ML20147H328
| ML20147H328 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  (NPF-011, NPF-018) |
| Issue date: | 03/04/1988 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML20147H333 | List: |
| References | |
| NPF-11-A-055, NPF-18-A-036 NUDOCS 8803080455 | |
| Download: ML20147H328 (10) | |
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COMMONWEAliH EDIS0N COMPANY DOC /ET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 55 License No. NPF-11 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by the Comonwealth Edison Company (thelicensee),datedAugust 24, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter It B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.
There is reasonable assurance: (1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; 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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the enclosure to this license amendment and para-graph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:
(2) TechnicalSpecificationsandEnvironmentalProtectionPM The Technical Specifications contained in Appendix A, as revised through Amendment No. 55, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifi-cations and the Environmental Protection Plan.
8803000455 000304 PDR ADOCK 05000373 P
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This amendment is effective 15 days after the date of issuance.
1 FOR THE NUCLEAR REGULATORY COMMISSION l
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Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III.
IV, Y and Special Projects
Enclosure:
Changes to the Technical Specifications Date of Issuance: March 4, 1988
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ENCLOSURE TO LICENSE AMENDMENT NO. 55 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
REMOVE INSERT 3/4 4-2 3/4 4-2 8 3/4 4-1 B 3/4 4-1 i
l i
R_EACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 Each of the above required jet pumps shall be demonstrated OPERABLE prior to the THERMAL POWER exceeding 25% of RATED THERMAL POWER and 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 measuring and recording each of the below specified parameters and verifying that no-two of the following conditions, occur when both recirculation loops are operating with balanced flow, a.
The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characteristics for two recirculation loop operation.
b.
The indicated total core flow differs by more than 10% from the established total core flow value derived from either the:
1.
Established THERMAL POWER-core flow relationship, or l
2.
Established core plate differential pressure-core flow relation-ship for two recirculation loop operation.
c.
The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established two recirculation loop operation patterns by more than 10%.
4.4.1.2.2 During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE 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 that no two of the following conditions occur:
a.
The indicated recirculation loop flow in the operating loop differs by more than 10% from the established single recirculation flow control valve position-loop flow characteristics.
b.
The indicated total core flow differs by more than 10% from the established total core flow value from single recirculation loop flow measurements derived from either the:
1.
Established THERMAL POWER-core flow relationship, or 2.
Established core plate differential pressure-core flow relationship for two recirculation loop operation.
c.
The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established single recirculation t
loop operational patterns by more than 10%.
LA SALLE - UNIT 1 3/4 4-2 Amendment No. 55
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and been found to be acceptable, provided the unit is operated in accordance with the single recirculation loop operation Technical Specifications herein.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design-basis-accident by increasing the blowdown area and reducing the capability of reficoding the core, thus, the requirement for shutdown of the facility with a jet pump inoperable.
Jet pump failure can be detected by monitoring jet pump performance on a prescribed scheduled for significant degradation.
During dual loop operation, the jet pump operability surveillance should be performed with balanced drive flow (drive flow mismatch less than 5%) to ensure an accurate indication of jet pump performance.
Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criterion.
The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
Where the recir-culation loop flow mismatch limits cannot be maintained during the recir-l culation loop operation, continued operation is permitted in the single recirculation loop operation mode.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50*F of each other prior to startup of an idle loop.
The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145*F.
The possibility of thermal hydraulic instability in a BWR has been investi-gated since the startup of early BWRs.
Based on tests and analytical models, it has been identified that the high power-low flow coraer of the power-to-flow map is the region of least stability margin.
This region may be encountered during startups, shutdowns, sequence exchanges, and as a result of a recircula-tion pump (s) trip event, To ensure stability, single loop operation is limited in a designated restricted region (Figure 3.4.1.1-1) of the power-to-flow map.
Single loop operation with a designated surveillance region (Figure 3.4.1.1-1) of the power-to-flow map requires monitoring of APRM and LPRM noise levels.
3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code.
A total of 18 OPERABLE safety /
relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.
