ML18092B563
| ML18092B563 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 05/05/1987 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18092B562 | List: |
| References | |
| NUDOCS 8705110288 | |
| Download: ML18092B563 (24) | |
Text
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ATTACHMENT A.
UNIT 1 EXISTING SPECIFICATION
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TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE. TIMES FUNCTIONAL UNIT RESPONSE TIME
- 1.
Manual Reactor Trip NOT APPLICABLE
- 2.
Power Range. Neutron flux
< 0.5 seconds*
- 3.
Power Range. Neutron flux, High Positive Rate NOT APPLICABLE
- 4.
Power Range. Neutron flux.
High Negattv~ Rate
~ 0.5 seconds*
- 5.
Intermediate Range. Neutron flux NOT APPLICABLE
- 6.
Source Range. Neutron flux NOT APPLICABLE
- 7.
Overtemperature AT
< 4.0 seconds*
- 8.
Overpower AT NO! APPLICABLE
- 9.
Pressurizer Pressure--Low
< 2.0 seconds
- 10. Pressurizer Pressure--High
< 2.0 seconds l l. Pressurizer W~ter Level--High NOT. APPLICABLE
- Neutron detectors* are exempt from response time testing. Response time of the neutron flux si9nal portion of the channel shall be measured from detector output or input of first electronic component in clldunel.
TABLE 3.3-5 (Continued)
EMGINE£RED* SAFETY F[ATIJRES RESPONSE T!MES INITlATING SIGNAL AND FUNCTION
- 6.
Steam Flow in Two Steam Lines-Hi h oinc:ident with team Llne Pressure-Low
- a.
Safety Injection (ECCS)
- b.
Reactor Trip (from SI)
- c.
Feedwater Isolation
- d.
Containment Isolation-Phase "A"
- e.
Containment Ventilation Isolation
- f.
Auxiliary Feedwater Pum;is
- g.
Service Water System
- h.
Steam Line Isolation
- 7.
Containment ?ressure--Hign-High
- a.
- b.
Containment Isolation-Phase "8"
- c.
Steam Line Isolation
- d.
Containment Fan Cooler
- a.
Steam Generator Water Leve1--High-Hign RESPONSE T!ME IN SECOr~os
!.. 12~0#/22.0#~
!.. 2.0
< 7. 0
!.. f?.0#/27.0##
Not A~glicable Not Applicable
!. 14.0'-/48.0!'*
!.. a.a
< 45.0 Not Apclicab1e
- c 7. 0
< 40. 0
- a.
Turbine Trip-Reactor Trip
~ 2.5
- b.
Feedwater Isolation
- 9.
Steam Generator Water Level **L~Lo~
- 1.
Motor-Qriven Auxiliary Feit~ater P~s
- b.
Turt>ine*Oriven "uxilial""j FH~ater Pumps.
- SALEM - UNIT 1 3/4 3-29
< 1l. 0
~ 60.0
- ~ 60. a Amendment No. 39
ATTACHMENT B UNIT 2 EXISTING SPECIFlCATION
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w TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME
- 1.
Manual Reactor Trip NOT APPLICABLE
- 2.
Power Range,.Neutron flux
~ 0.5 seconds*
- 3.
Power Range, Neutron flux, High Positiye Rate NOT APPLICABLE
- 4.
Power Range,,Neutron flux, High Negative* Rate
~ 0.5 seconds*
- 5.
Intermediate Range, Neutron flux NOT APPLICABLE
- 6.
Source Range, Neutron flux*
NOT APPLICABLE
- 7.
Overtemperature AT
~ 4
- 0 seconds*
- 8.
Overpower AT NOT APPLICABLE
- 9.
Pressurizer Pressure--Low
~ 2.0 seconds IO.
Pressurizer Pressure--High
~ 2.0 seconds
- 11.
Pressurizer*w~t~r Level--High NOT APPLICABLE fiNeutron detectors are exempt from response lime testing.
RespQnse time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic componEml in channel.
