ML20086U034
| ML20086U034 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/21/1995 |
| From: | Jabbour K NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20086U036 | List: |
| References | |
| NUDOCS 9508030246 | |
| Download: ML20086U034 (3) | |
Text
'
g@ M00 y-
-t UNITED STATES 4
S NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 2066H001
\\*****/
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORfdA DOCKET NO. 50-321 J
EDWIN I. HATCH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 196 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, i
Unit 1 (the facility) Facility Operating License No. DPR-57 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 13, 1994, as supplemented by letters dated January 13 and May 4, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; I
C.
There is reasonable assurance (i) 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 Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common 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 s
have been satisfied.
e 9508030246 950721 PDR ADOCK 05000321 P
. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.
DPR-57 is hereby amended to read as follows:
Technical Snecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 196
, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemerted within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMISSION
~
He art N. Bebkow, Director P ject Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: July 21, 1995
'\\
eS 44 g
je 4
UNITED STATES j
j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 2066H001 s
/
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NVCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendmen+, No.136 License No. NPF-5 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 2 (the facility) Facility Operating License No. NPF-5 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated October 13, 1994, as supplemented by letters dated January 13 and May 4, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) 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 Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common 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
'*' of the Commission's regulations and all applicable requirements
.iave been satisfied.
. 2.
Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.136
, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Herbert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: July 21,1995 e
\\
9
i ATTACHMENT TO LICENSE AMENDMENT NO. 196 FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 J
gg 4
TO LICENSE AMENDMENT NO. 136 FACILITY OPERATING LICENSE NO. NPF-5 1
DOCKET NO. 50-366 4
Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of :hange.
1 Rgmove Paaes Insert Paaes i
Unit 1 3.3-30 3.3-30 3.3-32 3.3-32 B 3.3-89 8 3.3-89 B 3.3-91 B 3.3-91 B 3.3-92 B 3.3-92 B 3.4-50 B 3.4-50 B 3.4-50a B 3.4-50b l
B 3.4-52 B 3.4-52 Unit 2 3.3-31 3.3-31 3.3-33 3.3-33 8 3.3-89 B 3.3-89 B 3.3-91 B 3.3-91 B 3.3-92 B 3.3-92 B 3.4-50 B 3.4-50 B 3.4-50a B 3.4-50b B 3.4-52 B 3.4-52
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ATWS-RPT Instrumentation 3.3.4.2 3.3-INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation LCO 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation function listed below shall be OPERABLE:
a.
Reactor Vessel Water Level -- ATWS-RPT Level; and l
b.
Reactor Steam Dome Pressure -- High.
APPLICABILITY:
MODE 1.
ACTIONS
____________________________________-NOTE-------------------------------------
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.I Restore channel to 14 days inoperable.
OPERABLE status.
98 A.2
NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
Place channel in 14 days trip.
(continued) i HATCH UNIT 1 3.3-30 Amendment No. 196
a ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST.
92 days SR 3.3.4.2.3 Perform CHANNEL CAllBI'.ATION. The 18 months Allowable Values sheil be:
a.
ATWS-RPT Level: 2: -73 inches; and b.
Reactor Steam Dome Pressure - High:
s 1095 psig.
SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
\\
HATCH UNIT 1 3.3-32 Amendment No. 196
ATWS-RPT Instrumentation l
B 3.3.4.2 B 3.3 INSTRUMENTATION B 3.3.4'.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System-initiates an RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level -- ATWS-RPT Level or Reactor Steam Dome Pressure -- High setpoint is reached, the recirculation pump drive motor breakers trip.
The ATWS-RPT System (Ref.1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT.
The channels I
include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.
The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure -- High and two channels of Reactor Vessel Water Level -- ATWS-RPT Level in l
each trip system.
Each ATWS-RPT trip system is a two-out-of-two logic for each Function.
Thus, either two Reactor Water Level -- ATWS-RPT Level or two Reactor l
Pressure -- High signals are needed to trip a trip system.
The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor breakers).
There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers.
The Output of each trip system is provided to both recirculation pump breakers.
(
)
(continued)
HATCH UNIT 1 B 3.3-89 Amendment No.196
ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE The individual functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1 to protect against common mode failures of the LCO, and Reactor Protection System by providing a diverse trip to APPLICABILITY mitigate the consequences of a postulated ATWS event. The (continued)
Reactor Steam Dome Pressure - High and Reactor Vessel Water Level - ATWS-RPT Level Functions are required to be OPERABLE l
in MODE 1, since the reactor is producing significant power and the recirculation system could be at high flow.
During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event.
In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event.
Therefore, the ATWS-RPT is not necessary.
In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible.
In MODE 5, the one rod out interlock ensures that the reactor remains subtritical; thus, an ATWS event is not significant.
In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists.
The specific Applicable Safety Analyses and LC0 discussions are listed below on a function by function basis.
a.
