ML20132B740
| ML20132B740 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/11/1996 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20132B735 | List: |
| References | |
| NUDOCS 9612170370 | |
| Download: ML20132B740 (25) | |
Text
... -
ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNIT 8 1, 2, AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE TS-386 NARKED PAGE8 I.
AFFECTED PAGE LIST Units 1.
2.
and 3 1.2/2.2-1 1.2/2.2-2 3.6/4.6-30 Unit 1 only 1.2/2.2-3 1.2/2.2-4 II.
MARF"U PAGES see attached.
9612170370 961211 PDR ADOCK 05000259 P
1 4
l I
1 1 2/2.2 REAcrom CX)OUWT SYSTEM INTEGRITY SAFRTY LIMIT LIMITING SAFETY SYSTEM SETTING 1.2 Reactor Coolant System Integrity 2.2 Reactor Coolant System Integrity Applicability ADD 11cability 4
Applies to limits on reactor coolant Applies to trip settings of the i
system pressure.
instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
1 objective obiective To establish a limit below which To define the level of the j
the integrity of the reactor process variables at which coolant system is not threatened automatic protective action due to an overpressure condition.
is initiated to prevent the pressure safety limit free being exceeded.
Specifications SDecifications i
A.
The pressure at the lowest point The limiting safety system j
of the reactor vessel shall not settings shall be as specified 3
exceed 1.3'l5 psig whenever below:
l i
irradiated fuel is in the M t~$ eat Limiting Safety I
i reactor vessel.
Protective Action System Settinq fA.
v#
Nuclear system 1,105 ps
[ A.
Verify the safety function lift settings o j
the required S/RVs are within 2,M of the open-nuclear (4 4Wes) setpoint as follows:
p system pressure ihree f'<MY Nusber of Setpoint 1,115 psig ~+
1 S/RVs fosial gg pgg 4
1105
/
(4 valves)
{
)
4 1115 1.125 psig i 11 psi 0 f t.ere 5
1125 (5 valves) 4 Following testing, lift settings shall be
^~
within t,,14p orle pycer1f.
8.
Scram--nuclear 11,055 psig j
system high
-e pressure I
hyw 1.2/2.2-1 unit I
I 1.2 BASES 4
REACTOR C00UWT SYSTEM INTOGRITY j
The safety limits for the reactor coolant system pressure have been selected such that they are below pressures at which it can be shown that 4
the integrity of the system is not endangered. However, the pressure
(-
safety limits are not high enough such that no foreseeable circumstances can cause the system pressure to rise over these limits. The pressure safety limits are arbitrarily selected to be the lowest transient i
overpressures allowed by the applicable codes. ASam Boiler and Pressure j
Vessel code Section III, and USAS Piping Code, Section 831.1.
l The design pressure (1,250 psig) of the reactor vessel is established l
such that, when the 10 percent allowance (125 psi) allowed by the ASME Boiler and pressure Vessel Code section III for pressure transients is 1
added to the design pressure, a transient pressure limit of 1,375 psig is established.
l Correspondingly, the design pressure (1,148 psig for suction and 1,326 psig for discharge) of the reactor recirculation system piping is such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS 1
Piping Code, Section B31.1 for pressure transients is added to the design i
pressures, transient pressure limits of 1.378 and 1,591 psig are established. Thus, the pressure safety limit applicable to power i
j operation is established at 1,375 psig (the lowest transient overpressure l
allowed by the pertinent codes). ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code Section 831.1.
1 (Thecurrentcycle'ssafetyanalysisconcerningthemostsevereabnormal operational transient resulting directly in a reactor coolant system pressure increase is given in -..._..__.. % e. ee b $JL
- k. c e u 4
' { *' 4
(> o < % e 4. v e < < wk y 3 i
( Cg,, a y The reactor el pressure code limit of 1,375 psig given in subsection 4 f the safety analysis report is well above the peak h
pressure prod ed by the overpressure transient described above. Thus, j
the pressure saYety limit applicable to power operation is well above the l
at can result due to reasonably expected overpressure kpeakpressure transients.
)
Higher design pressures have been established for piping within the
]
reactor coolant system than for the reactor vessel. These increased l
design pressures create a consistent design which assures that, if the f
pressure within the reactor vessel does not exceed 1,375 psig, the pressures within the piping cannot exceed their respective transient j
pressure limits due to static and pump heads.
