ML20100G718
| ML20100G718 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/27/1985 |
| From: | Thompson H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20100G727 | List: |
| References | |
| NUDOCS 8504080398 | |
| Download: ML20100G718 (51) | |
Text
_ _ _ _ -
.a ners.
k, UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION L
ij WASHING TON, D. C. 20555
~s...../
IOWA ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 115 License No. DPR-49 i
1.
The Nuclear Regulatory Comission (the Comission) has found that:
4 A.
The application for amendment by Iowa Electric Light & Power Company, et al, dated August 17, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Actl, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authori7ed by this amendment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regula,tions; D.
The issuance of this anendment will not be inimical to the common I
defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. DPR-49 is hereby amended to read as follows:
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_ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ - (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 115, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license amendment is effective as of the date of issuance.
F0P,THE NUCLEAR REGULATORY COMMISSION cfrak o
- c. -
/Hugh L.
omps n Jr., Director US Division of Licensing Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of Issuance: March 27.1985 l
1
ATTACHMENT TO LICENSE AMENDMENT NO. 115 FACILITY OPERATING LICENSE NO. DPR-49 DOCKET NO. 50-331 Revise the Appendix A Technical Specifications by removing the current pages and inserting the revised pages listed below. The revised areas are -identified by vertical lines.
LIST OF AFFECTED PAGES vii 1.1-17 3.1-21 3.6-28 3.7-44 l
1.0-2 1.1-18 3.2-16 3.6-41 3.7-48a
-1.0-3 1.1-21 3.2-17 3.7-1 3.7-49 1
1.0-5 1.1-23 3.2-23 3.7-3 3.12-1 3.5-6 3.7-4 3.12-11 1.1-2 1.2-1 3.5-8 3.7-6 3.12-16*
1.1-3 1.2-4 3.5-9 3.7-24 3.12-17 1.1-14 3.1-3 3.5-19 3.7-31 3.12-18 1.1-15 3.1-4 3.5-22 3.7-36
.3.12-19 1.1-16 3.1-6 3.5-26 3.7-37 3.12-20
~
- These pages have been deleted.
e 6 *
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DAEC-1 TECHNICAL SPECIFICATIONS LIST OF FIGURES Figure Nunber Title 1.1-1 Power / Flow Map 1.1-2 Deleted 2.1-1 APRM Flow Blased Scram and Rod Blocks 2.1-2 Deleted 4.1-1 Instrument Test Interval Determination Curves 4.2-2 Probability of System Unavailability Vs. Test Interval 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements 1
3.4-2 Saturation Temperature of Sodium Pentaborate Solution
\\
i 3.6-1 DAEC Operating Limits 6.2-1
.DAEC Nuclear Plant Staffing l
3.12-1 Kf as a Function of Core Flow 3.12-2 Deleted 3.12-3 Deleted 3.12-4 Deleted 3.12-5 Deleted l
3.12-6 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P80RB301L) 3.12-7 Limiting Average Planar Linear Heat Generation Rate (Fuel Type P8DPB289)
{
3.12-8 Limiting Average Planar Linear Heat Generation Rate (Fuel Type BP/P80RB299)
{
3.12-9
-Limiting Average Planar Linear' Heat Generation Rate (Fuel Type P80RB284H) vii Amendment No. 115
DAEC-1 S.
OPERABLE-OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).
Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that. are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
6.
OPERATING Operating means that a system or component is performing its intended functions in its required manner.
)
. 7.
IMMEDIATE Immediate means that the' required action will be initiated as soon as practical considering the safe operation of the unit and the importance of the required action.
8.
REACTOR POWER OPERATION Reactor power _ eperation is any operation with the mode switch in the "Startuo" or "Run" position with the reactor critical and above 1% rated power.
1 1
9.
HOT STANDBY CONDITION l
Hot standby condition means operation with coolant temperature greater than 212*F, reactor vessel pressure less than 1055 psig, and the mode switch in the Startup/ Hot l
Standby position.
- 10. COLD CONDITION Reactor coolant temperature equal to or less than 212*F.
- 11. HOT SHUTDOWN The reactor is in the shutdown mode and the reactor coolant temperature greater than 212*F.
- 12. COLD SHUTDOWN The reactor is in the shutdown mode, the reactor coolant temperature equal to or less than 212*F, and the reactor vessel is vented to atmosphere.
1.0-2 Amendment No. 115
DAEC-1
- 13. MODE OF OPERATION A reactor mode switch selects the proper interlocks for the operational status of the unit. The following are the modes and interlocks provided-a.
Startup/ Hot. Standby Mode - In this mode the reactor protection scram trips, initiated by main steam line isolation valve closure, are bypassed.
The reactor protection system is energized with IRM neutron monitoring I
system trip, the APRM 15% high flux trip, and control rod withdrawal interlocks in service. The lower pressure MSIV closure 850 psig trip is l
also bypassed. - This is intended to imply the Startup/ Hot Standby position of the mode switch.
- b.. Run Mode - In this mode the reactor vessel pressure is at or above 850 l
psig and the reactor protection system is energized with APRM protection (excluding the 15% high flux trip) and RBM interlocks in service.
I c.
Shutdown Mode - Placing the mode switch to the shutdown position initiates i
a reactor scram and power to the control rod drives is removed. After a short time period (about 10 seconds), the scram signal is removed allowing a scram reset and restoring the normal valve lineup in the control rod
' drive hydraulic system; also, the main steam line isolation scram is bypassed.
d.
Refuel Mode - With the mode switch in the refuel position interlocks are established so that one control rod only may be withdrawn when the Source Range Monitor indicates at least 3 cps and the refueling crane is not over the reactor; also, the main steam line isolation scram is bypassed.
If l
the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.
l i
- 14. RATED POWER Rated power (100% power) refers to operation at a reactor power of 1658 Mwt.
I 1.0-3 Amendment No. 115
DAEC-1
'19.
ALTERATION OF THE REACTOR CORE (CORE ALTERATION)
The addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completi5n of the movement of a component to a safe conservative position.
- 20. REACTOR VESSEL DRESSURE Unless otherwise indicated, reactor vessel pressures listed in the Technical Specifications are those measured by the reactor vessel steam space detectors.