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TABLE 3.3*5 (Cont1nued)
ENGINEERED SAFETY FEATURES RESPONSE TrMES INITIATING SIGNAL ANO FUNCTION RESPONSE TIME IN SECONDS
- 3.
Pressurizer Pressure-Low
- a.
Safety Injection (ECCS)
- b.
Reactor Trip (from SI)
- c.
Feeci'.tater Isolation
- d.
Containment Isolation-Phase 11A11
- e.
Containment Venti1ation Isolation
- f.
Auxiliary Feedwater Pumps
- g.
Service Water System
- 4.
Oif~erential Pr-essure Between Steam Lines*Hign
- a.
Safety Injection (ECCS)
- b.
Reactor Trip (from SI)
- c.
Feedwater Isolation
- d.
Containment Isolation-Phase 11A11
- e.
CJntainment Ventilation Isolation
- f.
Auxiliary Feedwater Pumps
- g.
Service Water System
- 5.
Steam F 1 aw in Two Steam Lfnes -
Hfg~ Coi~c~~e~t
\\
- a.
Saf~ty !njection (ECCS)
- b.
.~eac<:.or Trip (from SI)
- c.
Feedwater Isolation
- d.
Cuntainment rsulation-Pliase 11A11
- e.
Containment Ventilation !solatfon
- f.
Auxiliary Feedwater Pumps
- g.
Ser*lice 'r'later System
- h.
Steam Line Isolation SALE:!-i - UNIT 2 3/4 3-29 S Z7.o(l)/12.oCZ) s z.o s 7.0 s 1s.0CZJ Not App 1 i cab 1 e
( 60
~ 49.oCl)/l3.oC2J s iz.oC2)122.oC 3) s 2.0 s 7.0
~ 17.oC2)/27.0(J)
Not Applicable
< 60
~ 13.oC2)/48.oC 3)
~ l~.JC 2 )12~.0C 3 J
- 4. 0
- 9. 0
-~ l9.o( 2)/29.0C 3)
.~at App1:c3.c;e
< 60
(.,,
fl'
~ :~.o\\~,1~9.o\\~)
~ 3.0
ATTACHMENT C UNIT 1 PROPOSED SPECIFICATION
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1-3 TAB lE 3.3-2 REACTOR TRIP SYSTEM INSTRLMENTATION RESPONSE ITEMS F UNCT IONA L UN IT
- 2.
Power Range, Neutron Flux
- 3.
Power Range, Neutron Flux, High Positive Rate
- 4. Power Range, Neutron Flux, High Negative Rate
- 5.
Intermediate Range, Neutron Flux
- 6. Source Range, Neutron Flux
- 7.
Ove rtempe ratu re 6. T
- 8. Overpower 6 T
- 9.
Pressurizer Pressure--Low
- 10. Pressurizer Pressure--High
- 11. Pressurizer Water Level--High RESPONSE TIME NOT APP LICAB LE
< 0.5 seconds*
NOT APP LICAB LE
< 0.5 seconds*
NOT APP LI CAB LE NOT APP LI CAB lE
< 5. 75 seconds*
NOT APP LICAB LE
< 2.0 seconds
< 2.0 seconds NOT APP LICAB lE
- Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
.I
TABIE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE ITEMS INITIATING SIGNAL AND FUNCTION
- 1. Manual
- a. Safety Injection (ECCS)
Feedwater Isolation Rea ct or Tri p ( S I )
Containment Isolation-Phase "A" Containment Ventilation Isolation Auxiliary Feedwater Pumps Service Water System Containment Fan Cooler
- b. Containment _Spray Containment Isolation-Phase "B 11 Containment Ventilation Isolation
- c. Containment Isolation-Phase 11A11 Containment Ventilation Isolation
- d.