]
The safety limit of 1,375 psig actually applies to any point in the reactor vessel; however, because of the static water head, the highest pressure point will occur at the bottom of the vessel. Because the l
1 1
)
BFW 1.2/2.2-2 Unit 1 1
_ -.... ~ -. = _ - -
i 4
5 1.2 BASES (Cont'd) pressure is not monitored at this point, it cannot be directly determined if this safety limit has been violated. Also, because of the potentially varying head level and flow pressure drops, an equivalent pressure cannot be a priori determined for a pressure monitor higher in the vessel.
Therefore, following any transient that is severe enough to cause concern that this safety limit was violated, a calculation will be performed using all available information to determine if the safety limit was violated.
.=0 D (Q"kD N REFERENCES 1.
Plant Safety Analysis (BPNP PSAR Secti 4.0 2.
Ab.M Boiler and Pressure Vessel Code Section III 3.
USAS Piping Code, Section B31.1 4.
Reactor Vessel and Appurtenances Mechanical Design (BPNP PSAR Subsection 4.2) 5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-P-A and Addenda.
i BPW 1.2/2.2-3 Unit 1
~
4 A
a a-4_
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4 m.
.4_.
,__.J>
4
1 l
l 2.2 BASBS j
i REACTOR COOLANT SYSTEM INTEGRITY
- gJA, To meet the safety basis, 13 relief valves /
been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flowg. ; ;;f;;r.;; ;;;;;;.re ;f 41,10",
i _e wud voi @ The analysis of the worst overpressure transient (3-second closure of all main steam line isolation valves), neglecting the direct scram (valve position scram), results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves operable, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
l To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowable vessel overpressure of 1,375 psig.
i J
1 J
1 l
l i
a e
l I
BFW 1.2/2.2-4 Unit 1 I
3.6/4.6 BASES AUS 03193g 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would i
not result from a crack approaching the critical size for rapid j
propagation. Leakage less than the magnitude specified can be detected 2
reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a rv.sonably
.l.
short time, the unit should be shut down to allow further investigation and corrective action.
l The two spe limit for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> perfod is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
1 The. total leckage rate consists cf all leakage, identified and 4
i unidentified, which flows to the d.ywell floor drain and equipment drain sumps.
The capacity of the drywell floor aump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
REFERENCE
- 1. Nuclear System Ledage Rate Limits (BFNP FSAR Subsection 4.10)
]
- 2. Safety Evaluation ngort (SER) on IE Bulletin 82-03 i
3.6.D/4.6.D Relief Valves d e.$ 4 g t To meet the safety basis, 13 relief valves have been installed on the j
unit with_a total capacity of 84.1 per/ent of nuclear boiler rated steam flowkt_a_ refer:::c;rer::: Of (1,1^s. 1 perc a ) peig d the analysis i
of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position serem) 1 results in a maximum vessel pressure ich, if a neutron flux scram is assumed considering 12 valve _operible_. results in adequate margin to the code allowable overpressure liinit of 1,p5 psig.
,, ~.
To meet operational design, the analysis of the plant isolation traasient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
pqCAr Experience in reliefy alve operation shows that a testing of 50 percent of the valves per @J is adequate to detect failures or deteriorations.
_The relief valves are benchtested every second operatin -
le to ensure that their setpoints are within CG; i _1 peuent)toleran The relief valves are tested in place in accordance with Specificat on 1.0.MM to establish that they will open and pass! steam.
3 w.
S f ' U k' ' d
'J""'O AM ML170 BFN 3.6/4.6-30 Unit 1
1.2/2.2 mCTOR COOLANT SYSTEM TFfEGRITY SAFETY LIMIT LIMITING SAiziz SYSTEM SETTING 1.2 Reactor Coolant System Intearity 2.2 Reactor coolant System Intearity Ano11cability Acolicability Applies to limits on reactor coolant Applies to trip settings of the syntaa pressure.
instruments and devices which are provided to prevent the reactor system safety limits from being axceeded.
Obiective Obiective l
To establish a limit below which To define the level of the the integrity of the reactor process variables at which coolant system is not threatened automatic protective action due to an overpressure condition.
is initiated to prevent the Pressure safety limit from being exceeded.
