ML19330B483
| ML19330B483 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/31/1980 |
| From: | METROPOLITAN EDISON CO. |
| To: | |
| Shared Package | |
| ML19330B478 | List: |
| References | |
| NUDOCS 8008040076 | |
| Download: ML19330B483 (14) | |
Text
Thr:3 Mile Icicnd Nuleccr Stctirn, Unit II (TMI-2) n Oparcting Liccn;G No. DPR-73 b
Docket No. 50-320 Technical Specification Change No. 24 The licensee requests that the attached changed pages, 3.1-1, 3.1-2, 3.4-1, 3.7-1, 3.7-2, 3.8-4, B 3/4 4-1, B 3/4 7-1, B 3/4 7-la, be substituted for or added to the existing Technical Specifications.
Reason for Changes The Reactor Coolant System is currently in a natural circulation heat removal mode with heat rejection through the "A" steam generator. The "A" steam generator is steaming to the condenser through the turbine by-pass valve. With the passage of time and the associated reduction of the decay heat generation rate, natural circulation exhibits longer time periods between " burps" while maintaining ade-quate core cooling.
In order to discontinue use of OTSC "A" for removal of reactor decay heat, the licensee proposes the implementation of the Mini Decay Heat Removal System (MDHR System). Removal of reactor decay heat by this method would:
1.
Permit the removal from service of several major BOP systems and equipment, such as:
a.
b.
Main steam system and the "A" steam generator.
c.
Condensate pumps.
d.
Condensate and feedwater systems, e.
2.
Permit the removal of reactor decay heat with a system that is more appropriately sized to the present and future decay heat generation rates.
To facilitate operation of the MDHR system as the primary heat removal system, the Technical Specifications must be amended to reflect MDHR system usage.
With the MDHR system, Long Term "B" and " Loss to Ambient" cooling modes available, many of the components / systems used in the present cooling mode will no longer be necessary for cooling the core. These systems / components include portions of the Feedwater and Condensate System and the Emergency Feedwater pumps.
The Feedwater/ Condensate System including the Emergency Feedwater pumps are no longer needed; therefore, the Licensee requests that they be removed from the Technical Specifications.
Additionally, the licensee requests that the Technical Specifications relating to the Makeup Pumps be removed. The Makeup Pump is designed to inject Makeup j
and seal injection fluid into the Reactor Coolant System (RCS) at normal (2150 psig) pressures. The ultimate mode for TMI-II heat removal will have the RCS at 100 psig, being cooled by a low design pressure system (MDHR system, 235 psig).
8008040 C M
Tha Mak up Pump ch:uld b2 removed from esrvica cnd ths Tcchnicql Sp;cificatiana
' altered to remove Makeup Pumo Operability Requirements. The pump is capable of developing a discharge pressure of about 2800 psig which is mugh higher than the MDHR system design pressure of 235 psig. Therefore, inadvertant operation of the Makeup Pump would threaten HDHR system integrity.
The Licensee plans to retain the option to operate the pump (i.e., for degassing) but plans to eliminate the current focus of the Technical Specifications upon them as a " Backup System Component".
480 Volt Busses 2-35, 2-45, 2-36, 2-46, 2-38, and 2-48, and 4160 Volt Bus 2-5 do not provide electrical power to any electrical loads required to support any of the TMI-II core cooling modes: therefore, the Licensee requests they be removed from the Technicial Specifications.
Safety Evaluation Justifying Change Offsite radiological exposure estimates for a worst case postulated accident have been completed based on the radionuclide activity concentrations present in the RCS on July 24, 1979. This calculation employed the following assumptions:
1.
MDHR System isolation occurs at the inlet and outlet isolation valves.
2.
The fluid contained in the MDHR System (assumed to be 1200 gallons), is dumped to the auxiliary building floor.
3.
No flashing of coolant occurs.
4.
Source terms considered in this evaluation are Iodine, Xenon, and Krypton at RCS concentrations that existed on July 24, 1979.
5.
Airborne release to the atmosphere occurs as a puff release with no cleanup or partitioning. No other isotopes are released.
6.