- 21. THERMAL PARAMETERS Minimum Critical Power Ratio (MCPR) - The value of critical power a.
ratio (CPR) for that fuel bundle having the lowest CPR.
britical Power Ratio (CPR) - The ratio of that fuel bundle power b.
which would produce boiling transition to the actual fuel bundle power.
Transition Boiling - Transition boiling means the boiling regime c.
between nucleate and film boiling. Transition boiling is the regime in which both nucleate and film boiling occur intermittently with neither type being completely stable.
d.
Deleted Linear Heat Generation Rate - the heat output per unit length of e.
fuel pin.
f.
Fraction #of Limiting Power Density (FLPD) - The fraction of limiting power density is the ratio of the linear heat generation rate (LHGR) existing at a given location to the desigr. LHGR for that bundle type.
g.
Maximum Fraction of Limiting Power Density (MFLPD) --The maximum fraction of limiting power density is the highest value existing,in the core of the fraction of limiting power density (FLPD).
h.
Fraction of Rated Power (FRP) -_The fraction of rated power is the ratio of core thermal power to rated thermal power of 1658 l
MWth.
- i. Total Peaking Factor (TPF) - The ratio of local LHGR for any specific location on a fuel rod divided by the core average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.
- j. Maximum Total Peaking Factor -(MTPF) - The largest TPF which exists
~ in the core for a given class of fuel for a given operating condition.
1.0-5 Amendment No. 115
DAEC-1 p
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING C.
Power Transient Where:
S = Setting in percent of rated power (1,658 MWt)
)
To ensure that the Safety Limits-established in Specification W = Recirculation loop flow in 1.1.A and 1.1.8 are not exceeded, percent of rated flow.
each required scram shall be -
Rated recirculation loop initiated by its primary source flow is that recirculation
' signal. A Safety Limit shall be loop flow which assumed to be exceeded when scram corresponds to 49x106
. is accomplished by a means other
~ lb/hr core flow.
than the Primary Source Signal.
For a WLPD greater than FRP, the D.
With irradiated fuel in the APRM scram setpoint shall be:
reactor vessel,.the water level shall not be less than 12 in.
FRP above the top of the normal S < (0.66 W + 54) active fuel zone. Top of the WLPD active fuel zone is defined to be 344.5 inches above vessel zero NOTE: These settings assume (see Bases 3.2).-
operation within the basic thermal design criteria. These criteria are LHGR < 13.4 KW/ft (8x8 array) and MCPR I values as indicated in Table 3.17-2 times K, dere f
K is defined by Figure 3.12-f 1.
Therefore, at full power, operation is not allowed with MFLPD greater than unity even if the scram setting is reduced.* If (
it is determined that either 'of these design criteria is being violated during operation, action must be taken immediately to
+
return to operation within these criteria.
2.
APRM High Flux Scram When in the REFUEL or STARTUP and HOT STANDBY MODE the APRM scram i
shall be-set at less than or e to 15 percent of rated power. qual
- With WLPD greater than FRP during power ascenston up to 90% of rated power rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% of MFLPD, provided that the adjusted APRM reading does not exceed 100% of rated power and a notice of adjustment is control panel. posted on the reactor 1.1-2 Amendment No 115 f
i
l DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.
APRM Rod Block when in Run Mode.
For operation with MFLPD 1ess than or equal to FRP the APRM Control Rod Block setpoint shall be as shown on Figure 2.1-1 and shall be:
S < (0.58 W + 50) l The definitions used above for the APRM scram trip apply.
For a MFLPD greater than FRP, the APRM Control Rod Block setpoint shall be:
S < (0.65 W + 54)
~
MFLP0 4.-
IRM - The IRM scram shall be set at less than or equal to 120/125 of full scale.
B.
Scram and
> 514.5 inches Isolation above vessel on reactor zero (+170" low water indicated level level)
C.
Scram - turbine
-' 10 percent stop valve
valve closure closure D.-
Turbine control valve fast closure shall occur within 30 milliseconds of the start of turbine control valve f ast closure.
- see footnote to 2.1.A.1 l
1.1-3 Amendment No.115
OAEC-1
2.1 BASES
LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY The abnormal operational transients applicable to operation of the Ouane Arnold Energy Center have been analyzed throughout the spectrum of planned operating conditions up to the thermal power condition of 102% of 1658 MWt in accordance with Regulatory Guide 1.49.
The analyses were based upon plant operation in accordance with the operating map given in Figure 1.1-1 of the Technical Specifications.
In addition,1658 MWt is the licensed maximum power level of the Duane Arnold Energy Center, and this represents the maximum steady state power which shall not knowingly be exceeded.
Conservatism is incorporated in the transient analysis-in estimating the controlling fart:rs, such as void reactivity coefficient, control rod scram worth, scram delay time.. peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect'on the applicable transient results as determined by the current analysis mode. Conservatisms incorporated into the transient analysis is documented in Reference 1.
1.1-14 Amendment No. 115
i DAEC-1 i
I This choice of using conservative values of controlling parameters and initiating transients at the rated power level produces more conservative resu'lts than would be obtained by using expected values of control parameters and analyzing at higher power levels.
For analyses of the toermal consequences of the transients the MCPRs stated in Section 3.12 as a limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.
Steady-state operation without forced ~ recirculation will not be permitted, except during special testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation pumps.
In summary:
1.
The abnormal operational transients have oeen analyzed to a pc.ser level of 102% of 1658 MWt.
l 11.
The licensed maximum power level is 1658 MWt.
iii.
Analyses of transients employ adequately conservative values of the controlling reactor parameters.
1.1-15 Amendrent No.115
DAEC-1 iv.
The analytical procedures now used result in a more-logical answer than the alternative method of assuming a higher starting power in conjunction with the expected values for the parameters.
4 Trip Settings The bases for individual trip settings are discussed in the following paragraphs.
A.
Neutron Flux Trips i
1.
The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, reads in percent of rated thermal power (1658 MWt).
l Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux.
During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore, during abnormal operational transients, the thermal power of the fuel will be' less than that indicated J
by the neutron flux at the scram setting. Analyses demonstrate that with a 120 percent scram trip setting, none of the abnormal operational transients analyzed violate the fuel Safety Limit and there is a substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides even additional margin. An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached.
The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneJvering 1.1-16 Amendment No. 115 i
p.
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DAEC-1 i
.during operation. Reducing this operating margin would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses.