Steam Line Isolation
- 2. Containment Pressure-High
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Isolation-Phase "A11
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System SA LEM -
UN IT 1 3/4 3-27 RESPONSE TIME IN SECON a; Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable
<27.o<l)
< 2.0
< 7.0
- 2. 17.0(2)/27.0(3)
Not Applicable
< 60
- 2. 13.o(2)/4S.o(3)
I TAB LE 3. 3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE ITEMS INITIATING SIGNAL AND FUNCTION
- 3. Pressurizer Pressure-Low
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. F eedwater Isolation
- d. Containment Isolation-Phase "A"
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System
- 4.
Differential Pressure Between Steam lines-High
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Isolation-Phase 11A 11
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System
- 5. Steam Flow in two Steam Lines - High Coincident with T avg --Low-Low*
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Isolation-Phase "A"
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater. Pumps
- g. Service Water System
- h. Steam Line I sol at ion SA LEM -
UN IT 1 3/4 3-28 RESPONSE TIME IN SECON(l)
- 2. 27.0{l)/12.0(2)
< 2.0
< 7.0
< 18.0(2)
Not Applicable
< 60
< 49.o(l)/13.0(2)
- 2. 12.0(2)122.0(3)
< 2.0
< 7.0
- 2. 17.0(2)/27.0(3)
Not Applicable
< 60
- 2. 13.o(2)/48.o(3)
- 2. 15.75(2)/25.75{3)
< 5.75
< 10. 75
- 2. 20.75(2)/30.75(3)
Not Applicable
< 61.75
- 2. 15.75(2)/50.75(3)
< 10. 75
- TAB LE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION
- 6. Steam fl ow in Two Steam Lines-High Coincident with Steam Line Pressure-Low
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Isolation-Phase 11A 11
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System
- h. Steam Line Isolation
- 7. Containment Pressure--High-High
- b. Containment Isolation-Phase 11B 11
- c. Steam Line Isolation
- d. Containment Fan Cooler
- 8. Steam Generator Water Level--High-High
- a. Turbine Trip
- b. Feedwater Isolation
- 9. Steam Generator Water Level--Low-Low
- a. Motor-Driven Auxiliary Feedwater Pumps(4)
- b. Turbine-Driven Auxiliary Feedwater Pumps{5)
SA LEM -
UN IT 1 3/4 3-29 RESPONSE TIME IN SECONDS
- 2. 12.0(2)/22.0(3)
< 2.0
< 7.0
- 2. 11.0(2)121.0{3).
Not Applicable
< 60
- 2. 14.0(2)/48.0(3)
< 8.0
< 45.0 Not applicable
< 7.0
< 40.0
< 2. 5
< 11.0
. < 60.0
< 60.0 I
TAB LE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION
- 10. Undervoltage RCP bus
- a. Turbine-Driven Auxiliary Feedwater Pumps
- 11. Containment Radioactivity - High
- a. Purge and Exhaust Isolation*
- 12. Trip of Feedwater Pumps
- a. Auxiliary Feedwater Pumps
- 13. Undervoltage, Vital Bus
- a.
Loss of Voltage
- 14. Station Blackout
< 60.0
< 5.0(6)
Not App 1i cab 1 e
< 4.0
< 60 II I, l
TAB LE 3.3-5 (Continued)
TAB LE NOTATION (1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establish SJ path and attainment of discharge pressure for centrifugal chargi~g pumps, SI and RHR pumps.
(2 Diesel generator starting and sequence loading delays not included. Offsite power available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(3) Diesel generator starting and sequence loading delays included. Response time limit opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(4) On 2/3 in any steam generator.
(5) On 2/3 in 2/4 steam generators.
(6) The response time is the time the isolation circuitry input reaches the isolation setpoint to the time the Containment Pressure-Vacuum Relief are fully shut.
SALEM - UNIT 1
- 3/4 3-31.
ATTACHMENT 0 UNIT 2 PROPOSED SPECIFICATION
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l.O TAB LE 3. 3-2 REACTOR TRIP SYSTEM INSTRLMENTATION RESPONSE ITEMS F UNCT IONA L UN IT RESPONSE TIME
- 1. Manual Reactor Trip NOT APP LICAB LE
- 2. Power Range, Neutron Flux
< 0.5 seconds*
- 3. Power Range, Neutron Flux, NOT APP LICAB LE High Positive Rate
- 4.