LA SALLE-UNIT 1 B 3/4 4-1 Amendment No. 55 i
UNITED STATES E
NUCLEAR REGULATORY COMMISSION o
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COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 License No. NPF-18 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by the Comonwealth Edison Company (thelicensee),datedAugust 24, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the 4
provisions of the Act, and the regulations of the Comission; i
C.
There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR i
Chapter I; l
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.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the enclosure to this license amendment and para-graph 2.C.(2) of the Facility Operating License No NPF-18 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 36, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specif1-cations and the Environmental Protection Plan.
m
-2 3.
This amendment is effective 15 days after the date of issuance.
FOR THE NUCLEAR REGULATORY C0m !SSION x==Y A &
Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, Y and Special Projects
Enclosure:
Changes to the Technical i
Specifications Date of Issuance: March 4,1988 I
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4 ENCLOSURE TO LICENSE AMENDNENT NO. 36 FACILITY OPERATlhG LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
REMOVE INSERT 3/4 4-3 3/4 4-3 B 3/4 4-1 B 3/4 4-1
7
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4 REACTOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
$ TION:
With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 Each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER and 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 measuring and recording each of the below specified parameters and verifying that no two of the following conditions occur when both recircula-tion loops are operating with balanced flow.
l a.
The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characteristics for two recirculation loop operation, b.
The indicated total core flow differs by more than 10% from the established total core flow value derived from either the:
1.
Established THERMAL POWER-core flow relationship, or 2.
Established core plate differential pressure-core flow relationship for two recirculation loop operation.
c.
The indicated diffuser-to-lower plenum differential pressure o-any individual jet pump differs from established two recirculation loop operation patterns by more than 10%.
4.4.1.2.2 During single recirculation loop operation, each of the above required jet pumps shall be demonstrated OPERABLE 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 that no two of the following conditions occur:
a.
The indicated recirculation loop flow in the operating loop differs by more than 10% from the established single recirculation flow l
control valve position-loop flow characteristics.
b.
The indicated total core flow differs by more than 10% from the established total core flow value from single recirculation loop flow measurements derived from either the:
1.
Established THERMAL POWER-core flow relationship, or 2.
Established core plate differential pressure-core flow relation-ship for two recirculation loop operation.
c.
The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established single recirculation loop by more than 10%.
LA SALLE - UNIT 2 3/4 4-3 Amendment No. 36
J 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated and been found to be acceptable provided the unit is operated in accordance with the single recirculation loop operation Technical Specifications herein.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design-basis-accident by increasing the blowdown area and reducing the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.
Jet pump failure can be detected by monitoring jet pump performance on a prescribed scheduled for significant degradation.
During dual loop operation, the jet pump operability surveillance should be performed with balanced drive flow (drive flow mismatch less than 5%)
to ensure an accurate indication of jet pump performance.
Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criterion.
The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
Where the recir-culation loop flow mismatch limits cannot be maintained during the recir-l culation loop operation, continued operation is permitted in the single recirculation loop operation mode.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop.
The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
Since the coolant in the bottom of the vessel is at a lower temperature than the water in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145'F.
The possibility of thermal hydraulic instability in a BWR has been investi-gated since the startup of early BWRs.
Based on tests and analytical models, it has been identified that the high power-low flow corner of the power-to-flow map is the region of least stability margin.
This region may be encountered during startups, shutdowns, sequence exchanges, and as a result of a recircula-tion pump (s) trip event.
To ensure stability, single loop operation is limited in a designated restricted region (Figure 3.4.1.1-1) of the power-to-flow map.
Single loop operation with a designated surveillance region (Figure 3.4.1.1-1) of the power-to-flow map requires monitoring of APRM and LPRM noise levels.
3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves operate to prevent the reactor coolant system from being pressurized above the Salety Limit of 1325 psig in accordance with the ASME Code.
A total of 18 OPERABLE safety /
relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
Demonstration of the safety / relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.
LA SALLE - UNIT 2 B 3/4 4-1 Amendment No. 36