- Thus, f
the APRM scram trip setting was selected because it provides l
adequate margin for the fuel cladding integrity Safety Limit yet allows-operating margin'that reduces the possibility of unnecessary scrams.
I '
'The scram trip setting must be adjusted to ensure that the LHGR i
j transient peak is not increased for any combination of MFLPD and h
reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the 1
maximum fraction of-limiting power density _is greater than the fraction of rated power. This adjustment may be' accomplished by increasing the APRM gain and.thus reducing the slope and intercept point o'f the flow-referenced APRM High Flux Scram curve by the i
reciprocal of the APRM gain change.
I Analyses of the limiting transients show that no scram adjustment is ' required to assure MCPR greater than or equal to safety limit -
i-when.he transient is ' initiated from MCPR > values as indicated in
'r 1
Table 3.12.2.
l 2.
APRM High Flux Scram (Refuel or Startup & Hot Standby Mode)
For operation in these modes the APRM scram setting of 15 percent 7
of rated power and the IRM High FluxLScram' provide adequate thermal margin between the setpoint and the safety. limit, 25 4
percent 'of rated. The margin is adequate to accommodate-anticipated maneuvers associated with power plant startup.
I-Effects of increasing pressure at zero or low void content are
~:
j minor, cold water from sources available during startup is not much colder than that already.in the system, temperature j
coefficients are small 'and control rod patterns are constrained to be uniform by operating procedures backed up by the rod 4
l' Amendment No. 115 i
i
DAEC-1 worth minimizer and the Rod Sequence Control System. Worths of individual rods are very low in a uniform rod pattern.
- Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is not more than 5 percent of rated power per minute, and the APRM system would be more than adequate to assure a scram before the power could exceed the safety limit.
The 15 percent APRM scram remains active until the mode switch is placed 1
in the RUN position. This switch occurs when reactor pressure is l
greater than 850 psig.
3.
APRM Rod Block (Run Mode)
Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent rod withdrawal beyond a given parar level at constant recirculation flow rate, and thus prevents a MCPR less than safety limit.
This rod block trip setting, which is automatically varied with recirculation loop flow r, ate, prevents excessive reactor power level increase resulting from control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the Safety Limit increases 1.1-18 Amendment No.115
DAEC-1 With a scram setting at 10 percent of valve closure, the resultant increase in surf ace heat flux is such that MCPR remains above safety limit even during the worst case transient that assumes the turbine bypass is closed. This scram is by-passed when turbine steam flow is below approximately 30 percent of rated, as measured by the turbine l
first stage pressure.
O.
Turbine Control Valve Fast Closure (Loss of Control Oil Pressure Scram)
The control valve fast closure scram is provided to limit the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection.
It prevents MCPR from becoming less than safety limit for this transient.
E., F. and J.
Main Steam Line Isolat hn on Low Pressure. Low Condenser Vacuum, and Main Steam Line Isolation Scram The low pressure isolation of the main steam lines at 850 psig has been l
provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel.
Advantage is taken of the scram feature that occurs when the main steam line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the STARTUP l
position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams. Thus, the combination of main steam line low pressure isolation and isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.
In addition, the isolation valve closure scram anticipates the pressure and flux transients 1.1-21 Amendment No. 115
4 4
4 i
i MATED POWSR
- 1M M RATED PLOW = # 6 120
/
/
y' APRM FLOW
/
BIAS SCRAM
/
/,
100 I
/
/
/
/
/
/
80
/
N MINAL EMECTED
/
FLOW CONTROL LINE
/
/
8 l
/
5 60 I/
g l/
~
_--y O
en eCORE TEPJ1AL g
POWER LIMIT- !
40 g
WilEN REACTOR l o
IS f.785 PSIG l A
OR CORE FLOW'I
,,:6101 OF RATEDj a
5 2 20 25%
/
9
/
b I
NATURAL CIRCULATION LINE d.
I O
l u
/
0 0
_20 40 60 80 100 120 CORE FLOW RATE ( %OhRATED) 1 i
t DUANE ARNOLD ENERGY CENTER IOWA ELECTRIC LIGHT & POWER COMPANY TECHNICAL SPECIFICATIONS APRM FLOW BIAS SCRAM RELATIONSHIP TO NOR.t'AL OPERATING CONDITIONS
}
FIGURE 1.1-1
+
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1.1-23 AmeTdment. No.115
DAEC-1 SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.2-REACTOR COOLANT SYSTEM 2.2 REACTOR COOLANT SYSTEM INTEGRITY INTEGRITY Applicability:
Applicability:
Applies to limits on reactor Applies to trip settings of the coolant system pressure, instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
Objective:
Objective:
To establish a limit below To define the level of the which the integrity of the process variables at which reactor coolant system is not automatic protective action is threatened due to an initiated to prevent the pressure overpressure condition.
safety limit from *31ng exceeded.
Specification:
Specification:
1.
The reactor vessel dome l'.
The limiting safety system pressure shall not exceed 1335 setting shall be as specified psig at any time when below:
irradiated fuel is present in the reactor vessel.
Protective Action / Limiting Safety System Setting A.
Scram on Reactor Vessel high pressure 1055 psig l
B.
Relief valve settings 1110 psig j;11 psi (1 valve) i 1
1120 psig f; 11 psi (1 valve) 1130 psig)+~11 psi 4
(2 valves 1140 psig)+ 11. psi (2 valves
+
1.2-1 Amendment No. 115
~- --,
y-
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c.
DAEC-1 design pressure (120% x 1150 = 1380 psig; 120% x 1325 = 1590 psig).
The analysis of the worst overpressure transient, a 3 second closure of all main steam isolation valves with a direct valve position scram failure (i.e., scram is assumed to occur on high neutron flux), shows that the peak vessel pressure experienced is much less than the code allowable overpressure limit of 1375 psig (Reference 1).
Mus, the pressure safety limit is well above the peak pressure that can result from reasonably-expected overpressure transients.
A safety limit is applied to the Residual Heat Removal System (RHRS) when it is operating in the shutdown cooling mode.
At this time it is included in the reactor coolant system.
1.2 References 1.
" General Electric Boiling Water Reactor Supplemental Reload Licensing Submittal fer Ouane Arnold Energey Center", 23A1739.*
- Refer to analyses for the current operating cycle.
1.2-4 Amendment No. 115
TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Modes in Which Number of Minimum No.
