ML19312D711
| ML19312D711 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 02/21/1980 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19312D712 | List: |
| References | |
| NUDOCS 8003250228 | |
| Download: ML19312D711 (24) | |
Text
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.s fi fg UNITED STATES y $ eing g NUCLEAR REGULATORY COMMISSION
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N .Cp f PUBLIC SERVICE ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 24 License No. DPR-70 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Public Service Electric and Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated June 29, 1978 as supplemented by letter dated September 27, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Ac!) 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 Comission; 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; 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.
80 032 501N
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. 2.
Acconfingly, 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. OPR-70 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 24 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 the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION M b2LWNd-A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: February 21, 1980 9
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.o ATTACHMENT TO LICENSE AMENDMENT NO. 24 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET NO. 50-272 i
s Revise Appendix A as follows:
Remove Pages Insett Pages v
V x-X 3/4 4-3 3/4.4-3 3/4 4-30 3/4 4-30 1
2 3/4 4-31
- 3/4 4-31, 4-32 and 4-33 3/4 5-6 3/4 5-6 and 5-6a B3/4 4-1 B3/4 4-1 and la B3/4 4-12 B3/4 4-10
.B3/4 5-1 B3/4 5-1 G
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.s INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Co n ta i nment I n teg ri ty..................................
3/4 6-1 Co n ta i nme n t L e a ka g e....................................
3/4 6-2
' Co n ta i nmen t Ai r Loc ks..................................
3/4 6-i I n te rnal P res s u re......................................
3/4 6-5 A i r Tempe ra tu re........................................
3/4 6 Contai nment Structural Integri ty.......................
3/4 6-3 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System...............................
3/4 6-9 3/4 6-10 Spray Addi tive Sy s tem..................................
Conta i nment Cool i ng System.............................
3/4 6-11 3/4.6.3 CONTAINMENT ISOLATION VALVES...........................
3/4 6-12 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.....................................
3/4 6-18 Electric Hydrogen Recombiners..........................
3/4 6-19 l
3/4.7 PLANT SYSTEMS l
l 3/4.7.1 TURBINE CYCLE l'
3/4 7 1 Safety Valves..........................................
l Auxil i a ry Feedwa te r System.............................
3/47-5 l
Auxil iary Feed S tora ge Tank............................
3/4 7-7 3/4 7.B Activity...............................................
Main Steam Line Isol ation Valves.......................
3/4 7-10 S econdary Wa ter C hemi s try..............................
3/4 7-11 SALEM - UNIT 1 VI
a i
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Page 3/4.4.7.
CHEMISTRY..............................................
3/4 4-17 3 /4.4.8 SPECIFIC ACTIVITY......................................
3/4 4-20 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Rea ctor Coolant Sys tem.................................
3/4 4-24 P re s s u ri z e r............................................
3/4 4-29 Overpressure Protection Systems........................
3/4 4-30
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3/4.4.10 STRUCTURAL INTEGRITY ASME Code Class 1, 2 and 3 Components..................
3/4 4-32 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) l 3/4. 5.1 AC C W4U L AT ORS...........................................
3/45-1 avg - 350*F.........................
3/45-3 3/4.5.2 ECCS SUBSYSTEMS - T 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350* F.........................
3/45-5
~
3/4.5.4 BORON INJECTION SYSTEM Bo ron Inj ecti on Tan k...................................
3/45-7 Heat Tracing...........................................
3/45-8 3/4.5.5 REFUELING WATER STORAGE TANK...........................
3/45-9 1
)
l SALEM - UNIT 1 V
Amendment No. 24
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i eo INDEX BASES ~
SECTION Page 3/4.0 APPLICABILITY..........................................
B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.....................................
B 3/4 1-1 3/4.1.2 BORATION SYSTEMS.....................................
G 3/4 1-3 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...........................
B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE................................
B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACT 0RS......................................
B 3/4 ?-4 3/4.2.4 QUADRANT POWER TILT RATI0............................
B 3/4 2-5 3/4.2.5 D N B PA RAM ET E RS.......................................
B.3/4 2-6 l
l l
SALEM - UNIT 1 IX Amendment No. 16
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INDEX BASES 1
SECTION P,]fE, 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............................
