ML19323C772
| ML19323C772 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/24/1977 |
| From: | Desiree Davis Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323C736 | List: |
| References | |
| FOIA-80-162 NUDOCS 8005190164 | |
| Download: ML19323C772 (12) | |
Text
,
Q O
8 005 3 9 9 t& Y
/
UNITED STATES
- 4 NUCLEAR F EGULATORY COMMISSION 4'
f M.h i
WASHINGTON, D. C. 20055 i
g s.,*,,,<j 5y CONSUMERS POWER COMPANY E
F DOCKET NO. 50-155 y
~'
l l
BIG ROCK POINT PLANT AMENDMENT TO FACILITY OPERATING LICENSE fd J :
f i
Z Amendment No.14 p
License No. DPR-6 y
9 1.
Ti The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment by the Consumers Power y
b Company (the licensee) dated May 30,1975 (as supple-G
' mented by letter dated June 30, 1975), September 10, 1975 1
(as supplemented by letter dated May 25, 1977), May 26, 1976, April 21,1977 and May 18, 1977, comply with the h
standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; k
. B.
The facility will operate in conformity with the application, t
the provisions of the Act, and the rules and regulations of the Commission; i
C.
There is reasonable assurance (i) that the activities i
authorized by this amendment can be conducted without endan
.(ii) gering the health and safety of the public, and that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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.
O
^
N,
~
Consumers Power Company June 24, 1977 Our review of the remaining portions of your May 30, 1975 application will be the subject of a later action.
Copies of our Safety Evaluation and Notice of Issuance also are enclosed.
Sincerely, 9
1 t'
\\
Don L' avis, Acting Chief Operating Reactors Branch #2 Division of Operating Reactors
Enclosures:
1.
Amendment No.14 to DPR-6 2.
Safety Evaluation 3.
Notice cc w/ enclosures:
See next page e
t w--
gg.
J ' wo
- O y
ATTACHMEflT TO LICEf!SE AMEflDMEf!T fl0. la FACILITY OPERATIllG LICEriSE NO. DPR-6 DOCKET NO 50-155 Replace the following pages of the Technical Specifications contained in Appendix A of the above-indicated license with the attached pages bearing the same numbers, except as otherwise indicated. The changed areas on the revised pages are reflected by a marginal line.
Remove Insert 4-3 4-3 4-3a (new) 4-7 4-7 4-9 4-9 6-7 6-7 6-10 6-10 6-13 6-13 6
/
U u
,. :.1.1. (Contd)
?
(i) Operating Requiremnts 1.
The average rate of vessel temperature change during normal heatup or cooldown should not exceed 100 F/h when averaged over a one-hour period.
2.
Control rod withdrawal during power operation shall be such that the average rate-of-change of reactor power is less than 50 MW per minute when power is less than 120 MW '
t less than 20 MW per minute when power is between 120 MWt and 200 MW, an 10 MW per minute when power is between 200 MW anIf 240 MW
- t t
t 3.
Reactor vessel pressure shall be limited in accordance with Figure 4.1.
4.
The reactor shall not be made critical, with the exception of physics testing..at temperatures below the criticality limit shown on Figure 4.1.
I 4.1.2 Primary Coolant Recirculation System The primary coolant recirculation s'ystem shall consist of the reactor vessel, the steam drum, the reactor recirculation pumps, the interconnecting piping and valves, and the safety relief valves.
(a) Design Features Shall Be as Follows:
Number of Recirculation Loops 2
Number of Recirculation Pumps per Loop 1
Approximate Internal. Volume of 3830 System Excluding Reactor Core and Internals to Isolation Valves, Cubic Feet Approximate Volume of Coolant in 2689 System During 157 Mwt Operation,
-Cubic Feet Steam Drum:
Length, Overall, Feet 40 Inside Diameter, Inches 78 Wall Thickness, Excluding Cladding, 4-3/8 Inches Cladding Thickness, Minimum, Inches 5/32 Design Pressure, Psia 1700 Design Temperature. *F 650 Amendment No. la 4-3
O r!r;URE 4-1
'ig Roex Fo;nt - hen:::r Vescel Criticality, Cooldown, Heatup and Hydrotest Linitations Hydrotest Limit 1,500 Psi Operating Limit 4
1,300
) ;
1,000 llg ___.
n.
4 Criticality Limit 3
800 I
5
/
.5 500 Psi Curve applicable for heatup rates u to 100 F/h Tor 'the service period up to 2.8x10 9 nyt (approximately 5 years from 14/1/75).
"" ' -I ndt RT
~
ndt 50 100 150 200 3/26/T5 l
Indicated Reactor Water Temperature *F 4-3a Amendment No. 14
'.1.2 (Contd.)
~
4
/
(Micromho/cm)
Conductivity 5
Maximum 10 Maximum transient
- pH (Lower and Up;:er limits) 4.0 and 10.0 Chloride Ion (Ppm)
'l.0 35 l
Equilibrium Halogen Radioactivity (uc/ml) 100 Boron (Ppm)
(c) Leakaae Limits _
If the primary coolant system leakage exceeds 1 gpm and the l.
source of leakage is not identified, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If leakage from the primary coolant system exceeds 10 gpm, 2.
the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The high energy line sections identified in Table 9-3b 3.
shall be maintained free of visually observable through-wall leaks.