Function Must be Instrument of Operable Operable Channels Instrument Provided by Channels for Refuel Startup Run Design Action (1)
Trip System (1)
Trip Function Trip Level Setting (6) 1 Mode Switch in X
X X
1 Mode Switch A
Shutdown (4 sections) 1 Manual Scram
~X X
X 2 Instrument A
Channels 2
IRM High Flux i 120/125 of fuel X
X (5) 6 Instrument A
y Scale Channels 2
IRM Inoperative
,X X
(5) 6 Instrument A
Channels 2
APRM High Flux
(.66W+54) (FRP/MFLPD)
X 6 Instrument A or B (11) (12)
Channels 2
APRM Inoperative (10)
X X
X 6 Instrument A or B Channels 2
APRM Downscale
> 5 Indicated on Scale (9) 6 Instrument A or B Channels s
2 APRM High F. lux in
,g Startup
-< 15% Power X
X 6 Instrument A
Channels 5
2 High Reactor i 1055 psig X(8)
X X
4 Instrument A
l g
Pressure Channels m
TMt.E3.1-1(Continued)
REETm IHOTECTI(W SYSTEM (SCR#f) INSTRtKNTATifW RE()JIRDDIT Minimun No.
of (ber&1e instrunmt edes in Wiidi itsber of Diannels Function N st be Instrunent for Trip Opertie G mnels Refuel Startte llun Pmvided Systen (1)
Trip function Trip Level Setting (6)
By Design Atton (1) 2 lii$ Dryell Pressure
< 2.0 psig X(7)
X(8)
X 4 Instrunmt A
Diannels 2
Reactor low Water
> +170" Indicated X
X X
4 Instrunent A
Level revel (15)
Dimnels w
2 th#i Water Level in
< 60 Gallons X(2)
X X
4 Instrunmt A
aM Scran Discharge Volune 1,
mannels
?
~
2 min Stean Line
< 3 x Nonnal Rated X
X*
X 4 Instrunent A
flifiRadiation her Background *
&annels 4
Main Stean Line
< IOK Valve Closure X
X X(13) 8 Instrunent A or C isolation Valve Closure (3)(13) (3)(13)
Diannels 2
Turbine Control Valve Within 30 milliseconds X(4) 4 Instrunent A or D Fast Closure (Loss of the Start of Control of Contrul Oil Valve Fast Closure Oiannels
%m)
R 4
Turbine Stop Valve
<10% Valve Closure X(4) 8 Instrunent A or D 2
Closure
- s Oiannels 2
First Stage 8 pass below 165 psig X
X X
4 Instrunent A or D
\\
y Diannels h
- Alann setting <l.5 X lennal Rated fbur Background
DAEC-1 3.
A nain steam line isolation valve closure trip bypass is effective when the reactor mode switch is in the shutdown, refuel or startup positions.
4.
Bypassed when turbine first stage pressure is less than 165 psig (corresponding to 30% of rated core power). This value of first stage pressure assumes that the second stage reheaters are not in service below 30% of rated Core power.
5.
IRM's are bypassed when APRM's are on-scale and the raactor node switch is in the run position.
6.
When the reactor is subcritical and the reactor water temperature is less than 212*F, only the following trip functions need to be operable:
i a.
Mode switch in :hutdown b.
Manual scrar c.
High flux IRM d.
Scram discharge volune high level - may be bypassed in the refuel and shutdown nodes for the purpose of resetting the scran.
4 e.
APRM 15% flux 3.1-6 Amendment No.115
DAEC-1
'to the Refuel mode during reactor power operation does not diminish the protection provided by the reactor protection system.
Turbine stop valve; closure trip occurs at approximately 10% of valve closure. Below 165 psig turbine first stage pressure (corresponding to 30%
of rated core power), the scrari signal due to turbine stop valve closure is by-passeo because the flux and pressure scrams are adequate to protect the reactor below 30% of rated core power, i'
ij' Turbinc Contro1' valve f ast closure scram trip shall initiate within 30 1
l_
milliseconds of the start of control valve fast closure. The trip level setting is verified by measuring the time interval from energizing the fast i
i acting solenoid (from valve test switch) to pressure switch response; the measured result is compared to base line data taken during each refueling outage. Turbine control valve fast closure is sensed by measuring disc i
dump electro-hydraulic oil line pressure (Relay Emergency Trip Supply) which decreases rapidly upon generator load rejection. This scram is only l
effective when turbine first stage pressure is above 165 psig (corresponding to 30% of rated core power).
1 The requirement that the IRM's be inserted in the core' when the APRM's read 5 as indicated on the scale in the Startup and Refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation.
l l
3.1-21 Amendment No. 115 i
E
- =
TABLE 3.2-C INSTRUENTATION THAT INITIATES, CONTROL R00 BLOCKS l
Minimum No.
of Operable Instrument Number of Channels Per Instrument Channels Trip System (1)
Trip Function Trip Level Setting Provided by Design Action 2
APRM Upscale (Flow.Blased) f(0.66 W t 42)(W PD 2
APRM Upscale (Not in Run Mode) i 12 indicated on scale 6 Inst. Channels (1) 2 APRM Downscale
> 5 indicated on scale 6 Inst. Channels (1) 1 (7)
Rod Block Monitor (Flow Blased) 1(0.66 W + 39)( FRP )(2) 6 Inst. Channels (1) p 1 (7)
Rod Block Monitor Downscale
> 5 indicated on scale 2 Inst. Channels (1) 2 IRM Downscale (3)
> 5/125 full scale 6 Inst. Channels (1) 2 IRM Detector not in (8) 6 Inst. Channels (1)
Startup Position 2
IRM Upscale i 108/125 C Inst. 0$annels (1) 2 (5)
SRM Detector not in (4) 4 Inst. Channels (1)
Startup Position 2 (5)(6)
SRM Upscale i10 counts /sec.
4 Inst. Channels (1) 5 k
Water Level-High
-< 24 gallons 1 Inst. Channel (9) 2a E
t I
DAEC-1 NOTES FOR TABLE 3.2-C 1.
For the startup and run positions of the Reactor Mode Selector Switch, there shall be two operable or tripped trip systems for each function.
The SRM and IRM blocks need not be operable in "Run" mode, and the APRM
[except for APRM Upscale (Not in Run Mode)] and RBM rod blocks need not be operable in "Startup" mode.