Reactor Vessel Water level - ATWS-RPT Level l
Low RPV water level indicates the capability to cool the fuel may be threatened.
Should RPV water level decrease too far, fuel damage could result.
4 Therefore, the ATWS-RPT System is initiated at a low level to aid in maintaining level above the top of the active fuel.
The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
l Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
s (continued)
HATCH UNIT 1 B 3.3-91 Amendment No.1%
ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE a.
Reactor Vessel Water Level -- ATWS-RPT Level l
SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of Reactor Vessel Water Level -- ATWS-RPT Level, with two channels in each trip system, are available and required to be OPERABLE to ensure that i
no single instrument failure can preclude an ATWS-RPT from this function on a valid signal.
The Reactor Vessel Water Level -- ATWS-RPT Level Allowable Value is l
chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.
b.
Reactor Steam Dome Pressure -- Hiah Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization.
The Reactor Steam Dome Pressure -- High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation.
For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure -- High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Steam Dome Pressure -- High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal.
The Reactor Steam Dome Pressure -- High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
~
(
(continued)
HATCH UNIT 1 B 3.3-92 Amendment No. 196
l RCS P/T Limits B 3.4.9 BASES i
SURVEILLANCE SR 3.4.9.1 (continued)
REQUIREMENTS cooldown operations and RCS inservice leakage and hydrostatic testing.
SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design j
allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
If the 145*F temperature differential specified in SR 3.4.9.3 cannot be determined by direct indication, an hlternate method may be used as described below:
The bottom head coolant temperature and the RPV coolant can be assumed to be s 145 F if the following can be confirmed:
(
a.
One or more loop drive flows were > 40 percent of rated flow prior to the RPT, (continued)
HATCH UNIT 1 B 3.4-50 knendment No.1%
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued)
REQUIREMENTS b.
High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not injected since the RPT, c.
Feedwater temperature has remained > 300*F since the RPT, and d.
The time between the RPT and restart is < 30 minutes.
General Electric test data from BWR plants shows that stratification up to the 145*F differential does not occur any sooner than I hour following the RPT (Refs.10 and 11).
Adding HPCI and RCIC injection, and feedwater temperature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the RPT.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4.
In MODE 5, the overall stress on limiting components is lower. Therefore, oT limits are not required.
(
(continued)
HATCH UNIT 1 B 3.4-50a Amendment No.196 l
RCS P/T Limits B 3.4.9 This page intentionally left blank.
\\
HATCH UNIT 1 8 3.4-50b Amendment No. 196 l
RCS P/T Limits B 3.4.9 BASES 1
REFERENCES 3.
ASTM E 185-82, " Standard Practice for Conducting (continued)
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.
4.
5.
Regulatory Guide 1.99, Revision 2, May 1988.
6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
7.
FSAR, Section 14.3.6.2.
8.
George W. Rivenbark (NRC) letter to J. T. Beckham, Jr.
(GPC), Amendment 126 to the Operating License, dated June 20, 1986.
9.
NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993, 10.
GE-NE-668-13-0393, " Recirculation Pump Restart Without Vessel Temperature Indication for E.I. Hatch Nuclear Plant," April 9, 1993.
11.
DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant Hazards Consideration Assessment for RPV Stratification Prevention Improvements at Edwin I.
Hatch Nuclear Plant Units 1 and 2," April 1994.
e
\\
HATCH UNIT 1 B 3.4-52 Amendment No. 196
ATWS-RPT Instrumentation 3.3.4.2
~
3.3 INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation LCO 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation function listed below shall be OPERABLE:
a.
Reactor Vessel Water Level -- ATWS-RPT Level; and l
b.
Reactor Steam Dome Pressure -- High.
APPLICABILITY:
MODE 1.
ACTIONS
__________.----------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel.
b CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel to 14 days inoperable.
OPERABLE status.
QB A.2
NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
Place channel in 14 days trip.
v u
(continued)
(
i HATCH UNIT 2 3.3-31 Amendment No. 136
~.
m..
ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST.
92 days SR 3.3.4.2.3 Perform CiiANNEL CALIBRATION. The 18 months Allowable Values shall be:
a.
ATWS-RPT Level: 2 -73 inches; and b.
Reactor Steam Dome Pressure -- High:
s 1095 psig.
SR 3.3.4.2.4 Perfonn LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
\\
HATCH UNIT 2 3.3-33 Amendment No. 136
ATWS-RPT Instrumentation B 3.3.4.2 B 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following ever.Ls in which a scram does not (but should) occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level -- ATWS-RPT l
Level or Reactor Steam Dome Pressure -- High setpoint is l
reached, the recirculation pump drive motor breakers trip.
The ATWS-RPT System (Ref.1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT.
The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.
The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure -- High and two channels of Reactor Vessel Water Level -- ATWS-RPT Level in l
each trip system.