B 3/4 3-1 3/4.3.2 ENGINEEREO SAFETY FEATURES (ESF) INSTRUMENTATION........
B 3/4 3-1 3/4. 3. 3 ' MONITORING INSTRUMENTATI ON..............................
B 3/4 3-1
)
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS...................................
B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES...............................
B 3/4 4-la 3/4.4.4 PRESSURIZER.............................................
B 3/4 4-2 3/4.4.5 STEAM GENERATORS........................................
B 3/4 4-2 3/4.4.6 REACTOR COOLANT SY STEM L EAKAGE..........................
B 3/4 4-3 3/4.4.7 CHEMISTRY...............................................
B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY.......................................
B 3/4 4-5 3/4.4.9 PRESSURE /TEMPE RATURE LIMITS.............................
B 3/4 4-6 3 / 4. 4.10 STRU CTU RAL I NT EGR I TY...................................
B 3/4 4-12 SALEM - UNIT 1 X
Amendment No.
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REACTOR COOLANT SYSTEM ACTION (Continued)
Below P-7#:
With K
> 1.0, operation may proceed provided at least two a.
f reactof [oolant loops and associated pumps are in operation.
b.
With X
< l.0, operation may proceed provided at least one reactof [oolant loop is in operation with an associated reactor f
coolant or residual heat removal pump.*
I c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REOUIREMENTS 4.4.1.1 With one reactor coolant loop and associated _ pump not in opera-tion, at least once per 31 days determine that:
a.
The applicable reactor trip system and/or ESF actuation system instrumentation channels specified in the ACTION statements above have been placed in their tripped conditions, and b.
If the P-8 interlock setpoint has been reset for 3 loop opera-tion, its setpoint is < 76", of RATED THERMAL POWER.
All reactor coolant pumps and residual heat removal pumps may be de-l energized for up to I hour, provided no operations are permitted which could cause dilution of the reactor coolant system baron concentration.
'A reactor coolant pum;: shall not be started with one or more of the RCS cold leg temperatures less than 312'F unless 1) the pr>ssurizer water volume is less than 1630 cubic feet (equivalent to approximately 92%
of level) or 2) the secondary water temperature of each steam generatcr is l
less than 50 F above each of the RCS cold leg temperatures.
SALEM - UNIT 1 3/4 4-3 Amenoment No. 2<
s REACTOR COOLANT SYSTEM SAFETY ' VALVES - SHUTDOWN LIMITING CONDITION FOR OPERATION -
3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift. setting of 2485 PSIG + 1%.
APPLICABILITY: MODES 4 and 5.
ACTIOF,:
With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.
SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those requirad by Specification 4.0.5.
l SALEM - UNIT 1 3/4 4 4 1
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.s.
3 REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:
a.
A maximum heatup of 100'F in any one hour period, b.
A maximum cooldown of 200'F in any one hour period, and c.
A maximum spray water tempgrature differential of 320*F.
APPLICABILITY: At all times.
ACTION:
With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pres-surizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the fol-lowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
4 SURVEILLANCE REQUIREMENTS 4.4.9.2.The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.
The spray water tenperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.
SALEM - UNIT 1 3/4 4-29 Amendment No. - 9
w L
REACTOR COOLANT SY5 TEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3d.9.3 ' At least one of the following overpressure protection systems L
shall be OPERABLE:
i I
l Two Pressurizer Overpressure Protection System relief valves a.
(P0Ps) with a lift setting of less than or equal to 375 psig, or
)
b.
A reactor coolant system vent of greater than or equal to 3.14 square-inches.
l APiLICABILITY: When the temperature of one or.more the RCS cold legs is less than or equal to 312*F, except when the reactor vessel head is removed.
i.
ACTION:
With one POPS inoperable, either restore the inoperable POPS to a.
OPERABLE status within 7 days or depressurize and vent the RCS through a 3.14 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both POPS have been restored to OPERABLE status.
l b.
With both POPS inoperable, depressurize and vent the RCS through-a 3.14 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented conditin until both POPS have been restored to OPERABLE
- status, In the event either the POPS or the RCS vent (s) are used to c.
mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Comission pursuant to Specif-ication 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the POPS or vent (s) on. the transient and any corrective action necessary to prevent recurrence.