If the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall b'e' tripped.
If the first column cannot be met for both trip systems, the systems shall be tripped.
2.
W is the recirculation loop flow in percent of rated. Trip level setting is in percent of rated power (1658 MWt). A ratio of FRP/MFLP0
< 1.0 is permitted at reduced power. See Subsection 2.1.A.1.
3.
IRM downscale is bypassed when it is on its lowest range.
4.
This function is bypassed when the count rate is > 100 cps.
3.2-17 Amendment No. 115
TABLE 3.2-G INSTRUMENTATION THAT INITIATES RECIRCULATION PUPF TRIP s
Minimum Number of Operable Instrument Channels per Trip Number of Instrument Channels Provided System (1)
Instrument Trip Level Setting By Design Action 1
(ATWS) Reactor High
< 1140 psig 4
(2) l Pressure I
(ATWS) Reactor Low-
> +119.5 in.
4 (2)
Low Water Tndicated level (5)
Level 1
-< 140 msec (4) 2 (3)
Response Time) w NOTES FOR TABLE 3.2-G b
1.
Whenever the reactor is in the RUN Mode, there shall be one; operable trip system for each parameter for operating recirculation pump.
If this cannot be met, the indicated action shall be taken.
2.
Reduce power and place the mode selector-switch in a mode other than the RUN Mode.
3.
Two EOC RPT systems exist, either of which will trip both recirculation pumps. The systems will be individually functionally tested monthly.
hours, the system will be declared inoperable.If the test period for one RPT system exceeds two consecutive is inoperable for more than 72 consecutive hoursIf both RPT systems are inoperable or if one RPT system reactor power shall be less than 85% within four, hours.an orderly power reduction shall be initiated and the-4.
This response time is from initiation of Turbine control valve fast closure or Turbine stop valve closure to actuation of the breaker secondary (auxiliary) contact.
5.
Zero referenced to top of active fuel.**
[
- Top of active fuel zone is defined to be 344.5" above vessel zero (see bases 3.2).
h
_m-
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT D.
HPCI Subsystem D.
HPCI Subsystem 1.
The HPCI Subsystem shall be 1.
HPCI Subsystem testing shall be operable whenever there is performed as follows:
irradiated fuel in the reactor vessel, reactor pressure is Item Frequency greater than 150 psigf.up from and prior to reactor star a.
Simulated Once/ operating a Cold Condition,0.2 and except as Automatic cycle specified in 3.5.
Actuation 3.5.0.3 below.
Test b.
Pump Once/ month Operability c.
Motor Operated Once/menth Valve Operability d.
At rated reactor Once/3 months pressure demonstrate ability to deliver rated flow at a discharge pressure greater than or equal to that pressure required to accomplish vessel injection if vessel pressure were as high as 1040 psig.
I e.
At reactor Once/ operating pressure of cycle 150 + 10 psig demolistrate ability to deliver rated flow at a discharge pressure greater taan or equal to that pressure required to accomplish vessel injection.
The HPCI pump shall deliver at least 3000 gpm for a system head corresponding to a reactor pressure of 1040 to 150 psig.
l
.5-6 Amendment No. 115 l
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Item Frequency than or equal to that pressure required to accomplish vessel injection if vessel pressure were as high as 1040 psig.
l e.
At reactor Once/ operating pressure of cycle 150 + 10 psig demoWstrate ability to deliver rated flow at a discharge
. pressure greater than or equal to that pressure required to accomplish vessel injection.
The RCIC pump shall deliver at least 400 gpm for a system head corresponding to 1040 to 150 psig.
I 2.
From and af ter the date that f.
Verify that the Once/ operating the RCICS is made or found to suction for the cycle be inoperable for any reason, RCIC system is continued reactor power automatically operation is permissible only trans0
- red 4
during the succeeding seven from tae days provided that during such condensate seven days the HPCIS is storage tank to operable.
the suppression pool on a 3.
If the requirements of-3.5.E condensate cannot be met an orderly.
storage tank water shutdown shali be f aitiated level-low signal, and the reactor pressure shall be reduced to 150 psig within 2.
When it is determined that the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
RCIC subsystem is inoperable, the HPCIS shall be demonstrated to be operable immediately and weekly thereaf ter.
3.5-8 Auendment No. 115
b DAEC-1 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT F.
Automatic Depressurization F.
Automatic Depressurization System (ADS)
System _ _( ADS) 1.
The Automatic 1.
During each " operating cycle Depressurization Subsystem the following tests shall be shall be operable whenever performed on the ADS:
there is. irradiated fuel in the reactor vessel and the A simulated automatic reactor pressure is greater actuetion test shall be than 100 psig and prior to a performed prior to startup startup from a Cold after each refueling outage.
Condition, except as specified-in 3.5.F.2 below.
2.
From and after the date that 2.
When it is determined that one valve in the automatic one valve of the ADS is depressurization subsystem is inoperable, the ADS subsystem made or found to be actuation logic for the other inoperable for any reason, ADS valves and the HPCI continued reactor operation subsystem shall be demonstrated is permissible only during to be operable imediately the succeeding thirty days and 'at least daily thereafter.
unless such valve is sooner cade operable, provided that during such thirty. days the HPCI subsystem is operable.
3.
If the requirements of 3.5.F cannot be met, an orderly shutdown shall be initiated and the reactor pressure shall be reduced to at least 100 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
G.
Minimum Low Pressure Cooling G.
Minimum Low Pressure Cooling and Diesel Generator and Diesel Generator Availability Availability 1.
During any period when one 1.
When it is determined that diesel generator is one diesel generator is inoperable, continued reactor
. inoperable, all low pressure operation is permissible core cooling and containment only during the succeeding cooling subsystems shall be seven. days unless such diesel demonstrated to be operable generator is sooner made imediately and daily operable, provided thereafter.
In 3.5-9 Amendment No. 115 e
DAEC-1 1
i does not result in rapid depressurization of the reactor vessel.
The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or
. Core Spray System operation maintains core cooling.
The capacity of the system is selected to provide this required core cooling.
The. HPCI pump is designed to pump 3000 gpm at reactor pressures between approximately 1135 and 150 psig. Two l
sources of water are available.
Initially, demineralized water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor.
