ML20053F027
| ML20053F027 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 06/04/1982 |
| From: | J. J. Barton GENERAL PUBLIC UTILITIES CORP. |
| To: | Snyder B Office of Nuclear Reactor Regulation |
| References | |
| 4400-82-L-0079, 4400-82-L-79, NUDOCS 8206100322 | |
| Download: ML20053F027 (8) | |
Text
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GPU Nuclear h.
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P.O. Box 480 l
Middletown, Pennsylvania 17057 717-944-7621 Writer's Direct Dial Number:
June 4, 1982 4400-82-L-0079 TMI Program Office Attn:
Dr. B. J. Snyder, Program Director U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 2 (TMI-2)
Operating License No. DPR-73 Docket No. 50-320 Recovery Operations ?lan Change Request Number 15 Standby Pressure Control System The attached request to change the Standby Pressure Control System (SPCS) limits on water level and valve isolation is submitted for your review and approval.
These changes will allow temporary variations in surge tank water level and temporary isolation of the SPCS per NRC approved procedures for preparing the RCS for the
" quick look" project.
This change is required to gain valuable information on the condition of some of the reactor vessel internals.
The information will be used to aid the preparations for the reactor vessel head lift and fuel removal.
Sincerely, s
/
Acting Director TMI-2 JJB:RBS:djb Attachment cc:
L. H. Barrett, Deputy Program Director - TMI Program Office 8206100722 820604 PDR ADOCK 05000320 P
PDR GPU Nuclear is a part of the General Public Utilities System
I.
Recovery Operations Plan Change Request No.15 The licensee requests that the attached page 4.1-2 of the Recovery Operations Plan replace the existing page 4.1-2 of the Recovery Operations Plan and that the attached page B 3/4 4-1 replace existing basis page B 3/4 4-1.
II.
Reason For Change The proposed change will allow manipulation of the RCS and surge tank water level and isolation of the standby reactor pressure control system from the RCS.
These manipulations are necessary to permit RCS depressurization and lowaring the water level of the RCS which in turn is required to support core recovery efforts, the first of which will be the insertion of a TV camera into the reactor vessel via a lead screw penetration. Knowledge gained by the insertion of a camera into the upper plenum region will assist in the development of plans and equipment for core recovery efforts.
III. Safety Evaluation Justifyina Change The removal of lead screws is one in a series of steps in preparation for head removal and subsequent defueling. The insertion of a TV camera in the penetration for a lead screw will yield valuable information for the planning of future defueling operations. The insertion will take place with the RCS and steam generators depressurized and partially m
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draineo. Precautions will De taken to ensure that Doron concentrations are controlled in the RCS, that adequate cooling exists for the decay heat remaining in the core, and that leakage is accommodated with RCS grade makeup with the proper boron concentration.
Baron dilution will be further addressed in the safety evaluation for the quick looK project.
The proposed operations will take place with the RCS depressurized and cold and with containment integrity maintained.
Generally, the major evolutions for the quick look are as follows:
1.
The RCS will De depressurized to ~ 60 psig.
2.
The hot legs and pressurizer will be ventea.
3.
The SPC will De isolated.
4.
The RCS will be depressurized to : O psig in the hot legs.
5.
Add N blanket to the hot legs (: 2-5 psig) 2 6.
Drain and depressurize the RCS to : O psig in the pressurizer.
7.
AN blanket will be added to the pressurizer.
2 8.
A CRDM will be vented ano sampled.
9.
The RCS will be drained to between 333'-6" to 335'-0 level.
- 10. Remove lead screw and perform quick look project.
11.
Sample RCS.
The planned oecoupling of the control rod drives and the insertion of a TV camera will have no significant effect on the shutdown margin of tne core.
The draining of the RCS and Steam Generator will Iequire the transfer of radioactive effluents to storage. Tne venting of the RCS
will release gaseous activity.
The volume of free gas in the RCS is 3
estimated to De 45 ft at 108 psig. Details of the metnods to De used to control these effluents will be included in the safety evaluation that will De prepared to support the implementing proceoure(s).
NRC approval of the regJired test proceaures will be obtained oy GPU in accordance with technical specification 6.o.2.
A detaileo safety evaluation will be provided to the NRC staff for review prior to NRC approval of the required procedures.
The isolation of the standoy pressure control system (SPCS) operating mode for the proposed system operation will nave no effect on reactor safety.
The SPCS was initially designed to be the primary means for controlling RCS chemistry and inventory as well as RCS pressure to prevent culk boiling and to prevent loss of natural circulation due to void creation and vapor clock of circulation paths. Currently it is primarily used as a method of monitoring RCS integrity and controlling RCS chemistry. With three years of cecay, the concern for removal of decay heat has diminished considerably since the " loss to amolent" heat removal mode has been proven to De fully effective. Tne ouiloup of RCS temperature cue to decay heat addition to the lowered RCS level will be slow (<50F/ day). By opening valve SPC-V-71, the SPCS can ce useo to control decay neat buildup oy inventory acaition or, if aesireo, cecay heat can be controlled by operating the Mini Decay Heat System (MDHR),
should losses to ambient ce insufficient.