Atmospheric dispersion values used in the analysis are those in Tsble 2.3-7, TMI-II FSAR (ground level releases).
Results of this analysis indicate a whole body radiological exposure dose at the 4
site boundary of 1.16 X 10 rem. The off-site exposure estimate has been updated to reflect current RCS activity concentrations. The concentration of various radioisotopes in the RCS has decreased since July 24, 1979. Both Iodine and Xenon isotopic concentrations are insignificant.
In this updated estimate, the following assumptions have been employed:
1.
. - Same as above -
2.
- Same as above -
3.
- - Same as above -
4.
Source terms considered in this estimate are particulate radioisotopes (primarily Cesium and Strontium) at RCS concentrations that presently exist.
5.
Airborne release to the atmosphere occurs as a puff release with 1%
partitioning.
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- 9ane as abovo -
The results of this analysis indicate that a total of approximately 12 Ci would be released to the Auxiliary Building atmosphere as a result of this postulated, though
-improbable, accident. This potulated release to the Auxiliary Building atmosphere presents insignificant effects at the site boundary.
o utilizes natural circulation of reactor coolant through OTSG "B" and forced circulation of the secondary side of the steam generator). When in LTC "B"
the reactor decay heat is transferred to the Secondary Services Closed Cooling Water System, and then to Nuclear Services River Water.
In addition to UTC "B", steaming OTSG "A" to the condenser, with natural circulation of the reactor coolant through it, is an additional reactor decay heat removal method until the condenser is taken out of service.
In the casa where MDHR System heat removal capability is lost, reactor coolant bulk temperature will initially increase at a rate of approximately 0.4 F/ hour with =335,000 BTU / hour core heat input.
Our analysis indicates that this tempera-ture increase will gradually diminish such that at approximately 190 F reactor coolant heat up rate will be approximately zero. At this RCS temperature, heat loss to ambient is equal to the decay heat generation rate. Assuming that the time required to shift cooling modes from MDHR to forced circulation of the secondary side of OTSG "B" is =72 hours, reactor coolant system temperature could rise as much as 30 F.
Current RCS temperature in the "A" hot leg is approximately 145 F.
If, in this transient, the RCS temperature rises to approximately 175 F in the hot leg, substantial nargin to the saturation temp-erature exists because the RCS, and the HDHR System, is operated at 100 psig RCS pressure.
MDHR System Overpressurization:
The MDHR system design pressure is 235 psig. Consideration has been given to possible sources.of overpressurization of the system. The potential sources are defined to be:
a.
Makeup pump operation with MDHR in service, b.
Standby Pressure Control (SPC) System failure.
c.
Pressurizer heater operation.
The attached Proposed Technical Specification Change Request deletes the requirement to have an operable makeup pump in service. In order that makeup pump operation with the MDHR system in service can be precluded, the makeup pump breakers will be
" racked out".
Furthermore, prior to operation of a makeup pump, the MDHR system will be taken out of service and isolated from the RCS. This isolation will provide assurance that the potential for MDHR system overpressurization from makeup pump operation is minimized.
Failure of the Standby Pressure Control System to control the pressure of the RCS is possible, though remote. To preclude the remote r r:lbility of overpressurization of the MDHR system from the Standby Pressure Control System, the SPC N bank 2
pressure will be reduced from the present operating range of "1500 psig to 2400 psig" to a new operating range of "225 psig to 400 osig."
This reduction in N bank 2
pressure removes the possibility of MDHR system overpressurization from this source, yet provides the capability for makeup to the RCS even in the event of the most severe cooldown rate in the RCS.
Pressurize heater operation while in a " solid" operating mode on the RCS, does not create the potential for overpressurization of the MDHF system. Pressurizer heater operation, even though the MDHR system is in service, will cause a volumetric expansion of the RCS through the mechanism of heat input into the coolant.
Electrically. energizing all pressurizer heaters (1638 kw) will result in a volumetric expansion of the RCS that will require a relief capacity of 8.6 gpm.
Since each MDHR system relief valve capacity is rated at 53.5 gpm flowrate, pres-
Y, g l curinar h;ctsr oparstion is nst cenzidsred: to' bn a viable ovarpressurization source forlthe MDHR system. 'Furthermore, present operation of the pressurizer heaters'isLperformed.such that all banks of heaters are out of service except one (1) bank.