The provigions of Specification 3.0.4 are not applicable.
d.
SURVEILLANCE REQUIREMENTS 0.4.9.3.1. Each POPS shall be demonstrated OPERABLE by:
SALEM - UNIT 1 3/4 4-30 Amendment No. 24 m
.s REACTOR COOLANT SYSTEM SUP.VEILLANCE REQUIREMENTS (Continued) a.
Performance of a CHANNEL FUNCTIONAL TEST on the POPS actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the POPS is required OPERABLE.
b.
Performance of a CHANNEL CALIBRATION on the POPS actuation channel at least once per 18 months.
c.
Verifying the POPS isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the POPS is being used for overpressure protection.
l d.
Testing in accordance with the inservice test requirements for ASME-Category C valves pursuant to Specification 4.0.5.
4.4.9.3.2 The RCS vent (s) shall be verified to be open 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 the vents (s) is being used for overpressure protection.
6 b
0
- Except wnen the vent pathway is provided with a valve which i; locked, sealed, or otherwise secured in the open position,.then verif./ these valves open at least once per 31 days.
l SALEM - UNIT 1 3/4 4-31 Amendment io. 24 l
REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME _ Code Class 1, 2 and 3 com-ponents shall be maintained in accordance with Specification 4.4.10.1.
APPLICABILITY: ALL MODES ACTION:
a.
With the structural integrity of any ASME Code Class 1 com-ponent(s) not confonning to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.
b.
With the structural integrity of any ASME Code Class 2 com-ponent(s) not confonning to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature abo /e 200*F.
c.
With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.
d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.10.1.1 The structural integrity of ASME Code Class 1, 2 and 3 components shall be demonstrated:
a.
Per the requirements of Specification 4.0.5, and b.
Per the requirements of the augmented inservice inspection program specified in Specification 4.4.10.1.2.
SALEM - UNIT 1 3/4 4-32 Amendment No. 24
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)
In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14 Revision 1, August
'1975.
4.4.10.1.2 Augmented Inservice Inspection Program for Steam Generator Channel Heads _ - The steam generator channel heads shall be ultrasonic inspected during each of the first three refueling outages using the saae ultrasonic inspection procedures and equipment used to generate the base-line data. These inservice ultrasonic inspections shall verify that the cracks observed in the stainless steel cladding prior to operation have not propagated into the base material. The stainless steel clad surfaces of the steat generator channel heads shall also be 100% visually inspected during -he above outages and a television camera shall be used to make i videota:e recording of the condition of this cladding. Each videotape shall be cocpared with those obtained during the previous outages to determine that the cladding does not show any abnormal degradation.
SALE *. - UNIT 1 3/4 4-33 Amendment No. 24
'^^ '*
0; EMERGENCY CORE COOLING' SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
By a visual inspection which verifies that no loose debris c.
(rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions.
This visual inspection shall be performed:
1.
For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and 2.
Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.
d.
At least once per 18 months by:
1.
Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is above 580 psig.
2.
A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens.,
etc.) show no evidence of structural distress or corrosion.
e.
At least once per 18 months, during shutdown, by:
1.
Verifying that each automatic valve in the flow path actuates to its correct position on a safety injection test signa' 2.
Verifying that each of the following pumps start autor.atically
~
upon receipt of a safety injection tcst signal:
'a )
Centrifugal charging pump b)
Safety injection pump c)
Residual heat removal pump l
L SALEM - UNIT 1 3/4 E-5
EMERGDCY CORE COOLING SYSTEMS n
ECCS SL3 SYSTEMS - T,y, < 350*F
~"
LIMITIN3 CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem 3 comprised of the following shall be OPEP.1BLE:
a.
One OPERABLE centrifugal charging pump, l
b.
One OPERABLE residual heat removal heat exchanger, c.
One OPERABLE residual heat removal ptmp, and d.
An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recir-culation phase of operation.
AP:LICA3ILITY: MODE 4.
AC ION:
a.
With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the l
refueling water storage tank, restore at least one ECCS subsystem l
to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHlITDOWN within tne next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
b.
With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at leasz one ECCS subsystem to OPERABLE status or maintain the Reactar Coolant System T*V9 less than 350*F by use of alternate heat removal methods.
c.