(a)
If a leak is detected by the surveillance program of Table 9-3b, efforts to identify the source of the leak shall be started immediately.
(b)
If the source of leakage cannot be identified within eight hours of detection or if the leak is found to be from a break in the sections identified in Table 9-3b, the reactor shall be in a cold shutdown within'48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
4.1.3 Primary System Shielding Reactor sgielding is ordinary concrete with a density of approximately 150 lb/ft.
Thickness varies in plan and elevation to suit structural requirenents. The shielding thickness directly opposite the core shall be approximately 9 feet, 6 inches. The control rod drive room, which is directly beneath the reactor, has ordinary concrete walls which shall be approximately 4 feet thick. A removable shield plug of a thickness 4 feet, 6-1/2 inches, consisting of 4 feet, 4 inches of concrete and 2-1/2 inches of lead, shall close the opening above the top of the reactor.
- Conductivity is expected to increase temporarily after startups from cold shutdown. The maximum transient value here stated is the maximum pennissible and applies only to the period subsequent to a cold shutdown between criticality and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 20% rated power.
a-7
O P
The steam drum, risers and downcomers are primarily shielded by ordinary concrete walls which shall vary in thickness from 4 feet, 9 inches near the bottom to 3 feet, 3 inches at the top.
A large section, 12 feet by 42 feet, of the steam drum enclosure wall serves as a blowout panel and shall contain high density, loose aggregate to a thickness of approximately 4 feet, 9 inches.
This provides the shielding equivalent to 4 feet, 9 inches of ordinary concrete.
The reactor shielding shall be cooled by a water-filled jacket at the inside face. _The cooling water system shall be designed to remove 60,000 btu per hour with the inlet water temperature at 680F.
Cooling water shall be supplied from the closed loop reactor cooling water system.
The jacket shall be a carbon steel, annular tank div.ded into eight segments, with water entering the bottom and leaving at the top.
Provisions are made to convert to air cooling.
r Amendment Mc. 14 4-8
)
y a
y..
4.2.2 Main Condenser (a) Design Features Shall Be as Follows:
Type Radial Flow Surface Condenser With Deaerating Hot Well Condenser Surface Area, 27,500 Square Feet Design Condensing Pressure, 1.5 Inches Hg Absolute Condensing Capacity, Pounds 460,000 per Hour 01.5 Inches Hg Absolute Condensing Capac",ty During Full 948,000 Load Rejection, Pounds per Hour Air Ejector Capicity 10 Cubic Feet per Minute of Air Plus 1.1 Pounds per Hour of Hydrogen Plus 8.3 Pounds per Hour of Oxygen (b) Operating Requirements (1) The following condenser pressure trips shall be operative during reactor power operations:
Annunciate, Inches He Absolute 5.0 1 0.5 i
Turbine Trip and Bypais Valve Closure, Inches Hg Absolute 10.0 1 0.5 (2) The following condenser pressure trip shall be operable during reactor power operations when steam drum pressure is at least 500 psig or higher:
Reactor Scram, Inches Hg Absolute 8.0 1 0.5 4.2.3 Turbine Bypass Control System (a) Design Features Shall Be as Follows:
Flow Capacity at 1015 Psia, Pounds 739,000 per Hour i
Flow Capacity at 1465 Psia, Pounds 963,000 per Hour Maximum Speed, Full Valve Stroke, Approximately 0.2 Seconds-l Amendment No. 14 4-9
O O
l 6.1.3 (Contd.)
J l
(c) With the mode switch in the " shutdown" position, both the scram circuit and the control rod withdrawal circuit are open.
The ventilating duct circuit power supply is trans-ferred to a point which provides penetration closure pro-tection through signals from "high containment sphere pressure" and " low water level in reactor vessel." This permits normal ventilation in the containment sphere during shutdown when the control rods are held in the full-in position.
M:nc of the reactor safety system signals are bypassed since there is no need to withdraw control rods.
1 (d) With the mode switch in the refuel position and the crane positioned over the reactor vessel, crane operation is prevented if any one rod is withdrawn from full-in position.
(e) High condenser pressure reactor trip is automatically bypassed any time steam drum pressure is below a set point maximum of l
500 psig.
i 6.1. 4 Related Systems 6-7 Amendment No. 14
O O
6.3.2 Refueling Operation Controls Interlocks shall be provided to prevent all motion with any of the refueling cranes (namely, jib cranes, transfer I
cask winch) which are positioned over the reactor vessel whenever any control rod is not fully inserted in the core and the mode selector switch is in the " refuel" position.
6.3.3 0perating Requirements (a) All reactor refueling safety system sensors and trip devices shall be functionally tested at each major re-fueling shutdown and shall be maintained in the speci-fied condition during all refueling operations.
(b) The refueling operation controls including position interlocks shall be functionally tested at each major l
refueling shutdown.