When the HPCI System begins operation, the reactor depressurizes more rapidly than would occur if HPCI was not initiated due to the condensation of steam by the cold fluid pumped into the reactor vessel by the HPCI System. As the reactor vessel pressure continues to decrease, the HPCI flow momentarily reaches equilibrium with the flow through the break. Continued depressurization causes the break flow to decrease below the HPCI 4
3.5-19 Amendment No.115
DAEC-1
'Because the Automatic Depressurization System does not provide makeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the CSCS. Performance analysis of the Automatic Depressurization System is considered only with respect to its depressurizing effect in conjunction with LPCI and Core Spray and is based on 3 valves.
There are four valves in the ADS and each has a capacity of approximately 810,000 lb/hr at a l
set pressure of 1125 psig.
The allowable out-of-service time for one' ADS valve is determined as thirty days because of the redundancy and because the HPCIS is demonstrated to be operable during this period. Therefore, redundant protection for the core with a'small break in the nuclear system is still available.
The ADS test circuit permits continued surveillance on the operable relief valves to assure that they will be available if required.
3.5-22 Amendment No. 115-
DAEC-1
3.5 REFERENCES
1.
Jacobs, I.M., " Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards", General Electric Company, APED, April 1968 (APED 5736).
2.
General Electric Company, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NED0-20566, 1974, and letter MFN-255-77 from Darrell G. Eisenhut, NRC, to E.D. Fuller, GE, Documentation of the. Reanalysis Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-lead Plants, dated June 30, 1977.
i 3.
General Electric, Loss-of-Coolant Accident Analysis Report for Duane Arnold Energy Center (lead Plant), NE00-21082-03, June 1984.
l 4
s p
3.5-26
' Amendment V6. 115
DAEC-1 the direct scram (valve position scram) results in a peak vessel pressure less than the Code allowable overpressure limit of 1375 psig if a flux scram is assumed.
The relief valve setpoints given in Section 2.2.1.8 have been optimized to maximize the simmer margin, i.e., the difference between the normal operating pressure and the lowest relief valve setooint. The Reference 2 analysis shows that the six relief valves assure margin below the setting of the safety valves such that the safety valves would not be expected to open during any normal operating transient.* This analysis verifies that the peak system pressure during such an event is limited to greater than the 60 psi design margin to the lowest spring safety valve setpoint.
Experience in relief and safety valve operation shows that a testing of 50 percent of the valves per year is adequate to detect failures or deteriorations. The relief and safety valves are benchtested every second operating cycle to ensure that their setpoints are within the 21 percent tolerance. Additionally, once per operating cycle, each relief valve is tested manually with reactor pressure above 100 psig and with turbine bypass flow to the main condenser to demonstrate its ability to pass steam. By observation of the change in position of the turbine bypass valve, the relief valve operation is verified.
- A normal. operating transient is defined as an event whose probability of occurrence is greater than once per 40 years, e.g., Turbine Trip with Bypass, MSIV closure with. direct scram.
3.6-28 Amendment No. 115
DAEC-1 The records will be developed from engineering data available.
If actual installation data is not available, the service life will be assumed to commence with the initial criticality of the plant.
These records will provide statistical bases for future consideration of snubber service life.
The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.
3.6 and 4.6 References
.1) General Electric Company, Low-Low Set Relief Logic System and Lower MSIV
-Water Level Trip for the Duane Arnold Energy Center, NtDE-30021-P, danuary, 1983.
2)
" General Electric Boiling Water Reactor Increased Safety / Relief Valve Simmer Margin Analysis for.Duane Arnold Energy Center," NEDC-30606, May, 1984.
3.6-41 Amendment No. 115 l
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.7 PLANT CONTAINMENT SYSTEMS 4.7 PLANT, CONTAINMENT SYSTEMS Applicability:
Applicability:
Applies to the operating status Applies to the primary and of the primary and secondary secondary containment system containment systems.
integrity.
Objective:
Objective:
To assure the integrity of the To verify the integrity of the primary and secondary primary and secondary containment systems.
containments.
Specification:
Specification:
A.
At any time that the nuclear 1.a. The pressure suppression pool system is pressurized above water level and temperature shall atmospheric or work is being be checked once per day.
done which has the potential to drain the vessel, the
' ' ~ b. Whenever there is indication of suppression pool water volume relief valve operation or testing and temperature shall be which adds heat to the maintained with the following' temperature.shall be continually suppression pool, the pool 1imits.
monitored and also observed and a.
Maximum water volume - 61,500 logged every 5 minutes until the cubic feet heat addition is terminated.
b.
Minimum water volume - 58,900
- c. Whenever there is indication of cubic feet relief valve operation with the temperature of the suppression c.
Maximum water temperature pool reaching 200*F or more, an l
external visual examination of (1) During normal power the suppression chamber shall be 1
operation - 95F.
conducted before resuming power i
operation.
(2) During testing which adds heat to the suppression
- d. A visual inspection of the pool, the water temperature suppression chamber interior, shali not exceed 10*F above including water line regions, the =,'ormal power operation shall be made at each major limit specified in (1) above.- In connection with.
refueling outage, such. testing, the pool 2.
The primary containment integrity temperature must be reduced shall be demonstrated as follows:
to below the normal power operation limit specified in (1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
t 3.7-1 Amendment No. 115 J
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 2)
Closure of containment isolation valves for the Type A test shall be accomplished by normal mode of actuation and without any preliminary exercising or adjustments.
3)
The containment test pressure shall be allowed to stabilize for a period of about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of a leakage rate test.
4)
The reactor coolant pressure boundary shall be vented to the containment atmosphere prior to s
the test and remain open during the test.
5)
Test methods are to comply with ANSI N45.4-1972.
6Property "ANSI code" (as page type) with input value "ANSI N45.4-1972.</br></br>6" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.)
The accuracy of the Type A test shall be verified by a supplemental test. An acceptable method is described in Appendix C of ANSI N45.4-1972.
3Property "ANSI code" (as page type) with input value "ANSI N45.4-1972.</br></br>3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..7-3 knendment No.115 i
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 7)
Periodic Leakage Rate Tests l
Periodic leakage rate tests shall be performed at or above the peak pressure (Pa) of 43 psig.
8)
Acceptance Criteria Reduced pressure tests.
(Pt,
... reduced pressure) The leakage rate Ltm shall be less than 0.75 Lt.
Peak pressure test.
(Pp, peak pressure)
The leakage rate Lpm shall be less than 0.75 (La).