Specifications Snecifications i
A.
The pressure at the lowest point The limiting safety system of the reactor vessel shall not settings shall be as specified exceed 1,375 paig whenever below:
irradiated fuel is in the J3F Limiting Safety reactor vessel.
Protective Action System Settina A.
Nuclear systen 1,105 psig i relief valves 11 poi A.
Verify the safsty function lift settings of 4 of the open--nuclear (4 valves) the required S/RVs are within t )7 setpoint as follows:
system pressure 4hrte P'CtN Number of Setpoint S/RVs (osia) 1,115 peig i 11 pai 4
1105 (4 valves) 4 1115 1,125 psig i 11 pai 5
1125 i
(5 valves)
Following testing, lift settings shall be withint,14[o#1C p ycerif.
De e B.
Scras--nuclear 11,055 paig system high
_ _ =.
pressure i
BFN 1.2/2.2-1 Unit 2
m.._
1 j
l 1.2 B&114 j
NOV 17 tg REACTOR C001 ANT SYSTEM INTEGRITY 1
i The safety limits for the reactor coolant system pressure have been selected such that they are below pressures at which it can be shown that 4
the integrity of the system is not endangered. However, the pressure j
safety limits are not high enough such that no foreseeable circusstances j
can cause the system pressure to rise over these limits. The pressure i
safety limits are arbitrarily selected to be the lowest transient i
overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
i The design pressure (1,250 pais) of the reactor vessel is established such that, whos the 10 percent allowance (125 psi) allowed by the ASME j
Boiler and Pressure Vessel Code Section III for pressure transients is added to the design pressure, a transient pressure limit of 1,375 pais is established.
Correspondingly, the design pressures (1,143 for auction and 1,326 for f
discharge) of the reactor recirculation system piping are such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS Piping Code,
}
Section 531.1 for pressure transients is added to the design pressures, l
transient pressure limits of 1,378 and 1,591 pais are established. Thus, i
the pressure safety limit applicable to power operation is established at l
1,375 psis (the lowest transient overpressure allowed by the pertinent
{
codes), ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section 531.1.
-repor4 l
Thecurrentcycle'ssafety,analysisconcerningthemosts[vereabnormal l
operational transient resulting directly in a reactor coc lant system pressureincreaseisgiveninthereloadlicensing")I~t~of-@1,375pais W
or the j
current cycle. The reactor vessel pressure code li given in subsection 4 of the safety analysis report is well above the l
peak pressure produced the overpressure transient described above.
Thus, the pressure safety imit applicable to power operation is well j
above the peak pressure t can result due to reasonably expected
{
overpressure transients.
A 4
i Higher design pressures have been established for piping within the i
reactor coolant system than for the reactor vessel. These increased j
design pressures create a consistent design which assures that, if the pressure within the reactor vessel does not exceed 1,375 pais, the pressures within the piping cannot exceed their respective transient pressure limits due to static and pump heads.
i The safety. limit of 1,375 psig actually applies to any point in the reactor vessel; however, because of the static water head, the highest i
pressure point will occur at the bottom of the vessel. Because the i
i j
BFN 1.2/2.2-2 TS 370 Unit 2 Letter Dated 11/17/95
i l
3.6/4.6 BA H A NOV 171995 3.6.B/4.6.C (Cont'd) i five gym, as specified in 3.6.C, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage 1
detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further j
investigation and corrective action.
i The two spa limit for coolant leakage rate increases over any 24-hour l
il period is a limit specified by the NRC (Reference 2). This limit applies only during the RUN mode to avoid being penalized for the
{
expected coolant leakage increase during pressurization.
l
)
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain j
The capacity of the drywell floor sump pump is 50 spa and the capacity i
of the drywell equipment sump ptmp is also 50 syn. Removal of 25 spa from either of these sumps can be accomplished with considerable margin.
)
REFERENCE
)
- 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
- 2. Safety Evaluation Report (SER) on II Bulletin 32-03 1
1 3.6.D/4.6.D Relief Valves l
l To meet the safety basis, 13 relief valves have been installed on the l
unit with a total capacity of 34.1 percaat of nuclear boiler rated staan j
flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct i
scram (valve position scram) results in a maximum vessel pressure which, l
if a neutron flux scram is asstard considering 12 valves OPWRARf E, i
results in adequate margin to the code allowe' ole overpressure limit of 1,375 peig.