In the event the ECCS is actuated and injects water into the i
Reactor Coolant System, a Special Report shall be prepared and subctitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.
-i
=A..a.x mc of one safety injection pump shall be OPERABLE whenever the tercerature of one or more of the RCS. cold legs is less than or equal I
t: 312'F.
S/J.EF - l NIT 1 3/4 5-6 A:1endment No. 24
EMERCENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T
< 350*F ava I
SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the appli:able Surveillance Requirements of 4.5.2.
4.5.3.2 All safety injection pumps, except the OPERABLE pump allowed above, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> wr.enever the temperature of one or more of the RCS cold legs is less than or equal to 312*F by verifying that the motor circuit breakers have been removed from their electrical power supply circuits.
SALEM - UNIT 1 3/4 5-6a Amendment f.o.24 G
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m 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. With one reactor coolant loop not in operation, THERMAL POWER is restricted to < 36 percent of RATED THERMAL POWER until the Overtemperature AT trip is reset. Either action ensures that the DilBR will be maintained above 1.30.
A loss of flow in two loops will cause a reactor trip if operating above P-7 (11 percent of RATED THERMAL POWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (36 percent of RATED THERMAL POWER).
A single reactor coolant loop provides sufficient heat removal capability for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing a RHR loop into operation in the shutdown cooling mode if component -repairs and/or corrective actions cannot be made within the allowable out-of-service time, i
The operation of one Reactor Coolant Pump or one RHR pump provides adecuate flow to ensu.e mixing, prevent stratification and produce gradual reactivity changes during boren concentration reductions in the Reactor C:olant System. The reactivity change rate associated with boron reduc-tions will, therefore, be within the capability of operator recognition and centrol.
The restrictions on starting a Reactor Coolant Pump below P-7 with one or rore RCS cold legs less than or equal to 312 F are provided to prevent RCS pressure transients, :aused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water vclume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting from the RCPs to wten the secondary water temperature of each steam generator is less thin 50*F above each of the RCS cold leg temperatures.
SALEM - UNIT 1 B 3/4 4-1 Amendment No 24 g
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p 3/4.4 REACTOR COOLANT SYSTEM i
SASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES
'The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000.lbs per hour of saturated steam at the valve
(
set point. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
l During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 l
psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached'(i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
l l
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SALEM - UNIT I B 3/4 4-la Amendment No. 24 I
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REACTOR COOLANT" SYSTEM l
BASES 3/4.4.4. PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and.is capable of accommodating pressure.
surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including ~the desigr4 step load decrease with steam dump. -Operation of the power operated relief valves minimizes the undesirable opening of the spring-loaded pressurizer code' safety valves.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the s. team generator tubes ensure that the structural integrity of this portion ?f the RCS will be maintained. The program for inservice inspection of steam generator tubes -is based on a modification of Regulatory Guide 1.83 Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to 4
design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant. is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found i to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within these limits, localized corrosion may.likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage =
500 gallons per day per steam generator). Cracks having a primary-to-secondary-leakage less than this ' limit during operation will have an adequate margin of safety to withstand the loads imposed during nornal operation and by postulated accidents. Operating plants have demonstrat-ed that primary-to-secondary leakage of 500 gallons per day'per steac generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown
~
and an unscheduled inspection, curing which. the leaking tubes will be located and plugged.
8 SALEM
. UNIT-'1 B 3/4 4-2 t
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TABL,E B 3/4.4-1 (Continued)'
2> -
{j REACTOR VESSEL TOUGilNESS-Ei 50 FT-LB/35 MIN. UPPER SHELF 1
COMP MATERIAL CU P
NDTT HIL TEMP F RTNDT FT-LB COMPONENT CODE TYPE F
LONG TRANS F
LONG TRKNT LOWER SifL.
10DW A533B1 19 011
-40 45 77*
17
.138 90**
LOWER Sill.
100X A53381 19 012
-70 58 89*
29 124 81**
LOWER Sill.