9)
Additicnal Requirements If any periodic Type A test fails to meet the applicable acceptance criteria the test schedule applicable to subsequent Type A tests will be reviewed and approved by the Commission.
If two consecutive periodic Type A tests f ail to meet the acceptance criteria of 4.7.A.2.(a)(9) a Type A test shall be performed at each plant shutdown for major 1
refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet j
the subject acceptance criteria after which time the retest schedule of 4.7.A.2.(d) may be resumed.
3.7-4 Amendment No. 115 i
t-
OAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT d.
Periodic Retest Schedule 1)
Type A Test After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. (These intervals may be extended up to eight months if necessary to coincide with refueling outages.)
The third test of each set shall be conducted when the plant is shut down for the 10-year plant in-service inspections.
The performance of Type A tests
.., shall be limited to periods when the plant facility is nonoperational and secured in the shutdown condition under administrative control and in accordance with the plant safety procedures.
2)
Type B Tests a)
A continuous leakage monitoring system is provided to measure changes in containment leakage during service. Accordingly, penetrations and seals of this type (except air locks) shall be leak tested at greater than or equal to 43 psig (Pa) every other reactor shutdown for major fuel reloading.
b)
The personnel airlock shall be pressurized to greater than or equal to 43 psig (Pa) and leak tested at an interval no longer than one operating cycle. The airlock will be l
monitored for leakage with the continuous leakage monitoring system during plant operation.
A report 3.7-6 Amendment No. 115
DAEC-1 NOTES TO TABLE 3.7-2 I
ITest volume is filled with demineralized water then pressurized to greater than o.' equal to 43 psig with air or nitrogen for test.
I For all other penetrctions (except Main Steam Lines) test volumes are pressurized to greater than or equal to 43 psig with air or nitrogen for test.
l 2MO-4441, M0-4442 will be remote manually closed.
3Subject isolation valves to be installed at earliest practicable date per FSAR P. 6.4-10.C, dated 9/73.
I 1
3.7-74 Amendment No.115 l
DAEC-1 The pressure suppression pool water provides the heat sink for the reactor primary system energy release following a postulated rupture of the system.
The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1040 psig.
Since all of the gases in the drywell are purged into the l
pressure suppression chamber air space during a loss-of-colant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum allowable pressure. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approximately 43 l
psig which is below the design pressure of 56 psig. The minimum volume of 58,900 ft' results in a submergence of approximately 3 feet. Based on Humboldt Bay, Bodega Bay, and Marviken test facility data as utilized in General Electric Company document number NEDE-21885-P and data presented in Nutech document, Iowa Electric document number 7884-M325-002, the following technical ' assessment results were arrived at:
1.
Condensation effectiveness of the suppression pool can be maintained for both short and long term phases of the Design Basis Accident (DBA), Intermediate Break Accident (IBA), and Small Break Accident (SBA) cases with three feet submergence.
L l
3.7-31 Amendment No. 115 l
l l
DAEC-1 The primary containment preoperational test pressures are based upon the calculated primary containment pressure response corresponding to the design basis loss-of-coolant accident. The peak drywell pressure would be about. 43 psig which would rapidly l
reduce to 27 psig within 30 seconds following the pipe break.
Following the pipe break, the suppression chamber pressure rises to about 25 psig within 30 seconds, equalizes with drywell pressure shortly thereaf ter and then rapidly decays with the drywell pressure decay, (Reference 1).*
The design pressure of the drywell and s'uppression chamber is 56 psig, (Reference 2). The design basis accident leakage rate is 2.0%/ day at a pressure of 43 psig. As pointed out above, the l
drywell and suppression chamber pressure following an accident would equalize fairly rapidly. Based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual corrponents separately.
The design basis loss-of-coolant accident was evaluated by the AEC staff incorporating the primary containment design basis accident leak rate of 2.0%/ day, (Ref. 3). The analysis showed that
- NOTE: The initial leak rate testing performed during plant startup was conducted at a pressure of 54 psig in accordance with the original FSAR analysis of peak containment pressure (Pa).
3.7-36 Amendment No. 115
DAEC-1 with this leak rate and a standby gas treatment system filter efficiency of 90% for halogens, 90% for particulate iodine, and assuming the fission product release fractions stated'in TID-14844, the maximum total whole body passing cloud dose is about 2 rem and the maximum thyroid dose is about 32 rem at the site boundary over an exposure duration of two hours. The resultant thyroid dose that would occur over the course of the accident is 98 rem at the boundary of the low population zone (LPZ).
- Thus, these do,ses are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident. These
~
doses are also based on the assumption.of no holdup in the secondary containment, resulting in a direct release of fission products from the primary containment through the filters and stack to the environs. Therefore, the specified primary containment leak rate is conservative and provides additional margin between expected offs,ite doses and 10 CFR 100 guidelines.
The design basis accident leak rate at the peak accident pressure of 43 psig (P ) is 2.0 weight percent per day l
p (L ).
To allow a margin for possible leakage deterioration a
during the interval between Type A tests allowable containment operational leak rate (Lto), is 0.75 L to.
In addition to these 4
3.7-37 Amendment No.115
DAEC-1 the accidents analyzed, as the FSAR analysis shows compliance with 10 CFR 100 guidelines with an assumed efficiency of 99% for the l
adsorber. Operation of the fans significantly different from the design flow envelope will change the removal efficiency of the HEPA filters and charcoal adsorbers.
Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 11 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability, pressure drop and air distribution should be determined at least once per operating cycle to show system performance capability.
The frequency of tests and sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as i
evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accordance with USAEC Report OP-1082.
Iodine removal efficiency tests shall follow RDT Standard M-16-1T.
(The design of the SGTS system allows the removal of charcoal samples from the bed directly through the use of a grain thief.) Each sample should be at least two inches in diameter and a length equal to the thickness of the bed.
If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according 3.7-44 Amendment No. 115
DAEC-1 Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below 200*F during any period of relief valve operation. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings, (see Bases Section 3.7.A.1).
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:
(1) use of all available means -to close the valve, (2) initiate suppression pool water cooling heat exchangers, (3) initiate reactor shutd'own, and (4) if other relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
4 4
4 Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountereo.
Particular attention should be focused on structural discontinuities in-the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
8a Amendment No. 115
DAEC-1 3.7.A & 4.7.A REFERENCES 1.