To meet operational design, the analysis of the plant isolation j
transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a j
value which is well below the allowed vessel overpressure of 1,375 pais.
c.pJLe Experience in relief 17e operation shows that a testing of 50 percent of the valves per is adequate to detect failures or i
deteriorations. The relief valves are benchtested every second
,_ operating cycle to ensure that their setpoints are within 6 r 1,
withSpecificati@on1.0.191toestablishthattheywillopenandpass percent)toleranc The relief valves are tested in place in accordance 4
steam.
4 ibelv spWUhle BFN 3.6/4.6-30 TS 370 Unit 2 Letter Dated 11/17/95 1
1.2/2.2 DRACTOR COOLANT SYSTEM INTEGRITY LIMITING SAFETY SYSTEM SETTING SAmi LIMIT i
l 1.2 Reactor Coolant Sveten Intearity 2.2 Reactor Coolant System Intearity Anolicability Acolicability Applies to limits on reactor coolant Applies to trip settings of the instruments and devices which system pressure.
are provided to prevent the reactor system safety limits from being exceeded.
i Obiective j
Obiective 4
To establish a limit below which To define the level of the j
the integrity of the reactor process variables at which coolant system is not threatened automatic protective action due to an overpressure condition.
is initiated to prevent the pressure safety limit from being exceeded.
r SnecificationS Snecificatibn3 A.
The pressure at the lowest point The limiting safety system 4
of the reactor vessel shall not settings shall be as specified belovt i
exceed 1,375 paig whenever 4
irradiated fuel is in the reactor vessel.
~ 1,105 psig Nuclear system A.
relief valves 11 psi open-nuclear a ves) 1 A.
verify the safety function lift settings of system pressure the required S/RVs are within 2 Jt of the setpoint as follows:
f 1,115 pais i g ee p,cenf 11 pai Number of Setpoint S/RVs (maia)
(4 gglygg) 4 1105 1,125 psig i 11 psi 4
1115
)
(5 valves) w DElm 5
1125 Following testing, lift settings shall be 1,o55 pois 1
n.
sers.-nuclear within t J4 c'" P"ccrit.
system high pressure Limiting Safety Protective Action Svaten Settina 1.2/2.2-1 BFN Unit 3
1.2 BASES g 17 g
)
REACTOR COOLANT SYSTDI INTEGRITT The safety limits for the reactor coolant system pressure have been selected such that they are below pressures at which it can be shown that the integrity of the system is not endangered. However, the pressure safety limits are set high enough such that no foreseeable circumstances can cause the system pressure to rise over these limits. The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicabis codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section 531.1.
The design pressure (1,250 pais) of the reactor vessel is established such that, when the 10 percent allowance (125 psi) allowed by the ASME Boiler and Pressure Vessel Code Section III for pressure transients is added to the design pressure, a transient pressure limit of 1,375 pais is established.
Correspondingly, the design pressures (1,143 for suction and 1,326 for l
discharge) of the reactor recirculation system piping are such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS Piping Code, Section B31.1 for pressure transients is added to the design pressures, l
transient pressure limits of 1,378 and 1,591 pais are established. Thus, the pressure safety limit applicable to power operation is established at 1,375 pais (the 3owest transient overpressure allowed by the pertinent codes), ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
( n g o'4 The current cycle's safety analysis concerning the most vere abnormal operational transient resulting directly in a reacto lant system pressure increase is given in the reload licensing @
for the current cycle. The reactor vessel pressure code limit of 1,375 pais given in subsection 4.($ of the safety analysis report is well nova,the peak pressure produced by the overpressure transient described above.
Thus,thepressuresafety3imitapplicabletopoweroperationiswell above the peak pressure that can result due to reasonably expected overpressure transients. (_,y Higher design pressures have been established for piping within the reactor coolant system than for the reactor vessel. These increased design pressures create a consistent design which assures that, if the pressure within the reactor vessel does not exceed 1,375 pais, the pressures within the piping cannot exceed their respective transient pressure limits due to static and pump heads.