Each ATWS-RPT trip system is a two-out-of-two logic for each Function.
Thus, either two Reactor Water Level -- ATWS-RPT Le'.el or two Reactor l
Pressure -- High signals are needed to trip a trip system.
The outputs of the ch3nnels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (oy tripping the respective drive motor breakers).
There is one drive mo or breaker provided for each of the two recirculation pumps for a total of two breakers.
The o'utput of each trip system is provided to both recirculation pump breakers.
\\
(continued) j HATCH UNIT 2 B 3.3-89 Amendment No. 136 E
ATWS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE The individual Functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1 to protect against common mode failures of the LCO, and Reactor Protection System by providing a diverse trip to APPLICABILITY mitigate the consequences of a postulated ATWS event.
The (continued)
Reactor Steam Dome Pressure -- High and Reactor Vessel Water Level -- ATWS-RPT Level Functions are required to be OPERABLE l
in MODE 1, since the reactor is producing significant power and the recirculation system could be at high flow. During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event.
In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event.
Therefore, the ATWS-RPT is not necessary.
In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a signi.ficant pressure increase or low water level is negligible.
In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant.
In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists.
The specific Applicable Safety Analyses and LC0 discussions are listed below on a Function by Function basis.
a.
Reactor Vessel Water Level -- ATWS-RPT Level l
Low RPV water level indicates the capability to cool the fuel may be threatened.
Should RPV water level decrease too far, fuel damage could result.
Therefore, the ATWS-RPT System is initiated at a low level to aid in maintaining level above the top of the active fuel.
The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
s (continued)
HATCH UNIT 2 8 3.3-91 Amendment No. 136
h ATUS-RPT Instrumentation B 3.3.4.2 BASES APPLICABLE a.
Reactor Vessel Water Level -- ATWS-RPT level l
SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of Reactor Vessel Water Level -- ATWS-RPT Level,with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this function on a valid signal. The Reactor ressei Water Level -- ATWS-RPT Level Allowable Value is l
chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation, b.
Reactor Steam Dome Pressure -- Hiah Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization.
The Reactor Steam Dome Pressure -- High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation.
For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure -- High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Steam Dome Pressure -- High, with two channels in each trip system, are available and are required to be OPERABLE to ensure thtt no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Steam Dome Pressure -- High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
i (continued)
HATCH UNIT 2 B 3.3-92 Amendment No.136
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)
REQUIREMENTS cooldown operations and RCS inservice leakage and hydrostatic testing.
SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temporature must be verified witbin the appropriate limits before withdrawing control rods that will make the reactor cri tical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
4 i
SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup i.
of an idle recirculation loop (Ref. 7) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
If the 145*F temperature differential specified in SR 3.4.9.3 cannot be determined by direct indication, an alternate method may be used as described below:
The bottom head coolant temperature and the RPV coolant can be assumed to be s 145 F if the following can be confirmed:
(
a.
One or more loop drive flows were > 40 percent of rated flow prior to the RPT, (continued)
HATCH UNIT 2 B 3.4-50 Amendment No. 136
4 RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued)
REQUIREMENTS b.
High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not injected since the RPT, c.
Feedwater temperature has remained > 300 F since the RPT, and d.
The time between the RPT and restart is < 30 minutes.
General Electric test data from BWR plants shows that stratification up to the 145 F differential does not occur any sooner than I hour following the RPT (Refs. 10 and 11).
Adding HPCI and RCIC injection, and feedwater temperature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the RPT.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4.
In MODE 5, the overall stress on limiting enmponents is lower. Therefore, oT limits are not required.
i (continued)
HATCH UNIT 2 B 3.4-50a Amendment No. 136 l
RCS P/T Limits B 3.4.9 This page intentionally left blank.
s HATCH UNIT 2 B 3.4-50b Amendment No. 136 l
l
RCS.P/T Limits B 3.4.9 BASES REFERENCES 3.
ASTM E 185-82, " Standard Practice for Conducting i
(continued)
Surveillance Tests for Light-Water Cooled Nuclear i
Power Reactor Vessels," July 1982.
4.
5.
Regulatory Guide 1.99, Revision 2, May 1988.
I 6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
7.
FSAR, Section 15.1.26.
8.
Kahtan N. Jabbour (NRC) letter to W. G. Hairston, III (GPC), Amendment 118 to the Operating License, dated January 10, 1992.
9.
NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
i 10.
GE-NE-668-13-0393, " Recirculation Pump Restart Without Vessel Temperature Indication for E.I. Hatch Nuclear Plant," April 9, 1993.
11.
DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant Hazards Consideration Assessment for RPV Stratification Prevention Improvements at Edwin I.
i Hatch Nuclear Plant Units 1 and 2," April 1994.
j i
i j
i J
l
's l
HATCH UNIT 2 8 3.4-52 Amendment No.136