ML19336A636
| ML19336A636 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 10/16/1980 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML19336A635 | List: |
| References | |
| NUDOCS 8010300377 | |
| Download: ML19336A636 (11) | |
Text
D_0CKET No. 50-213 O
ATTACID1ENT 1 HADDAM NECK PLINT PROPOSED REVISIONS TO TEClINICAL SPECIFICATIONS 1
8010300377 OCTOBER, 1980
3.3 REACTOR COOLANT SYSTEM OPERATIONAL COMPONENTS Applicability:
Applies to the operating status of the reactor coolant system.
Objective:
To specify those limiting conditions for operation of the reactor coolant system which must be met to insure safe reactor operation.
Specification:
A.
At least one pressurizer code safety valve shall be in service whenever the reactor is subcritical 0
and the reactor coolant system is above 375 F or 350 psig except during hydrostatic tests.
B.
One or more reactor coolant pumps or the residual heat removal system shall be in operation when reductions are made in the-boron concentration of l
the reactor coolant.
C.
The reactor shall not be critical unless the following conditions have been satisfied:
(1) Three self-actuated, spring loaded safety valves, having a combined relieving capability of 720,000 f/hr. shall be in service and shall be in accordance with Section VIII of the ASME Boiler and Pressure Code.
(2) Above 1 percent of Nominal Operating Power Level, at least one reactor coolant pump operating.
(3) Above 10 percent of Nominal Operating Power Level, at least three reactor coolant pumps operating.
(4) Above 65 percent of Nominal Operating Power Level. Four reactor coolant pump operating.
(5) Two steam generators are capable of performing their heat transfer function.
(6) Two power operated relief valves (PORV's) and their associated block valves shall be operable except that:
a.
With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />'either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s);
otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
D 5
b.
With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s); otherwir.e, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTIDWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
(7) The pressurizer shall be operable with at least 150 KW pressurizer heaters.
~
a.
With the pressurizer inoperable due to the inoperability of both emergency power supplies to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in the HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
D.
Each steam generator in a non-isolated reactor coolant loop shall be restored to operable status prior to increasing Tave above 200 F.
When starting a reactor coolant pump, and the reactor coolant cold leg temperature in any non-isolated loop is at or below 340 F, the secondary water temperature of each non-isolated steam generator shall not be more than 200F higher than the water temperature of each of the non-isolated reactor coolant cold legs.
E.
Ti. RCS Overpressure Protection System (OPS) shall be in operation when the RCS temperature is below 340 F unless the RCS is vented through a minimum opening of three (3) inches (nominal diameter) or its equivalent. If one or more of the relief trains is taken out of service and the RCS is not vented, the following actions shall be taken:
(1) With one relief train inoperable, either restore that train within. days, or depressurize and vent the RCS through a minimum 3 inch diameter or equivalent opening within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both relief trains have been restored to operable status.
All pressurizer code safety valves are to be in service 1
prior to criticality to permit the design relieving flow to occur if required.
Part C of the specification requires that a sufficient number of reactor coolant pumps be operating to provide core cooling in the event loss of flow occurs. The flow provided in each case will keep DNB well above 1.30
.as discussed in FDSA Section 10.3.2.
Therefore, cladding damage release of fission products to the reactor coolant cannot occur.
By limiting the temperature differential between the 4
primary and secondary sides to twenty (20) degrees in Part D, the resulting pressure transient will be prevented by the RCS OPS (See Reference 1) from exceeding the limits in Specification 3.4.
As described in Reference (1), the RCS OPS, in conjunction with administrative controls, prevents exceeding the temperature and pressure limits in Specification 3.4 while RCS temperature is under 340 F.
Considerations have been incorporated to provide for the inoperability of one or more relief trains (relief valve, motor operated isolation valve, and associated instrumentation) when the RCS OPS is required to be operable.
The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.
The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.
The requirement that (150) kw of pressurizer heaters-and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at HOT STANDBY.
h
(2) With both relief trains inoperable, depressurize and vent the RCS through a minimum 3 inch diameter opening within 8-hours; maintain the RCS in a vented condition until both relief trains have been restored to operable status.
(3) A 30 day report shall be prepared and submitted to the Ccamission if either actions (1) or (2) are used to mitigate inoperability ~ one or both relief trains.
F.
Whenever the reactor is in Mode 3 the following conditions shall be met:
1.
At least two of the reactor coolant loops listed below shall be OPERABLE:
a.
Reactor Coolant Loop (1) and its associated steam generator and reactor coolant pump, b.
Reactor Coolant Loop (2) and its associated steam generator and reactor coolant pump, c.
Reactor Coolant Loop (3) and its associated steam generator and reactor coolant pump, d.
Reactor Coolant Loop (4) and its associated steam generator and reactor coolant pump.
2.
At least one of the above coolant loops shall be in operation.
a.
With less than the above required reactor coola.t loops OPERABLE, restore the requited loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next 12 hou's.
b.
With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to. return the required coolant loop to operation.
G.
Whenever the reactor is in Mode 4 or 5 the following conditions shall be met:
1.
At least two of the coolant loops listed below shall be OPERABLE:
a.
Reactor Coolant Loop (1) and its associated steam generator and reactoc coolant pump, b.
Reactor Coolant Loop (2) and its associated steam generator and reactor coolant pump,
c.