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ALARA Considerations:-
lIt is recognized.that the MDHR system and the facilities in which it is housed
'have potential to become high radiation areas. In order that.the guidance of Regulatory Guide 8.8 could be applied to the system and facilities, the following design considerations have.been. incorporated:
1.'
Remote flushing connections have been provided throughout the system to permit flushing of components prior to the performance of maintenance functions on that component. Flushing with non-contaminated water callows the washing away of radioactively contaminated materials from the internals ofLthe components, and hence, a reduction of the radiation field produced prior to the performance of operational or maintenance evolutions.
- 2. -
Radiation shielding has been provided throughout MDHR system for the
' equipment and piping to minimize the radiation exposure to personnel.
Shielding ~ between the MDHR pumps has been provided for the reduction of personnel exposure during pump maintenance. Reach rods to permit remote valve operation have been provided for manually-operated valves that will enable the reduction of radiation exposure to operations personnel.
3.
The MDHR system is provided with an inlet filter to protect the pumps from debris that may be entrained in the influent flow stream. This filter, because of the possibility of radioactively contaminated
-loading,'is ' shielded to minimize area radiation levels in the vicinity of the filter.
4.
'The MDHR system employs welded connected to the maximum extent
.possible to minimize' system leakage. Where the possibility of leakage exists, such as at pump seal leak-offs, flanges, and valve stems,
' provisions have been madefor the containment of the leakoff. This design consideration will permit minimizing the radioactive contamination in'the vicinity of the JDHR components. As a result, personnel con-ta'mination potential is minimized.
5.
Operation and observation of the MDHR systen operating parameters is
- performed from a panel that is located in a low radiation area. This design philosophy permits the infrequent-operational adjustments or the gathering or operating data to be performed in a reasonably low radiation level area.
An analysis of-~ expected occupational exposures, as a result of typical NDHR system
. operational ~and maintenance functions has been performed. The results are listed i
~below:
11.-
Valve operator replacement 3.12 Man-Rem
- 2.. fMDHR system filterfreplacement 1.84 Man-Rem-3.~
.MDHR system heat exchanger change-over 0.75 Phn-Rem L4.
JPump seal replacement 4.26 Man-Rem
[
'5.
LInstrument replacement 2.~79 Man-Rem p
-Th2 cnalyrio is brasd on r dionuclida c nctntrations in tha RCS of 57 pCi/cc due to Cs-137 and 11 pCi/cc due to Cs-134. No plate out sources were considered.
It is concluded that the above expected occupational exposures are acceptable and
~that the concept of "as low as reasonably achieved" has been considered and com-plied with.
The removal from service and the Technical Specification changes concerning the RC pump, the makeup pump, the Emergency Feedwater pump, Feedwater/ Condensate System, and the listed 480 B Busses, do not constitute an unreviewed safety question.
Amendment Class The Licensee has determined that because the amendment request involves a single safety concern, it represents a Class III License Amendment (per 10 CFR 170.22).
Therefore, enclosed please find a check in the amount of $4,000.00.
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LIMITING CONDITIONS FOR OPERATION 3.1 WATER INJECTION COOLING AND REACTIVITY CONTROL SYSTEMS 3.1.1 BORATION CONTROL BORON INJECTION 3.1.1.1 at least two systems capable of injecting borated cooling water into the Reactor Coolant System shall be OPERABLE
- with:
a.
One system comprised of:
1.
Delete 2.
One OPERABLE decay heat removal pump.
3.
An OPERABLE flow path from the BWST. The BWST shall contain at least 100,000 gallons of br. rated water at a minimum temperature of 50 F and at a boron concentration of between 3000 and 4500 ppm.
b.
The second system comprised of the Standby Reactor Coolant System Pressure Control System.
APPLICABILITY: When fuel is in the reactor pressure vessel.
ACTION:
With one of-the above required systems inoperable, restore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
BORON CONCENTRATION 3.1.1.2 The reactor coolant shall be maintained at a boron concentration of between 3000 and 4500 ppm and at a temperature above 50 F.