6.4 PLANT MONITORING SYSTEMS The plant monitoring systems include the process radiation monitoring systems and the area monitoring system.
6.4.1 Process Radiation Monitoring Systems The process radiation monitoring systems consist of the air ejector off-gas monitoring system including the fuel rupture detection system; stack-gas monitoring system, the emergency condenser vent monitor, and process liquid monitor system.
(a) Air Ejector Off-Gas Monitoring Systems Continuous monitoring of the air ejector off-gas radio-activity shall be provided by either tv.o ion chamber type systems or two single-channel gamma scintillation spectrometer systems designed to detect noble gas fission products indicative of a fuel element rupture.
One system (either ion chamber or scintillation detector) w;11 l
always be in service with an identical system as an operational spare. The sampling system shall be designed j
to hold up the gas sample to allow time for the decay of Nitrogen-16 and other short-lived activation gases.
The off-gas monitoring channels shall be calibrated so that the indicated and recorded count rate output of the channel in service, combined with the off-gas flow, permit Amendment No. 14 6-10 1
O o
6.4.2 (Cont'd)
(b) Two of these nineteen area monitors shall be located in the vicinity of the fuel storage areas to provide gamma monitoring of the fuel storage areas and refueling operations.
Local alarms shall be provided for these monitors, and alarm settings shall be in accordance with the provisions of 10 CFR 70.
However, notwithstanding the requirements of Section 70.24(a)(1),
alarm settings may be raised above 20 mR/hr as long as the overall detection criterion in Section 70.24(a)(1) is satisfied and the requirements specified in paragraph 6.4.3(e) below are met.
(c) At least five environmental film monitoring stations shall be provided for determining the integrated gamma dose rate in the site environs. These stations shall be placed on an arc of about 1,350 meters from the stack.
6.4.3 Operating Requirements (a) At least one of the two air ejector of t ;as monitoring systems shall be in service during power operation and set to initiate closure of the off-gas isolation valve as described below. Alarms normally shall be set to annunciate in the control room if the off-gas radioactivity reaches a level that corresponds to a stack release of 0.1 curie per second. At stack releases above 0.1 curie per second, the alarm shall be set approximately a factor of two above the expected off-gas release rate bu[* g no event above that level corresponding to a stack release of y---
curie per second where E is the a g ge gamma energy per disintegration (MEV/ dis).
If the limit of curie per second is exceeded, reactor power shallbeimmediathlyreducedsuchastomeetthelimits.
The monitors shall be set to initiate closure of the off-gas isolation valve (af ter a time adjustable from 0 to 15 minutes) if '.h vii gas radioactivity reaches a level that would correspond to a stack release rate of ten curies per second. Off-gas samples shall be taken monthly during power operation and analyzed for calibration of the off-gas radiation monitors. The automatic closure function of the monitors shall be tested monthly during power operation.
~
(b) The stack-gas monitoring system shall normally be in service.
Adequate spare parts shall be on hand to allow necessary repairs to be made promptly. The alarm normally shall be set to annunciate in the control room at a level that corresponds to a stack release rate of 0.1 curie per second.
At stack release rates above 0.1 curie per second, the alarm shall be set approximately a factor of g above the expected stack release rate, but in no event above 7 curien per second.
l The calibration of the system shall be checked at least monthly.
The particulate filter and iodine filter shall be analyzed at least weekly.
Amendment No. lA 6-13
.,l1.2 (Contd.)
O o
Conductivity
('licro.ho/cm)
Maximum 5
f Maximum transient
- 10 pH (Lower and Upper limits) 4.0 and 10.0 Chloride Ion (Ppm) 1.0 Equilibrium Halogen Radioactivity (uc/ml) 35 Boron (Ppm) 109 (c) Leakage Limits 1.
If the primary coolant system leakage exceeds 1 gpm and the source of leakage is not identified, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a cold butdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
If leakage from the primary coolant system exceeds 10 gpm, the reactor shall be placed in the hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a cold shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
The high energy line sections identified in Table 9-3b shall be maintained free of visually observable through-wall leaks.
(a)
If a leak is detected by the surveillance program of Table 9-3b, efforts to identify the source of the leak shall be started immediately.
(b)
If the source of leakage cannot be identified within eight hours of detection or if the leak is found to be from a break in the sections identified in Table 9-3b, the reactor shall be in a cold shutdown within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
4.1.3 Primary System Shielding Reactor sgielding is ordinary concrete with a density of approximately 150 lb/ft.
Thickness varies in plan and elevation to suit structural requirements. The shielding thickness directly opposite the core shall be apprcximately 9 feet, 6 inches.
The control rod drive room, which is direct,1y b,eneath the reactor, has ordinary concrete walls which shall be approximately 4 feet thick. A removable shield plug of a thickness 4 feet, 6-1/2 inches, consisting of 4 feet, 4 inches of concrete and 2-1/2 inches of lead, shall close the opening above the top of the reactor.
l
- Conductivity is expected to increase temporarily after startups from cold shutdown.
The maximum transient value here stated is the maximum pennissible and applies only to the period subsequent to a cold shutdown between criticality and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 20% rated power.
Amendment No. 14
_