ML20070M289
| ML20070M289 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/18/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070M292 | List: |
| References | |
| NUDOCS 9404250290 | |
| Download: ML20070M289 (15) | |
Text
-
7 i) #/.
4 es 3
UNITED STATES 5%
>l NUCLEAR REGULATORY COMMISSION
\\..s e,/
WASHINGTON, o.C. 20555-0001 PUBLIC SERVICE ELECTRIC & GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY ILOCKET NO. 50-354 HOPE CREEK GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 69 License No. NPF-57 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Public Service Electric &
Gas Company (PSE&G) dated March 4, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations 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 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 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by 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-57 is hereby amended to read as follows:
(2)
Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 69 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license.
PSE&G shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
9404250290 940418 PDR ADOCK 05000354 p
~~
.: 3.
The license amendment is effective as of its date of issuance and shall be implemented within 60 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Charles L. Miller, Director Project Directorate I-2 Division of Reactor _ Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 18, 1994 s
i
+
ATTACHMENT TO LICENSE AMENDMENT N0. 69 FACILITY OPERATING LICENSE N0. NPF-57 DOCKET N0. 50-354 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages provided to maintain document completeness.*
Remove Insert xv xv xvi xvi*
xxi xxi xxii xxii
- 1-11 1-11 j
3/4 10-7 3/4 10-7*
3/4 10-8 8 3/4 5-1 B 3/4 5-l*
B 3/4 5-2 B 3/4 5-2 B 3/4 5-3 B 3/4 10-1 B 3/4 10-l*
B 3/4 10-2
I 4
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1........................................
3/4 9-17 1
Low Water Leve1.........................................
3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 1
1 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY...........................
3/4 10-1 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM.............................
3/4 10-2 i
1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS..........................
3/4 10-3 l
3/4,10.4 RECIRCULATION LOOPS.....................................
3/4 10-4 3/4.10.5 OXYGEN CONCENTRATION....................................
3/4 10-5 3/4.10.6 TRAINING STARTUPS.......................................
3/4 10-6 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING..........
3/4 10-7 3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING.................
3/4 10-8 l 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration...........................................
3/4 11-1 Table 4.11.1.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program...
3/4 11-2 Dose....................................................
3/4 11-5 Liquid Waste Treatment..................................
3/4 11-6 Liquid Holdup Tanks.....................................
3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Dose Rate...............................................
3/4 11-8 Table 4.11.2.1.2-1 Radioactive Gaseous Waste Sampling and Analysis Program...
3/4 11-9 Dose - Noble Gases......................................
3/4 11-12 Dose - Iodine-131, Icdine-133, Tritium and Radionuclides in Particulate Form................
3/4 11-13 Gaseous Radwaste Treatment..............................
3/4 11-14 Ventilation Exhaust Treatment System....................
3/4 11-15 HOPE CREEA xv Amendment No. 69 l
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Explosive Gas Mixture.....................................
3/4 11-16 Main Condenser............................................
3/4 11-17 Venting or Purging........................................
3/4 11-18 3/4.11.3 SOLID RADIOACTIVE WASTE TREATMENT.........................
3/4 11-19 3/4.11.4 TOTAL 00SE................................................
3/4 11-20 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...........................
3/4 12-1 Table 3.12.1-1 Radiological Environmental Monitoring Program...................
3/4 12-3 Table 3.12.1-2 Reporting Levels For Radioactivity Concentrations In Environmental Samples..............................
3/4 12-9 Table 4.12.1-1 Detection Capabilities For Environmental Sample Analysis........
3/4 12-10 3/4.12.2 LAND USE CENSUS...........................................
3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM........................
3/4 12-14 HOPE CREEK xvi
i 1.
INDEX l
BASES l
EE9 TION Ehg1
)
I.
l l
3/4.10 QEggIAL TEST EXCEPTIONS l
3/4.10.1 PRIMARY CONTAINMENT INTEGRITY....................
B 3/4 10-1 t
l 3/4.10.2 ROD SEQUENCE CONTROL SYSTEM......................
B 3/4 10-1 l
l 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS...................
B 3/4 10-1 1
3/4.10.4 RECIRCULATION LOOPS..............................
B 3/4 10-1 3/4.10.5 OXYGEN CONCENTRATION.............................
B 3/4 10-1 3/4.10.6 TRAINING STARTUPS................................
B 3/4 10-1 3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING...