Power Range, Neutron Flux,
< 0.5 seconds*
High Negative Rate
- 5.
Intermediate Range, Neutron Flux NOT APP LICAB LE
- 6. Source Range, Neutron Flux NOT APPLICABLE
- 7. Overtemperatu re.6. T
< 5.75 seconds*
- 8. Overpower.6 T NOT APPLICABLE
- 9. Pressurizer Pressure--Low
< 2.0 seconds
- 10. Pressurizer Pressure--High
< 2.0 seconds
- 11. Pressurizer Water Level--High NOT APP LICAB LE
- Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
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TAB LE 3. 3... 5 ENGINEERED SAFETY FEATURES RESPONSE ITEMS INITIATING SIGNAL AND FUNCTION
- 1. Manual
- 2.
- a. Safety Injection (ECCS)
Feedwater Isolation Reactor Trip (SI)
Containment Isolation-Phase "A" Containment Ventilation Isolation Auxiliary Feedwater Pumps Service Water System Containment Fan Cooler
- b. Containment Spray Containment Isolation-Pha~e "B" Containment Ventilation Isolation
- c. Containment Isolation-Phase "A" Containment Ventilation Isolation d..
Steam Line Isolation Containment Pressure-High
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Isolation-Phase "A"
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System SA LEM - UNIT 2 3/4 3-28
.. RESPONSE TIME IN SEC ON a; Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable
< 27.0 (l_)
< 2.0
< 7.0
~ 17.0(2)/27.0(3)
Not Applicable
< 60
- 2. 13.0(2)/48.0(3)
f' TAB LE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE ITEMS INITIATING SIGNAL AND FUNCIION
- 3. Pressurizer Pressure-Low
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment.Isolation-Phase "A"
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System
- 4.
Differential Pressure Between Steam Lines-High
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Iso 1 at ion-Phase "A"
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System
- 5. Steam Flow in two Steam Lines - High Coincident with T avg --Low-Low
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment Isolation-Phase "A"
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service Water System
~ 21.0(1)112.0(2)
< 2.0
< 7.0
< 18.0(2)
Not Applicable
< 60
~ 49.0(l)/13.0(2)
~ 12.0(2)122.0(3)
< 2.0
< 7.0
~ 11.0(2)121.0(3)
Not Applicable
< 60
~ 13.o(2)/4a.o(3)
~ 15.75(2)/25.75(3)
< 5.75
< 10. 75
~ 20.75(2)/30.75(3)
Not Applicable
< 61.75
~ 15.75(2)/50.75(3)
< 10. 75
)
i TAB lf 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES IN IT IA TING SIGNAL AND FUNCTION
- 6. Steam fl ow in Two Steam Lines-High Coincident with Steam Line Pressure-Low
- a. Safety Injection (ECCS)
- b. Reactor Trip (from SI)
- c. Feedwater Isolation
- d. Containment I sol ati on-Phase uAu
- e. Containment Ventilation Isolation
- f. Auxiliary Feedwater Pumps
- g. Service.Water System
- h. Steam Line Isolation
- 7. Containment Pressure--High-High
- b. Containment Isolation-Phase 118 11
- c. Steam Line Isolation
- d. Containment Fan Cooler
- 8. Steam Generator Water Level--High-High
- a. Turbine Trip
- b. Feedwater Isolation
- 9. Steam Gene-rator Water Level--Low-Low
- a. Motor-Driven Auxiliary Feedwater
- Pumps(4)
- b. Turbine-Driven Auxi*liary Feedwater Pumps(5)
SA LEM - UN IT 2 3/4 3-30 RESPONSE TIME IN SECONDS
- 2. 12.0(2);22.0(3)
< 2.0
< 7.0
- 2. 11.0(2);21.0(3)
Not Applicable
< 60
! 14.0(2)/48.0{3)
< 8.0
< 45.0 Not applicable
< 7.0
< 40.0
< 2. 5
< 11.0
< 60.0
< 60.0
't(
l TAB LE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION
- 10. Undervoltage RCP bus
- a. Turbine-Driven Auxiliary Feedwater Pumps
- 11. Containment Radioactivity - High
- a. Purge and Exhaust Isolation
- 12. Trip of Feedwater Pumps
- a. Auxiliary Feedwater Pumps
- 13. Undervoltage, Vital Bus
- a.