10DY A533B1 19 010
-40 46 93*
33 124 81**
BOT.llD.SEG 120Z A533B1 10 009 10 28 71*
11 117 76**-
BOT.llD.SEG 12EA A53381 11 010
-50 40 76*
16 131 85**
BOT.ilD.SEG 12EB A53381 12 008 10 27 64*
10 118 76**
BOT.llD. DOM 13EC A53301 15 010
-20 37 84*
24 104 68**
oo
NA
-38 0
NA 97 um HAZ CORE 15ED HAZ NA NA NA
-28 NA NA 107 NA 2
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ESTIMATED (0 F OR 30FT-LB TEMP, WilICHEVER IS LESS)
ESTIMATED (77FT-LB/54 MIL TEMP FOR LONGITUDINAL DATA)
ESTIMATED (65 PER CENT OF LONGITUDINAL SilELF) 1
I o.EACTOR COOLANT SYSTEM BASES The OPERABILITY of two POPSs or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more the RCS cold legs are less than or equal to 312*F.
Either POPS has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to eithe.- (1) the start of an idle ACP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures, or (2) the start of a safety injection pump and its injection into a water solid RCS.
3/4.4.10 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 comaonents ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent apolical e, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.
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S*LEM - UNIT 1 33/44 Amendment No. 24 e
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3/4.5 EMERGENCY CORE COOLING ~ SYSTEMS' F
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L BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each RCS accumulator ensures that a sufficient volume of t orated water will be insnediately forced into the reactor core hrough each oJ the cold. legs in the event the RCS pressure falls below the pressure of -
the accanulators.
This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.
The limits' on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safe;y analysis are met.
The accanulator powe'r operated isolation valves are considered to be
" operating by9 asses" in the context of.IEEE Std. 279-1971, which requires that by: asses of a protective function be. removed automatically whenever pemissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, _ removal of power to the valves is required.
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The limits for operation with an accumulator inoperable for any reason except.an isolation valve closed minimizes the time exposure of the plant to a 1.0% event occurring concurrent with failure of an additional accumulttor which may result in unacceptable peak cladding temperatures.
If a cl: sed isolation' valve cannot be insnediately opened, the full capability of one accumulator is not available and prompt action is recui-ed to place the reactor in a mode where this capability is not recaired.
3 /4.5.2 'and 3/4.5.3 ECCS SUBSYSTEMS The 0?ERASILITY of two independent ECCS subsystems ensures that sufficient emergen:y core cooling capability will be available in the event of a LOCA assumin; the loss of one subsystem through any single failure consideration.
Either subsystem operating in conjunction with the accumulators i~s capable of 'su:piyinc sufficient core cooling to limit the peak cladding temperatures wit,in accentable limits for all postul-ated break sizes ranging from tFe double enced break of the-largest' RCS cold leg pipe downward.
In-addition, eac, EC:S subsystem provides long term _ core cooling capability in the recircu a:icn mode during the accident recovery period.
Tne lin tation for a s.tximum of one safety injection pump to be OPERABLE d
and *9e Surveillance. i.equirement to verify all safety injection pumps except -he allowed OPERABLE pump to be inoperable below 312 F provides assuran:e tcat a mass addition pressure transient can be relieved by the ose a:icn o'f a single POPS relief valve.
SALEM '- UNIT.1 B 3/4 5-1 Amendment No. 24 i
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EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued)
With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single fai'ure consideration on the basis of the stable reactivity condition of the raactor and the limited core cooling requirements.
The Surveillance Requirements provided to ensure OPERABILITY of each ccm-ponent ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintais.ed.
3/4.5.4 BORON INJECTION SYSTEM The OPERABILITY of the boron injection system as part of the ECCS ensur es that sufficient negative reactivity is injected into the core to counter-l act any positive increase in reactivity caused by RCS system cooldown.
RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture.
The limits en injection tank minimum contained volume and boron concen-tration ensure that the assumptions used in the steam line break analysis are met. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
The OPERABILITY.of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will be maintained above the solubility limit of 135'F at 21000 ppm boron.
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3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event l
of a LOCA.
The limits on RWST minimum volume and boron concentration ensure i
that 1) sufficient water is available within containment to permit recir-culation cooling flow to the core, and 2) the reactor will remain subcritical L
in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly.
These assumptions are consistent with the LOCA analyses.
Tne contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
SALEM - UNIT 1 B 3/4 5-2 e
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