its limits.
l SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the primary coolant shall 1:e determined to be within the limits be performance of the sampling and analysis program of Table 4.4-12.
l 5
- With Tavg 7,$00*F g
BEAVER VALLEY - UNIT 1 3/4 4-18 (next page is 3/4 4-20) i
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r ~ < ~ n e w..:...+1.u s m = r.. - =.~.,
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TABLE 4.4-12
~
PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE Alin ANALYSIS PROGifAR m
TYPE OF HEASUREMENT HINIHilH H00ES IN IAllCil AND ANALYSIS FRE@ENCY SURVEILLANCE RE@lRED Q
1.
Gross Activity Da'tarminption 3 times par 7 days with a 1, 2, 3, 4 t-maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> e
betwaar. samplas.
c:
2 2.
Isotopic Analysis for DOSE E@ IVA-1 par 14 days 1.
4
~
LENT l-131 Concentration 3.
Radiochemical for E Determination 1 par 6 months 1,
I I
I I
4.
Isotopic Analysis for Iodine a) Once par 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, I,Z,3,4,6 including I-131. I-133, and I-135 whanaver the s cific activity exceeds 1.0 pCl/ gram DOSE
,g E@lVALENT I-131
[.
or 100/E pCl/ gram, and b) Caa sample betwaan 1, 2, 3 2 & 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> followjng a TilERHAL POWER change exceeding 15 percent of the RATED IllERiiAL POWER within a one hour period.
Untti the specific activity of the primary coolant system is restored within its limits.
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0 20 30
' 40 50 60 70 '
80 90-100 5
{
PERCENT OF RATED THERMAL POWER J
d
... l. ~ - FIGU RE 3.4-1 l.
DOSE EQUlVALENT l-131 Primary Coolant Specific Activity Limit Versu f
Percent of RATED THERMAL POWER with the Primary Coolant Specific i
Activity > 1.0gCi/ gram Dese Equivalent I-131 I
1 1
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BEAVER VALLEY - UNIT 1 3/4421 1
i 1
Nrrschn7enf 0-ATTACHMENT 4 NRC RESPONSE TO FACILITY COMMENTS 1.04d Accept comment 1.05a Accept comment 1.06b Accept comment 1.06c Consider during grading 1.08b Consider during grading 1.08d Accept comment 2.01a Accept comment - Reference material inaccuracy 2.02a Consider during grading 2.06b Accept comment 2.07a Accept comment - Reference material inaccuracy 2.08b Changed answer to CLOSE 2.09b Accept comment 3.03a Consider during grading 3.03b Consider during grading 3.04b Consider during grading 3.04d-Accept comment 3.05b Consider during grading 3.08c Consider during grading 3.09a Consider during grading 3.09b Accept comment 3.10 Accept comment 4.01a Accept comment 4.02b Question deleted 4.03b Accept comment 4.04c Consider during grading 4.07c Accept comment 4.08c Accept comment - Reference material inaccuracy 4.08e Accept comment 4.09a Accept comment 5.10a Accept comment 5.10b Accept comment - Math error 6.02a Question deleted 6.02b Consider during grading 6.03a Consider during grading 6.04c Consider during grading - Reference material inaccuracy 6.05 Question deleted 6.06a Accept comment 6.09d Accept comment 6.10c Accept comment 7.0lb Accept comment - Redistributed points 7.02a Accept comment 7.02c Accept comment 7.03a Accept comment - Four answers required 7.03c Accept comment 7.06a Question deleted
r-'
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7.09c
. Accept comment 7.10c Accept. :omment
- - 7.10d Accept comment 7.07-Consider during grading 8.03 Accept comment - Redistribute points 8.10b
- Accept comment r
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