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This evolution will change the RCS control mode from a pressure control mode to a level control mode.
In the level control mode, the SPCS is not necessary and in fact, must be isolated from the RCS. During the course of this test RCS pressure will be decreased.
This decrease in pressure should reduce the driving force for an already low (=.1 gpm) leak rate. The SPCS will be available to control inventory and chemistry, should this become necessary. Additionally, there is no longer a need for the SPCS to fulfill the original design intent of pressure control for NPSH control and void control to prevent flow blockage since primary pumps are no longer needed and void creation is considered unlikely due to low RCS bulk temperatures.
Therefore it is acceptable to close SPC-V-71 under the controlled circumstances postulated here.
The monitoring for RCS leakage during this evolution will be performed by a backup monitoring system, enabling the SPCS to be isolated for the period of time involved in the proposed operation without losing RCS leakage monitoring.
The backup level monitoring system uses a level i
transmitter connected to the normal decay heat removal line. The transmitter will indicate the head of water in the RCS above the center line of the reactor vessel hot leg nozzle. Indication will be a digital readout and strip chart recorder in the control room.
The system will l
compensate for changes in RCS Nitrogen overpressure.
The range of the system will be O to 600 inches with an accuracy of i 5 inches. A local l
1evel indicator will be used in parallel with the above system. The range of the local system is 0 to 600 inches 1 6 inches of water.
These systems will give adequate notification of any level change to allow time for operator action to correct the situation.
Thus, for the above discussed reasons, the proposed change allowing isolation of the SPC system and manipulation of the RCS and Surge Tank water levels for the proposed inspection is essential to present core recovery plans and can be done without adversly affecting the health and safety of the public.
l l
(
3/4.4 REACTOR COOLANT SYSTEM 8ASES 3/4.4.1 REACTOR COOLANT LOOPS Several alternative methods are available for removal of reactor decay heat.
These methods include use of the Mini Decay Heat Removal System, the
" Loss to Ambient" cooling mode.
Elther of these cooling methods provides ade-l quate cooling of the reactor and each method is available for decay heat removal.
Procedures have been prepared and approved for use of these cooling methods.
3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig.
Each safety valve is designed to relieve 348,072 lbs per hour cf saturated steam at the valve's setpoint.
3/4.4.9 PRESSURE / TEMPERATURE LIMIT The RCS pressure and temperature will be controlled in accordance with
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approved procedures to prevent a nonductile failure of the RCS.
l Reactor coolant chemistry surveillance requirements are included in the Recovery Operations Plan.
These req'uirements provide assurance that localized corrosion or pitting in crevice areas, which could tend to promote stress corrosion cracking in heat affected zones of welds in stainless steel piping or components, will not occur.
This assurance is provided by maintaining the reactor coolant dissolved oxygen concentration and pH to within the specified limits.
The oxygen concentration must be limited since the chloride concen-tration is relatively high and cannot be reduced due to the unavailability of the purification domineralizers.
Hydrazine is used to control the oxygen concentration in the presence of metallic impurities in the reactor coolant.
1 e
THREE MILE ISLAND - UNIT 2 8 3/4 4-1
SURVEILLANCE REQUIREMENTS BORON INJECTION (Continued) h.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the BWST temperature is at least 50V when the outside air temperature is less that 50T.
i.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (when system is in operation) by verifying that the standby reactor coolant system pressure control system:
1.
Surge tank water volume is filled to between 55% and 804 of tank capacity and the tank is pressurized to the operating RCS pressure 1 25 psig but not higher than 600 psig. The above limits may be varied in accordance with procedures approved pursuant to Tech Spec 6.8.2.
2.
Isolation valves on the discharge side of the water filled tank nearest the reactor coolant system are open, except when closed in accordance with procedures approved pursuant to Tech. Spec. 6.8.2.
Valves closed by procedure shall be verified open within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of the procedure.
3.
The in-service nitrogen supply bank is pressurized to between 225 and 400 psig.
j.
At least once per 7 days by verifying that the standby reactor coolant system pressure control system surge tanks and the charging water storage tank contain borated water with:
1.
A boron concentration or between 3000 and 4500 ppm.
2.
A dissolved gas concentration of less than 15 sec/kg of water.*
k.
At least once per 31 days by verifying that the standby reactor coolant system pressure control system isolation valve on the discharge side of the water filled tank nearest the reactor coolant l
system closes automatically on a tank low level test signal.
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- Dissolved gas concentration for the SPC System is determined by taking a representative sample from the sampling point located downstream of SPC-T-1.
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t THREE MILE ISLAND - UNIT 2 4.1-2 Change No. 9