APPLICABILITY:
When fuel is in the reactor pressure vessel.
ACTION None except as provided in Specification 3.0.3.
- Both systems shall be considered OPERABLE when aligned per procedures approved pursuant to Specification 6.8.2.
..THREE MILE-ISLAND - UNIT 2 3.1-1
LIMITING CONDITIONS FOR OPERATION 3.4 -REACTOR COOLANT SYSTEM
_ REACTOR COOLANT LOOPS 3.4.1 The Reactor Coolant System shall be operating in accordance with procedures approved pursuant to Specification 6.8.2.
_ APPLICABILITY: RECOVERY MODE.
ACTION None except as provided in Specification 3.0.3.
SAFETY VALVES 3.4.3 All prgssurizer code safety valves shall be OPERABLE with a lift setting of 2435 PSIG - 1%*.
APPLICABILITY: RECOVERY MODE.
ACTION:
.None except as provided in Specification 3.0.3.
3.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM 3.4.9.1 The Reactor Coolant System shall be maintained at a T of less than 2800F and at a oressure of less than 600 psig.
avg AP_PLICABILITY: When fuel is in the reactor pressure vessel.
' ACTION:
None except as provided in Specification 3.0.3.
- The lift setting pressure shall corresoond to ambient conditions of the valve at nominal operating temperature and pressure.
'THREE MILE ISLAND - UNIT 2 3.4-1
DELETE Figure 3.11 Minimum Boric Acid Tank Contained Volume as a Function of Stored Boric Acid Concentration THREE MILE ISLANQ - UNIT 2 3.1-2 I
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- LIMITING CONDITIONS FOR OPERATION 3.7 PLANT SYSTEMS 3.7.1 FEEDWATER SYSTEM 3.7.1 The long term "B" steam generator cooling system shall be maintained in an OPERABLE status.
APPLICABILITY: RECOVERY MODE.
ACTION:
With the Long Term Cooling "B" Steam Generator cooling system inoperable restore the system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.7.2 SECONDARY CLOSED COOLING WATER SYSTEM 3.7.2.1 At least two secondary services closed cooling water l
system pumps and heat exchangers and the associated flow poth shall be OPERABLE with each pump capable of being powered from separate busses.
APPLICABILITY: RECOVERY MODE.
ACTION:
With only one secondary services closed cooling water pump or only one secondary services heat exchanger OPERABLE, restore the inoperable pump or heat exchanger to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
3.7.3 CLOSED CYCLE COOLING WATER SYSTEM NUCLEAR SERVICES CLOSED CYCLE COOLING SYSTEM 3.7.3.1 At least two independent nuclear services closed cycle cooling water pumps and heat exchangers'and the associated flow path shall be OPERABLE with each pumo-capable of being powered from separate emergency busses.
APPLICABILITY:
RECOVERY MODE.
4
- ACTION:
With only one nuclear services closed cycle cooling water pump or only one nuclear services heat exchanger OPERABLE, restore the inoperable pump or heat exchanger to OPERABLE status within.72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
THREE MILE ISLAND - UNIT 2 3.7-1
m
. o LIMITING CONDITIONS FOR OPERATION DECAY HEAT CLOSED COOLING WATER SYSTEM 3.7. 3. 2 - At least one decay heat closed cooling water loop shall 've OPERABLE.
{
APPLICABILITY: RECOVERY. MODE.
ACTION:
With no' decay heat closed cooling water loop OPERABLE, restore the inoperable
-loop to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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MINI DECAY MEAT REMOVAL SYSTEM 3.7.3.3 At-least one mini decay heat removal pump and heat exchanger shall be OPERABLE.
j APPLICABILITY: When fuel is in the reactor pressure vessel.
ACTION:
i With no mini decay heat removal pump and/or heat exchanger OPERABLE, restore the
. inoperable pump and/or heat exchanger to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Should the inoperable pump and/or heat exchanger not be rendered OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, bring a backup cooling system, into operational readiness within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> l
3.7.4 NUCLEAR' SERVICE RIVER WATER SYSTDI 3.7.4.1 Two independent nuclear service river water loops shall be OPERABLE.
APPLICABILITY:
RECOVERY MODE.