The safety limit of 1,375 pais actually applies to any point in the reactor vessel; however, because of the static water head, the highest pressure point will occur at the bottom of the vessel. Because the BFN 1.2/2.2-2 TS 370 Unit 3 Letter Dated 11/17/95
3.6/4.6 R&gfd NOV 171995 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not resvit from a crack approaching the critical size for rapid propagation. Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The two spa limit for coolant leakage rate increases over any 24-hour l
period is a limit specified by the NRC (Reference 2). This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
The capacity of the drywell floor sump punp is 50 spa and the capacity of i
the drywell equipment stamp punp is also 50 syn. Removal of 25 spa from either of these sumps can be accomplished with considerable margin.
References 1.
Nuclear System Leakage Rate Limits (BFEP FSAR Subsection 4.10) 2.
Safety Evaluation Report (SER) on II Bulletin 82--03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam l
flow. The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure whica, if a neutron flux scram is assaned considering 12 valves 0FunART.I, results in adequate margin to the code allowable overpressure limit of l
1,375 pais.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 peig.
c a cJt e Experience in reli alve operation shows that a testing of 50 percent i
of the valves per is adequate to detect failures or deteriorations.
The relief valves are benchteste Levert second operating Ac cle to ensure i that their setpoints are withint%; i i ; rt)tolerancu > The relief valves are tested in place in accordan with Specification 1.0.991 to establish that they will open and pass steam.
Ye'e r $f e(sfiek BFN 3.6/4.6-30 TS 370 Unit 3 Letter Dated 11/17/95
(
-+-
ENCLO3URE 3 TENNESSEE VALLEY AUTHORITY BRONN8 FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2,
AND 3 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE T8-386 l
REVISED PAGES I.
AFFECTED PAGE LIST Units 1.
2.
and 3 i
1.2/2.2-1 1.2/2.2-2 3.6/4.6-30 Unit 1 only 1.2/2.2-3 1.2/2.2-4 II.
REVISED PAGES See attached.
l 1
i I
?
~
. - -_. ~ - - ~ -. - -
~.. - -. - ~ ~
~. _.-.+
++-x 1.2/2.2 REACTOR COOLANT SYSTE*4 INTEGRITY I
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.2 Reactor coolant System Inteority 2.2 Reactor coolant System Intecrity l
Aeolicability Aeolicability Applies to limits on reactor coolant Applies to trip settings of the system pressure.
instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
Obiective Obiective To establish a limit below which To define the level of the the integrity of the reactor process variables at which coolant system is not threatened automatic protective action due to an overpressure condition, is initiated to prevent the pressure safety limit from being exceeded.
Soecifications Snecifications A.
.The pressure at the lowest point The limiting safety system of the reactor vessel shall not settings shall be as specified exceed 1,375 psig whenever below:
irradiated fuel is in the reactor vessel.
A.
Verify the safety function lift settings of the required S/RVs are within i three percent of the setpoint as follows:
Number of Setpoint S/RVs (esia) 4 1105 4
1115 5
1125 Following testing, lift settings shall be within one percent.
(
Limiting Safety Protective Action System Settina j
1,055 psig B.
Scram--nuclear 1
system high pressure BFN 1.2/2.2-1 Unit 1
l 1.2 BASES j
REACTOR COOLANT SYSTEM INTEGRITY l
The safety limits for the reactor coolant system pressure have been selected such that they are below pressures at which it can be shown that i
the integrity of the system is not endangered. However, the pressure l
safety limits are not high enough such that no foreseeable circumstances can cause the system pressure to rise over these limits. The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure l
Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The design pressure (1,250 psig) of the reactor vessel is established such that, when the 10 percent allowance (125 psi) allowed by the ASME Boiler i
and Pressure Vessel Code Section III for pressure transients is added to the design pressure, a transient pressure limit of 1,375 psig is established.
Correspondingly, the design pressures (1,148 psig for suction and 1,326 1
psig for discharge) of the reactor recirculation system piping is such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS Piping Code, Section B31.1 for pressure transients is added to the design pressures, transient pressure limits of 1,378 and 1,591 psig are established.
Thus, the pressure safety limit applicable to power operation is established at 1,375 psig (the lowest transient overpressure allowed by the pertinent codes), ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
)
The current cycle's safety analysis concerning the most severe abnormal operational transient resulting directly in a reactor coolant system pressure increase is given in the reload licensing report for the current cycle. The reactor vessel pressure code limit of 1,375 psig given in subsection 4.4 of the safety analysis report is well above the peak pressure produced by the overpressure transient described above.