B 3/4 10-1 4
3/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING...........
B 3/4 10-2 l l
J/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS 1
Concentration.........................................
B 3/4 11-1 Dose..................................................
B 3/4 11-1 Liquid Radwaste Treatment System......................
B 3/4 11-2 Liquid Holdup Tanks...................................
B 3/4 11-2 3/4.11.2 GASEOUS EFFLUENTS Dose Rate.............................................
B 3/4 11-2 Dose - Noble Gases....................................
B 3/4 11-3 Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form...................
B 3/4 11-3 Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment Systems...............
B 3/4 11-4 Explosive Gas Mixture.................................
B 3/4 11-4 Main Condenser........................................
B 3/4 11-5 Venting or Purging....................................
B 3/4 11-5 3/4.11.3 SOLID RADIOACTIVE WASTE TREATMENT.....................
B 3/4 11-5 3/4.11.4 TOTAL DOSE............................................
B 3/4 11-5 HOPE CREEK xxi Amendment No. 69 l
INDEX BASES SECTION PAGE 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM........................................
B 3/4 12-1 3/4.12.2 LAND USE CENSUS...........................................
B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM........................
B 3/4 12-2 HOPE CREEK xxii
TABLE 1.2 OPERATIONAL CONDITIONS HODE SWITCH AVERAGE REACTOR CONDITION POSITION COOLANT TEMPERATURE 1.
POWER OPERATION Run Any temperature 2.
STARTUP Startup/ Hot Standby Any temperature 3.
HOT SHUTDOWN Shutdown $,...
> 200*F 4.
COLD SHUTDOWN Shutdown #'##'***
l s 200*F 5.
REFUELING Shutdown or Refuel s 140*F The reactor mode switch may be placed in the Run, Startup/ Hot Standby, or Refuel position to test the switch interlock functions and related instrumentation provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
If the reactor mode switch is placed in the Refuel position, the one-rod-out interlock shall be OPERABLE.
- The reactor mode switch may be placed in the Refuel position while a single control rod drive is being removed from the reactor pressure vessel per specification 3.9.10.1.
.Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
See Special Test Exceptions 3.10.1 and 3.10.3.
- . The reactor modo switch may be placed in the Refuel position while a single control rod is being recoupled or withdrawn provided that the one-rod-out interlock is OPERABLE.
+
s See Special Test Exception 3.10.8.
I HOPE CREEK 1-11 Amendment No.
69 l
l
4 SPECIAL TEST EXCEPTIONS 3/a.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING LIMITING CON 0!T!0N FOR OPERATION 3/4.10.7 with the issuance of Amendment No.14.The material originally containe reference to this section, Section 3/4.10.7 is intentionally left blankHow fs-1
\\
1 4
i F
'l 1
HOPE CREEK
-3/4 10-7 Amendment No.14 e
- [3 k
Y j
th
SPECIAL TEST EXCEPTIONS 1/4.10.8 INSERVICE LEAK AND HYDROSTATIC TESTING LIMITING CONDITION FOR OPERATION 3.10.8 When conducting inservice leak or hydrostatic testing, the average reactor coolant temperature specified in Table 1.2 for OPERATIONAL CONDITION 4 may be increased to 212'F, and operation considered not to be in OPERATIONAL CONDITION 3, to allow performance of an inservice leak or hydrostatic test provided the following OPERATIONAL CONDITION 3 LCO's are mets a.
3.3.2,
" ISOLATION ACTUATION INSTRUMENTATION", Functions 2.a, 2.b, 2.c, 2.d and 2.e of Table 3.3.2-1; b.
3.6.5.1,
" SECONDARY CONTAINMENT INTEGRITY";
c.
3.6.5.2,
" SECONDARY CONTAINMENT AUTOMATIC ISOLATION DAMPERS"; and d.
3.6.5.3,
" FILTRATION, RECIRCULATION AND VENTILATION SYSTEM."
APPLICABILITY:. OPERATIONAL CONDITION 4, with average reactor coolant temperature > 200*F.
ACTION:
With the requirements of the above specification not satisfied, bnmediately enter the applicable condition of the affected specification or immediately suspend activities that could increase the average reactor coolant temperature or pressure and reduce the average reactor coolant temperature to 6 200*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS
...............u.............................................................
4.10.8 Verify applicable OPERATIONAL CONDITION 3 surveillances for specifications listed in 3.10.8 are met.
HOPE CREEK 3/4 10-8 Amendment No. 69
)
3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUT 00WN The core spray system (CSS), together with the LPCI mode of the RHR system, i
is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS.