Reactor Coolant Loop (3) and its associated steam generator and reactor coolant pump, d.
Reactor Coolant Loop (4) and its associated steam generator and reactor coolant pump, e.
Residual Heat Removal Loop (A)**
f.
Residual Heat Removal Loop (B)**
2.
At least one of the above coolant loops shall be in operation.***
a.
With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 % nts.
b.
With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
- The normal or emergency power source may be inoperable in MODE 5.
- All reactor coolant pumps and decay heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 100F below aaturation temperature.
Basis:
Each of the pressurizer code safety valves is designed to relieve 240,000 lbs per hr. of saturated steam at the valve set point. They are described more fully in FDSA Section 5.2.2.
Below 375 F and 350 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no decay heat were remove,d by any of the means available, the amount of steam which could be generated at safecy valve relief press./e would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization.
When the boron concentration of the reactor coolant system is to be changed, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the primary system volume in approximately one-half hour.
J In MODE 3, a single reactor coolant loop pro rides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
In MODES 4 and 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR
~
pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
'r-
3.13 REFUELUG Applicability:
Applies to operating limitations during refueling operations.
Obiective:
To insure that no incident could occur during refueling operations that would affect public health and safety.
Spec if icat ion:
A.
Radiation levels in the containment and fuel storage building shall be monitored continuously.
B.
Core suberitical neutron flux shall be continuously monitored by at least two neutron monitors, each with continuous visual and audible indication available, whenever core geometry is being changed.
When core geometry is not being changed, at least one neutron flux monitor shall be in service.
C.
Whenever the water level in the refueling cavity is less than 21 feet above the flange of the reactor pressure vessel, there shall be two residual heat removal pumps and heat exchangers available. Also, with less than the required depth of water described above, suspend all operations involving movement of fuel assemblies or control rods within the reactor pressure vessel.
At least one RHR pump and heat exchanger shall be in operation except that the residual heat removal system may be removed from operation for up to I hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during performance of core alterations in the vicinity of the reactor pressure vessel hot legs.
With less than one residual heat removal pump and heat exchanger in operation except as described above, suspend all operation involving any increase in reactor decay heat load or a reduction in boron concentration of the reactor coolant system. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
D.
During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration shall be maintained at' not less than that required to shut down the core to ak
- 0.92 (see Specification 3.11).
i j
E.
One charging pump capable of injecting borated water to the reactor coolant shall be available at'all times when changes in core geometry are taking place.
F.
Whenever new fuel is added to the reactor core, a 1/M plot shall be maintained to verify the sub-criticality of the core.
4 E
a 4
4 1
i 2
4
i
~
G.
Direct communication between the control room and the refueling cavity manipulator crane shall be available whenever changes in core geometry are taking place.
H.
Spent fuel casks shall not-be hcndled above the spent fuel pool or its edge except as provided in Section 3.13.1, until such time as NRC has received and approved the spent fuel cask drop evaluation.
I.
After April 23, 1980, a spent fuel cask may be brought into the spent fuel building and may be moved into or over the spent fuel pool a total of ten times in order to remove fuel from the pool for study at an off-site laboratory, or to return the fuel from the laboratory to the pool.
Movement of the spent fuel cask under the provisions of this paragraph is conditioned on compliance (by the licensee) with all commitments made by the licensee in its letters to the NRC dated April 18, 1980 and April 23, 1980.
In addition, all fuel within the spent fuel pool shall have decayed for at least 90 days before a spent fuel caeb.is handled above the pool.
~
.- - - -. ~.
, j 4
t
~
Basis:
The equipment and general provedures to be utilized during refueling are discussed in the Facility i
Description and Safety Analysis. Detailed instructions will be available for use by refueling personnel.
These instructions, the above-specified precautions, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance.that no incident could occur during the refueling operations that would result in a hazard to public health and safety. Whenever no change is being made in core geometry, one flux monitor is suf ficient.
.his permits maintenance of the instrumentation. Continuous monitoring of radiation levels (A"A above) and neutron flux provides immediate indication of an unsafe condition. The residual 2
j heat pump is used to maintain a uniform boron concentration. The shutdown margin indicated in l
Part D will keep the core suberitical, even if all control rods were withdrawn from the core. Weekly checks of refueling water boron concentration insure i
the proper shutdown margin. Part G allows the control room opetator to inform the manipular operator of any impending unsafe condtiion detected from the main
].
control board indicators during fuel movement.
I The requirement that at least one residual heat removal pump and heat exchanger be in operation ensures that (1) 4
. sufficient cooling capacity is available to remove decay
~
l heat and maintain the water in the reactor pressure i
vessel below 140 F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the 'effect of a boron 3
dilution inc: dent' and prevent boron stratification.
1
(
The requirent it to have two RHR pumps and heat exchangers operable when there is less than 21 feet of water above i
the reactor ptessure vessel flange ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain water in the reactor pressure vessel below 1400 as required during refueling mode and (2) i sufficient coolant circulation is maintained through the t
reactor core to minimize the effect of a boron dilution incident and prevent boron stratification and (3) that sufficient water depth is available to remove 99% of the i
assumed 10% iodine gap activity released from the rupture of.an irradiated fuel assembly.
I.
I
References:
(1)
FDSA Sect nr 5.2.9 (2)
FDSA Section 7.4 i
~ ___
_.