Thus, the pressure safety limit applicable to power operation is well above the peak pressure that can result due to reasonably expected overpressure transients.
Higher design pressures have been established for piping within the reactor coolant system than for the reactor vessel. These increased design
~
pressures create a consistent design which assures that, if the pressure within the reactor vessel does not exceed 1,375 psig, the pressures within i
the piping cannot exceed their respective transient pressure limits due to static and pump heads.
The safety limit of 1,375 psig actually applies to any point in the reactor vessel; however, because of the static water head, the highest l
pressure point will occur at the bottom of the vessel.
Because the i
l BFN 1.2/2.2-2 Unit 1 i
i l
_ ~.
.. -. -.. -.. -.. - ~.
l l
\\
1.2 BASES (Cont'd) pressure is not monitored at this point, it cannot be directly determined if this safety limit has been violated. Also, because of the potentially l
varying head level and flow pressure drops, an equivalent pressure cannot l
be a priori determined for a pressure monitor higher in the vessel.
l Therefore, following any transient that is severe enough to.cause concern that this safety limit was violated, a calculation will be performed using all available information to determine if the safety limit was violated.
i i
REFERENCES l
1.
Plant Safety Analysis (BFNP FSAR Sections 14.0 and Appendix N) l 2.
ASME Boiler and Pressure Vessel Code Section III 3.
USAS Piping Code, Section B31.1 4.
Reactor Vessel and Appurtenances Mechanical Design (BFNP FSAR Subsection 4.2) 5.
Generic Reload Fuel Application, Licensing Topical Report, NEDE-24011-
'P-A and Addenda.
l b
l 1
{
l
(
BFN 1.2/2.2-3 Unit 1
.__..m..
..m
_q l
2.2 BASES REACTOR COOLANT SYSTEM INTEGRITY 1
l j.
To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow.
l The analysis of the worst overpressure transient (3-second closure of all main steam line isolation valves), neglecting the direct scram (valve position scram), results in a maximum vessel pressure which, if a neutron I
flux scram is assumed considering 12 valves operable, results in adequate g
l margin to the code allowable overpressure limit of 1,375 psig.
J To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well l'
below the allowable vessel overpressure of 1,375 psig.
l 1
I l
i 1
l d
1 l
i BFN 1.2/2.2-4 L
Unit i l
l-I' I
.~
l d
3.6/4.6 BASEE 3.6.C/4.6.C (Cont'd) l suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid l
propagation.
Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably i
short time, the unit should be shut down to allow further investigation and corrective action.
The two gpm limit for coolant leakage rate increase over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant 1
' leakage increase during pressurization.
l The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
l The capacity of the drywell floor sump pump is 50 gpm and the capacity of j
the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
REFERENCE
- 1. Nuclear System Leakage Rate Limits -(BFNP FSAR Subsection 4.10)
I
- 2. Safety Evaluation Report. (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent.of nuclear boiler rated steam flow, d
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate-l margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well betow the allowed vessel overpressure of 1,375 psig.
Experi.ence in relief valve operation shows that a testing of 50 percent of the valves per cycle is adequate to detect failures or deteriorations.
l The relief valves are benchtested every second operating cycle to ensure l
that their setpoints are within their specified tolerances. The relief l
l valves are-tested in place in accordance with Specification 1.0.MM to l
establish that they will open and pass steam.
I l
I BFN 3.6/4.6-30 Unit 1 l
m n
w-
i I
1.2/2.2 REACTOR COOLANT SYSTEM INTEGRITY i
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1.2 Reactor coolant System Intecrity 2.2 Reactor coolant System Intecrity Aeolicability Aeolicability Applies to limits on reactor coolant Applies to trip settings of the system pressure.
instruments and devices which are provided to prevent the reactor system safety limits from being exceeded.
Obiective Obiective To establish a limit below which To define the level of the the integrity of the reactor process variables at which coolant system is not threatened automatic protective action due to an overpressure condition.
is initiated to prevent the pressure safety limit from being exceeded.
Snecifications Soecifications A.
The pressure at the lowest point The limiting safety system of the reactor vessel shall not settings shall be as specified exceed 1,375 psig whenever below:
irradiated fuel is in the reactor vessel.