"Duane Arnold Energy Center Power Uprate", NEDC-30603-P, May,1984 and to letter L. Lucas to R.E. Lessly, " Power Uprata B0P Study Report," June 18, 1984 2.
ASME Boiler and Pressure Vessel Code, Nuclear Vessels,Section III, maximum allowable internal pressure is 62 psig.
3.
Staff Safety Evaluation of DACC, USAEC, Directorate of Licensing, January 23, 1973.
4.
10 CFR 50.54, Appendix J, Reactor Containment Testing Requirements, Federal Register, August 27, 1971.
i 5.
DAEC Short-Term Program Plant Unique Analysis, NUTECH Doc. No.
10W-01-065, August 1976.
6.
Supplement to DAEC Short-Term Program Plant Unique Analysis, NUTECH Doc. No. 10W-01-071, October 1976.
3.7-49 Amendment No. 115
DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.12 CORE THERMAL LIMITS 4.12 CORE THERMAL LIMITS Applicability Apolicability The Limited Conditions for The Surveillance Requirements Operation associated with the apply to the parameters which fuel rods apply to those monitor the fuel rod operating 2
parameters which monitor the conditions.
j fuel rod operating i
conditions.
Objective Objective The Objective of the Limiting The Objective of the Conditions for Operation is to Surveillance Requirements is to assure the performance of the specify the type and frequency fuel rods, of surveillance to be applied to the fuel rods.
3 Specifications Specifications A.
Maximum Average Planar Linear A.
Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)
Heat Generation Rate (MAPLHGR)
During reactor power The MAPLHGR for each type of operation, the actual MAPLHGR fuel as a function of average for each type of fuel as a planar exposure shall be function of average planar determined daily during reactor exposure shall not exceed the operation at > 25% rated limiting value shown in Figs, thermal power and following any l
3.12-6, -7, -8 and -9.
If at change in power level or at any time during reactor distribution that would cause power operation it is operation with a limiting determined by normal control rod pattern as surveillance that the limiting described in the bases for value for MAPLHGR (LAPLHGR) is Specification 3.3.2.
During being exceeded, action shall operation with a limiting then be initiated within 15 control rod pattern, the minutes to restore operation MAPLHGR shall be determined at to within the prescribed least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hours.
limits.
If the MAPLHGR (LAPLHGR) is not returned to within the prescribed limits l
within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, reduce reactor power to < 25% of Rated Thermal Power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Surveillance and corresponding action shall j
continue until the prescribed limits are again being met.
3.12-1 Amendment No. 115
i OAEC-1
)
3.12 REFERENCES 1.
Duane Arnold Energy Center loss-of-Coolant Accident Analysis Report, NED0-21082-03, June 1984 l
2.
" General Electric Standard Application for Reactor Fuel,"
l NEDE-24011-P-A**.
3.
" Fuel Densification Effects on General Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDM-19735, August 1973.
4.
Supplement I to Technical Reports on Densifications of General Electric Reactor Fuels, December 14,1973 (AEC Regulatory Staff).
5.
Communication:
V. A. Moor.e to I.S. Mitchell, " Modified GE Model for Fuel Densification," Docket 50-321, March 27, 1974.
6.
R.B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NED0-10802).
7.
General Electric Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10CFR50, Appendix K, NEDE-20566, August 1974 8.
Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, NED0-24087, 77 NED 359, C1 ass 1, December 1977.
9.
Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 2: Revised Fuel Loading Accident Analysis, NED0-24087-2.
- 10. Boiling Water Reactor Reload-3 Licensing Amendment for Duane Arnold Energy Center, Supplement 5: Revised Operating Limits for Loss of Feedwater Heating, NED0-24987-5.
- Approved revision number at time reload fuel analyses are performed.
3.12-11 Amendment No. 115
DAEC-1 DELETED-i 3.12-16 Amendment No. 115 i
D1 13 a
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\\
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11.5 17,3 "5
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0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (CWd/t)*
1/ When core flow is equal to or less than 70% of rated not exceed 95% of the limiting values shown.
, the HAI"JICR shall
TICENICAL SPICIFICATIONS LIMITING AVERAGE PLANAR LINEAR HIAT GENERATION RATI AS A IUNCTION OF PLAN AVIRAGE EXPOSURE FUIL TTPE: BP/ P8DRB301L TIGURI 3.12-6 3.12-17 Amendment No. 115
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5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Plana: Average Exposure (GWd/t)
- 1/ When core flow is equal to or less than 70! of rated, the MAPLIGE shall not exceed 95! of the 11=1 ing values shown.
- 1 GWd/t = 1000 Mud /t DUANI APJICLD DTIiGY CIhiri 10'a'A MC""AIC LIGEI AND POWIR COMPANT TICENICAL 57ECIFIC. CIONS LIMITING AVERAGE PLANAR LUIT.AR IZAT GINIRATION RATE AS A FUNCTICN OF PLA'iAR AVIRAGE II?OSCRI l
iM.' ITPI: P8DP3289 FIGURI 3.12-7 3.1.2-18 Amendment No. 115
9
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9 9.0 c
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0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (GWd/t)*
1/ When core flow is equal to or less than 70I of rated, the MAPLHCR shall not exceed 95I of the limiting values shavn.
- 1 GWd/t = 1000 mwd /t DUANE ARNOLD ENERGY CEh'IER IOWA MC"AIC LIGEI AND PC'n'ER COMPANY TECENICAL SPECIFICATION 3 LIMITING AVERAGE FUL'ULR LINEAR REAT GENERAIION RL~I AS A FUNCTION OF PUL%Ut AVERAGE EIPOSURE FUEL TIPE: BP/PSDRB299 FIGURE 3.12-8 4
l 3.12-19 Amendment No. 115
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0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 Planar Average Exposure (CL'd / c )*
f
, the M HGR shall n t ex eed 95 o g
us
- 1 GWd/t = 1000 mwd /t DUANE ARNOLD INERCT CD II2 10k'A "'C11C LICE; AND Pok'ER C012 ANT TICENICAL SPECIFICAT!ON3 L.NITING A7ERAGE PLANAR LINEAR HEA:
CDiWION RATE AS A IITNCTION OT PLANAR AVERAGE EIPOSURE WEL TTPE: PSDR3284H
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FIGURE 3.12-9 3.12-20 Amendment No. 115