Loss of Voltage
- 14. Station Blackout
- a. Motor-Driven Auxiliary Feed Pumps SA LEM -
UN IT 2 3/4 3-31 RESPONSE TIME IN SECONDS
< 60.0
< s.0(6)
Not Applicable
< 4.0
< 60
TAB LE 3.3-5 (Continued)
TAB LE NOTATION
( 1) Diesel generator starting and sequence loading delays included. Response time limit includes opening of valves to establJsh SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pumps.
(2 Diesel generator starting and sequence loading delays not included. Offsite power available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(3) Diesel generator starting and sequence loading delays included. Response time limit opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.
(4) On 2/3 in any steam generator.
(5) On 2/3 in 2/4 steam generators.
(6) The response time is the time the isolation circuitry input reaches the isolation setpoint to the time the Purge and Exhaust Valves Relief are fully shut.
SALEM - UNIT 2 3/4 3-32
BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION The proposed change does not involve a significant hazards consideration because operation of Salem Generating Station Units 1 and 2 in accordance with this change would-not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated.
The probability of previously analyzed accident is discussed first.
The proposed change in measured response time will not increase the probability of such an accident because the numerical value of the LCO on OT~T or steam flow in two steam lines -
high coincident with Tavg - low-low ESF actuation response times is not a factor in the initiation of a previously evaluated accident.
The removal and replacement of the existing RTDs and bypass line el imi nation wi 11 not significantly increase the probability of occurrence of an accident previously eva)uated.
The events of interest are those initiated by a failure of those components affected by the proposed change.
There are four such events:
(1) Uncontrolled Withdrawal of a Control Rod at Power, (2) Excessive Load Increase, (3) Accidental Depressurization of the Main.*
Steam System, and (4) Small Break Loss of Coolant Accident (SBLOCA).
The Uncontrolled Rod Withdrawal event is an ANS Condition II (moderate frequency) event potentially initiated by a failure of the reactor control system.
The Excess Load and Accidental Oepressurization of the Main Steam System events are also Condition II events.
They are potentially initiated by a failure of the steam dump control system.
The input to the reactor control system and steam dump control system from the replacement RTDs will be equivalent to those currently provided by the existing RTDs.
The proposed modification will be done in a manner consistent with the plant design bases.
As such, there will be no degradation in the per~ormance of or increase of the number of cha 11 enges to safety systems assumed to function in the accident analysis.
Further-more, there will be no increase in the probability of failure of or degradation of the performance of the systems designed to reduce the number of challenges to safety systems.
Hence, the first three events will remain Condition II events.
~*.
The SBLOCA is an ANS Condition III (infrequent) event.
It could be initiated by the highly unlikely ejection of a thermowell or the failure of a cap covering one of the existing pump suction leg penetrations.
The scoops, cross over leg buttweld caps RVLIS, and thermowells will be analyzed to the ASME Boiler and Pressure Vessel Code,Section III, Class 1 and installed in accordance with the requirements of Section XI of this Code.
As such, the RCS pressure boundary will not be degraded.
The SBLOCA will thus remain a Condition III event.
Additionally, approximately 280 feet of small diameter pipe and the associated valves will be removed from the primary system pressure boundary, eliminating the possibility of a SBLOCA from these locations.
Hence, there will be no significant increase in the probability of occurrence of an accident previously evaluated in the SAR.