A.
Verify the safety function lift settings of the required S/RVs are within 1 three percent of the setpoint as follows:
Number of Setpoint S/RVs (esic) 4 1105 4
1115 5
1125 Following testing, lift settings shall be within one percent.
Limiting Safety Protective Action System Settina B.
Scram--nuclear 11,055 psig j
system high pressure l
I BFN 1.2/2.2-1 Unit 2
1.2 gaggs REACTOR COOLANT SYSTEM INTEGRITY l
l The safety limits for the reactor coolant system pressure have been selected such that they are below pressures at which it can be shown that the integrity of the system is not endangered. However, the pressure l
safety limits are not high enough such that no foreseeable circumstances can cause the system pressure to rise over these limits.
The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The design pressure (1,250 psig) of the reactor vessel is established such that, when the 10 percent allowance (125 psi) allowed by the ASME Boiler and Pressure Vessel Code Section III for pressure transients is added to the design pressure, a transient pressure limit of 1,375 psig is established.
Correspondingly, the design pressures (1,148 for suction and 1,326 for discharge) of the reactor recirculation system piping are such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS Piping Code, Section B31.1 for pressure transients is added to the design pressures, transient pressure limits of 1,378 and 1,591 psig are established.
- Thus, the pressure safety limit applicable to power operation is established at 1,375 psig (the lowest transient overpressure allowed by the pertinent codes),
ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The current cycle's safety analysis concerning the most severe abnormal operational transient resulting directly in a reactor coolant system pressure increase is given in the reload licensing report for the current l
cycle. The reactor vessel pressure code limit of 1,375 psig given in subsection 4.4 of the safety analysis report is well above the peak l
pressure produced by the overpressure transient described above. Thus, the pressure safety limit applicable to power operation is well above the peak pressure that can result due to reasonably expected overpressure transients.
Higher design pressures have been established for piping within the reactor coolant system than for the reactor vessel.
These increased design pressures create a consistent design which assures that, if the pressure within the reactor vessel does not exceed 1,375 psig, the pressures within the piping cannot exceed their respective transient pressure limits due to static and pump heads.
The safety limit of 1,375 psig actually applies to any point in the reactor vessel; however, because of the static water head, the highest pressure point will occur at the bottom of th vessel. Because the BFN 1.2/2.2-2 Unit 2 l
3.6/4.6 EASES 3.6.B/4.6.C (Cont'd) five gpm, as specified in 3.6.C, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.
Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage j
detection schemes, and if the origin cannot be determined in a reasonably l
short time, the unit should be shut down to allow further investigation l
and corrective action.
l The two gpm limit for coolant leakage rate increases over any 24-hour period is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant j
leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
l REFERENCE
- 1. Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10)
- 2. Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow.
The analysis of the worst overpressure transient, (3-second closure of all main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
Experience in relief valve operation shows that a testing of 50 percent of the valves per cycle is adequate to detect failures or deteriorations.
l The relief valves are benchtested every second operating cycle to ensure l
that their setpoints are within their specified tolerances. The relief l
valves are tested in place in accordance with Specification 1.0.MM to l
establish that they will open and pass steam.
5 l
BFN 3.6/4.6-30 Unit 2 t
. ~ _.- -. -
-... ~ -
1.2/2.2 REACTOR COOLANT SYSTEM INTEGRITY l
SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING l
1.2 Reactor Coolant System Intearity 2.2 Reactor Coolant System Intearity Applicability Anolicability Applies to limits on reactor coolant Applies to trip settings of the system pressure, instruments and devices which are provided'to prevent the reactor system safety limits from being exceeded.
Obiective Obiective
)
To establish a limit below which To define the level of the the integrity of the reactor process variables at which
)
coolant system is not threatened automatic protective action due to an overpressure condition.
is initiated to prevent the pressure safety limit from being exceeded.
1 l
Specifications Snacificatinna l
A.
The pressure at the lowest point The limiting safety system of the reactor vessel shall not settings shall be as specified exceed 1,375 psig whenever below:
)
irradiated fuel is in the reactor vessel.
A.