The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of eccidental draining.
The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident.
Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for small breaks following depressurization by 3
the ADS.
The surveillance requirements provide adequate assurance that the LPCI pystent will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
The high pressure coolant injection (HPCI) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.
The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.
The HPCI system continues to operate until reactor vessel pressure is below the pressure at which CSS operation or LPCI mode of the RHR system operation maintains core cooling.
The capacity of the system is selected to provide the required core cooling, The HPCI pump is designed to deliver greater than or equal to 5600 gpm at reactor pressures between 1120 and 200 psig.
Initially, water from the condensate storage tank is used instead of injecting water from the supprersion pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.
T m
HOPE CREEK B 3/4 5-1
w EBERGENCY CORE COOLING SYSTEM BASES ECCS-OPERATING and SHUTDOWN (Continued)
With the HPCI system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the CSS and LPCI systems.
In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the safety atilysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCI out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems and the RCIC system.
The surveillance requirements provide adequate assurance that the HPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor to be in HOT SHUTDOWN with vessel pressure not less than 200 psig. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.
Upon failure of the HPCI system to function properly after a small break loss-of-coolant accident, the automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200*F.
ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 100 peig. This pressure is substantially below that for which the low pressure core cooling systems can provide adequate core cooling for events requiring ADS.
ADS automatically controls five selected safety-relief valves although the safety analysis only takes credit for four valves.
It is therefore appropriate to permit one valve to be out-of-service for up to 14 days without materially reducing system reliability.
3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCI, CSS and LPCI systems in the event of a LocA.
This limit on suppression chamber minimum water volmne ensures that suf ficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is also required by Specification 3.6.2.1.
Repair work might require making the suppression chamber inoperable.
This specification will permit those repairs to be made and at the same tLme give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.
HOPE CREEK B 3/4 5-2 Amendment No. 69 l
"i l'
1 l
EMERGENCY CORE COOLING. SYSTEM l
BASES w................................
3/4.5.3 SUPPRESSION CHAMBER (Continued)
In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required' water yolume is reduced because the reactor coolant is maintained at or below 200*F.
Since pressure suppression is not required below 212*F, the minimum l
i water volume is based on NPSH, recirculation volume and vortex prevention plus a safety margin for. conservatism.
1 See Special Test Exception 3.10.8.
l l
.i i
1
'l 1
i t
HOPE CREEK B 3/4 5-3 Amendment No. 69 l
1 3/4.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PMSICS TESTS.
3/4.10.2 R00 SEQUENCE CONTROL SYSTEM In order to perform the tests required in the technical specifications it is necessary to bypass the scquence restraints on control rod mover,ent.
The additional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.
3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations during open vessel testing requires additional restrictions in order to ensure that criticality is properly monitored and controlled.
These additional restrictions are specified in this LCO.
+
3/4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow c6nditions and is required to perform certain PHYSICS TESTS while at low THERMAL l
POWER levels.
3/4.10.5 OXYGEN CONCENTRATION The material originally contained in this Technical Specification was deleted with the issuance of Amendment No. 35 However, to maintain the historical reference to this specification, this section has been intentionally left blank.
3/4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one flHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.
3/4.10.7 SPECIAL INSTRUMENTATION - INITIAL CORE LOADING The material originally contained in Bases Section 3/4.10.7 was deleted with the issuance of Amendment No. 14.
However, to maintain the historical reference to this section, Bases Section 3/4.10.7 is intentionally left blank.
e HOPE CREEK B 3/4 10-1 Amendment No. 35 DEC 181989
(
w.
^
l 3/4.10 SPECIAL TEST EXCEPTIONS BASES
.....................u........................................................
3/4.10.8 INSERVICE' LEAK AND HYDROSTATIC TESTING This special te st exception allows reactor vessel inservice ' leak and -
a hydrostatic testing to be performed in OPERATIONAL CONDITION 4 with reactor coolant temperaturis $l 212*F.
The additionally imposed OPERATIONAL CONDITION-s 3 requirement fn SECONDARY CONTAINMENT operability provides conservatism in the response of the unit to an operational event. This allows _ flexibility since temperatures approach 200'F during the testing and can drift higher because of decay and mechanical heat, f
i i
i 3
HOPE CREEK B 3/4 10-2 Amendment No. 69
.c
...