There will be no increase in the consequences of a previously evaluated~
~:*
. accident..In assessing the impact on the consequences of a previously,
evaluated accident, there are four events of interest:
(1) Uncontrolled Boron Dilution During Full Power, (2) Loss of External Load, (3) Uncon-trolled Withdrawal of a Control Rod at Pbwer, and (4) Major Secondary Pipe Rupture.
The first three events are of interest because the OT~T
__ --~rip is the primary trip credited in the safety ana 1 yses.
The fourth event is considered because steam flow in two steam lines - high
~coincident with Tavg-low-low is one of the signals credited to initiate
~n Engineered Safety Features actuation.
The OT~T trip will continue to function in a manner consistent with the existing analysis assumptions for the first three events.
The actual response time will be within the six seconds currently assumed.
Similarly, the ESF response times will remain within-those assumed in the safety analysis.
Hence there wi1*1 be no increase in the consequences of previously evaluated accident.
(2) create the possibility.of a new or different kind of accident from any previously analyzed.
The proposed change will be performed in a manner consistent with the applicable standards, preserve the existing design bases, and will not adversely impact the qualification of any plant systems.
This will preclude adverse control/protection systems interactions.
The design, installation, and inspection of the new
equipment will be done in accordance with ASME Boiler and Pressure Vessel Code criteria.
By adherence to industry standards, the pressure boundary integrity will be preserved.
As such, the possibility of a new or different kind of accident is not created.
(3) involve a significant reduction in a margin of safety.
The applicable margins of safety are defined in Technical Specification Bases Sections 2.1.1 and 2.1.2.
Bases Section 2.1.1 states that the minimum value of the Departure from Nucleate Boiling Ratio (ONBR) during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that Departure from Nucleate Boiling (DNB) will not o~cur. The restrictions of this fuel cladding integrity safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products.to the coolant.
The proposed change will not result in a decrease in the minimum DNBR reported in the UFSAR accident analyses.
Bases Section 2.1.2 states that the Safety Limit on maximum RCS pressure is 2735 psig.
This Safety Limit protects the integrity of the RCS from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The proposed change will not result in an increase in the maximum RCS pressure reported in the UFSAR accident analyses.
The Commission has pr~vided guidance concerning the application of the standards for determining whether a significant hazards consideration exists by providing examples (51 FR 7751) of amendments that are considered not likely to involve significant hazards consideration.
Example (ix) is:
A repair or replacement of a major component or system 1mportant to safety if the following conditions are met:
I (1)
The repair or replacement process involves practices which have been successfully implemented at least once on similar components or systems elsewhere in the nuclear industry or in other industries, and does not
- ~,
involve a significant increase in the probability or consequences of an accident previously evaluated or create the possibility of a new or different kind of accident from any accident previously evaluated; and (2)
The repaired or replacement component or system does not result in a significant reduction in any safety limit (or limiting condition of operation) associated with the component or system.
The proposed changes to the Salem Units 1 and 2*Technical Specifications is similar to changes approved at Byron Station Units 1 and 2 (52 FR 2785).
As discussed earlier, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated or create the possibility of a new or different kind of accident from any previously evaluated. It does not result in a significant change in a component or system safety function or a significant reduction in any associated safety limit or limiting condition of operation.
Therefore, based on the above consideratfons, it has been determined that the proposed change does not involve a s1gnificant hazards consideration.
The addition of response times to the Auxiliary Feed Pump Section of table 3.3.5 and the addition of Station Blackout requitements corresponds to example 2 of 48 FR 14870 as changes that impose an additional limitation not currently in the Technical Specifications. Changes to the footnotes were done to correct typographical errors and as such correspond to example 1 of 48 FR 14870.
In either case the changes will not involve an increase in the probability or consequences of a previously analyzed accident, create a new or different kind of accident that previously analyzed, or reduce the margin of safety since the changes are being done to be consistent with previously reviewed and approved analyses.