ACTION:
'With one nuclear river water loop inoperable, restore the inoperable loop to OPERABLE, status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
O THREE MILE ISLAND - UNIT 2
'3.7-2 J
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LIMITING CONDITIONS FOR OPERATION 3.8.2 ONSITE POWER DISTRIBUTION SYRTEMS A.C.
DISTRIBUTION 3.8.2.1 The following A.C. electrical busses shall be OPERABLE and energized with tie breakers open (unless closed in accordance with procedures approved pursuant to Specification 6.8.2) between redundant busses:
4160 volt' Emergency Bus # 2-1E and 2-3E 4160 volt Emergency Bus # 2-2E and 2-4E 4160 volt Busses # 2-3, and 2-4 480 volt Emergency Bus # 2-11E, 2-12E and 2-31E 480 volt Emergency Bus # 2-21E, 2-22E and 2-41E 480 valt Busses # 2-31, 2-32, 2-41, and 2-42 120 volt A.C. Vital Bus # 2-1V 120 volt A.C. Vital Bus # 2-2V 120 volt A.C. Vital Bus # 2-3V 120 volt A.C. Vital Bus # 2-4V APPLICABILITY:
RECOVERY MODE.
ACTION:
With less than the above complement of A.C. busses OPERABLE, restore the inoperable bus to OPERABLE status with 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
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~ THREE MILE' ISLAND - UNIT-2' 3.8-4 i
3/4.4 REACTOR COOLANT SYSTD1 BASES' 3/4.4.1 REACTOR COOLANT LOOPS Removal of reactor decay heat will be accomplished by the use of the Mini Decay Heat Removal System.
Operation of the Reactor Coolant Loops in the natural circulation mode will be a backup mode of' heat removal.
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 of saturated steam at the valve setpoint.
3/4.4.9 PRESSURE /TE!PERATURE LIMIT The RCS pressure and te=perature will be controlled in accordance with approved procedures to prevent a nonductile failure of the RCS while at the same time permitting the RCS pressure to be maintained at a sufficient high value to permit operation of the reactor coolant pumps.
e THREE FHLE ISLAND - UNIT 2 B 3/4 4-1
3/4.7 PLANT SYSTEMS BASES 3/4.7.1 FEEDWATER SYSTEM The "B" Steam Generator is required to be maintained in an OPERABLE status since it.is a backup means of removing the decay heat from the reactor coolant-system.
3/4.7.2 SECONDARY SERVICES CLCSED COOLING WATER SYSTEM The secondary services closed' cooling water system is required to be maintained in an OPERABLE condition since it is used to cool the "B" steam generator closed loop cooling system.
3/4.7.3 CLOSED CYCLE COOLING WATER SYSTDI 3/4.7.3.1 NUCLEAR SERVICES CLOSED CYCLE COOLING SYSTDI Operability of the nuclear services closed cycle cooling system is required during operation of the mini decay heat removal system (MDHR) since it provides the heat sink for the MDHR System.
3/4.7.3.2 DECAY HEAT CLOSED COOLING WATER SYSTEM
.The decay heat closed cooling water system is required to be maintained in an OPERABLE status since it is provided to remove heat from the DHR system which is being maintained OPERABLE in a backup status for possible core cooling.
THREE MILE ISLAND - UNIT 2 B 3/4 7-1
e 3/4.7 PLANT SYSTEMS
_ BASES 3/4.7.3.3 MINI DECAY HEAT ret 3 VAL SYSTDi The mini decay heat removal system is required to be maintained in an OPERABLE status to enable the removal of the decay heat from the reactor.
3/4.7.4 NUCLEAR SERVICE RIVER WATER SYSTEM The nuclear service river water system uses river water to cool the nuclear services closed cycle cooling system, the secondary services closed cooling water system, and decay heat closed cooling water system; therefore, it must be OPERABLE too.
This system rejects its heat to the river as the ultimate heat sink.
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THREE MILE ISLAND - UNIT 2 B ' 3/4 7-1 a
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