Verify the safety function i
lift settings of the required S/RVs are within 1 three percent of the setpoint as follows:
Number of Setpoint S/RVs (osia) 4 1105 4
1115 5
1125 Following testing, lift settings shall be within one percent.
Limiting Safety Protective Action Svatem Settina B.
Scram--nuclear 11,055 psig system high pressure BFN 1.2/2.2-1 Unit 3
1.2 BASES REACTOR COOLANT SYSTEM INTEGRITY The safety limits for the reactor coolant system pressure have been selected suca that they are below pressures at which it can be shown that the integrity of the system is not endangered.
However, the pressure safety limits are set high enough such that no foreseeable circumstances can cause the system pressure to rise over these limits. The pressure safety limits are arbitrarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The design pressure (1,250 psig) of the reactor vessel is established such that, when the 10 percent allowance (125 psi, allowed by the ASME Boiler and Pressure Vessel Code Section III for pressure transients is added to the design pressure, a transient pressure limit of 1,375 psig is established.
Correspondingly, the design pressures (1,148 for suction and 1,326 for discharge) of the reactor recirculation system piping are such that, when the 20 percent allowance (230 and 265 psi) allowed by USAS Piping Code, Section B31.1 for pressure transients is added to the design pressures, transient pressure limits of 1,378 and 1,591 psig are established.
- Thus, the pressure safety limit applicable to power operation is established at 1,375 psig (the lowest transient overpressure allowed by the pertinent codes), ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.
The current cycle's safety analysis concerning the most severe abnormal operational transient resulting directly in a reactor coolant system pressure increase is given in the reload licensing report for the current l
)
cycle.
The reactor vessel pressure code limit of 1,375 psig given in j
subsection 4.4 of the safety analysis report is well above the peak l
pressure produced by the everpressure transient described above. Thus, j
the pressure safety limit applicable to power operation is well above the l
peak pressure that can result due to reasonably expected overpressure transients.
Higher design pressures have been established for piping within the reactor coolant system than for the reactor vessel. These increased l
design pressures create a consistent design which assures that, if the pressure within the reactor vessel does not exceed 1,375 psig, the pressures within the piping cannot exceed their respective transient pressure limits due to static and pump heads.
l The safety limit of 1,375 psig actually applies to any point in the reactor vessel; however, because of the static water head, the highest pressure point will occur at the bottom of the vessel.
Because the l
BFN 1.2/2.2-2 l
Unit 3 i
3.6/4.6 BASES 3.6.C/4.6.C (Cont'd) suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.
Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the unit should be shut down to allow further investigation and corrective action.
The two gpm limit for coolant leakage rate increases over any 24-hour period is a limit specified by the NRC (Reference 2).
This limit applies only during the RUN mode to avoid being penalized for the expected coolant leakage increase during pressurization.
The total leakage rate consists of all leakage, identified and unidentified, which flows to the drywell floor drain and equipment drain sumps.
The capacity of the drywell floor sump pump is 50 gpm and the capacity of i
the drywell equipment sump pump is also 50 gpm.
Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.
References i
1.
Nuclear System Leakage Rate Limits (BFNP FSAR Subsection 4.10) 2.
Safety Evaluation Report (SER) on IE Bulletin 82-03 3.6.D/4.6.D Relief Valves To meet the safety basis, 13 relief valves have been installed on the unit with a total capacity of 84.1 percent of nuclear boiler rated steam flow.
The analysis of the worst overpressure transient, (3-second closure of all j
main steam line isolation valves) neglecting the direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed considering 12 valves OPERABLE, results in adequate margin to the code allowable overpressure limit of 1,375 psig.
To meet operational design, the analysis of the plant isolation transient (generator load reject with bypass valve failure to open) shows that 12 of the 13 relief valves limit peak system pressure to a value which is well below the allowed vessel overpressure of 1,375 psig.
Experience in relief valve operation shows that a testing of 50 percent of the valves per cycle is adequate to detect failures or deteriorations.
The l
relief valves are benchtested every second operating cycle to ensure that their setpoints are within their specified tolerances. The relief valves
,l i
are tested in place in accordance with Specification 1.0.MM to establish l
that they will open and pass steam.
BFN 3.6/4.6-30 Unit 3
ENCLOSURE 4 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)
UNITS 1, 2,
AND 3 CURRENT BFN UNIT 2 RELOAD LICENSING REPORT l
l I
t i
l l
4 i