ML20058E293
| ML20058E293 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 07/16/1982 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20058E290 | List: |
| References | |
| NUDOCS 8207280143 | |
| Download: ML20058E293 (37) | |
Text
- -,
ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS (TVA BFNP TS 175)
BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 8207200143 820716 PDR ADOCK 05000259 p
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TABLE 4.1.1 IIACTOR PEDTICTICW SYSTEN (SGAN) INSTRUMENTATION FUNCTIONAL TESTS MINDUM FUNCTIONAL TEST FREQUENCIES FOR SAFETT INSTR. AND CONTROL CIRCUITS Croco f 2)
Functional Test Minimum Frequency (3) i Mode Switch in Shutdown A
Flace Mode Switch in Shutdown Each Refueltog Outage Manual Scram A
Trip Channel and Alarm Every 3 Months IRM High Flut C
Trip Channel and Alaru (4)
Once Fer Veak Nring Refuelin and Before Each Startup Inopera tive C
Trip Channel and Alata (4)
Once Fer Week During Refuelta and Before Each Startup AFLM M
High Fluz (15I scram)
C Trip Output Relays (4)
Before Each Startup and Veek1 When Required to be Operable High Flus E
Trip output Relays (4)
Once/ Week inoperative 2
Trip output Relaya (4)
Once/Veek l
Downscale 5
Trip Output Relays (4)
Once/Veek l
l Flow Bias B
(6)
(6) l l
Righ Reactor Fressure A
Trip Channel and Alarm Once/Honth (1) l l
Righ Drywell Fressure A
Trip Channel and Alsru Once/ Month (1) j Reactor for Vater level A
Trip Chaamel and Alarm Once/ Month (1) i High Vater I.evel in Scram Discharge Tank' A
Trip Channel and Alarm Once/ month Turbine Condenser tw Vacuum A
Trip Channel and Alarm Once/Honth (1)
Main Steas Line R!gh Radiation 5
Trip Channel and Alarm (4)
Once/ Week
NOTES FOR TAHLE 4.1. A 1.
Initially the minimum frequency for the indicated tests shall be once per month.
2.
A description of the three groups in included in the Bases of this npecification.
3.
Tunctional tests are not required when the systems are not required to be operable or are operating (i.e., already tripped). If tests are missed, they shall be performed prior to returning the systems to an operable status.
4.
This instrumentation is exempted from the instrunent channel test definition. This instrument channel functional test will consist of injecting a sinulated electrical signal into the measurement channels.
5.
(deleted) 6.
The 'unctional test of the flow bian network is performed in accordance with Table 4.2.C.
r e
e 39
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RFOUIREMENTS te.6.A Micrmi and Ps esmriration 3.6.A Mysyl and Pressuri tation
{*g7m t a t i on:
Li mi ta t ion; 3.
Test specimens representing the 3
During heatup by non-nuclear reactor vessel, base weld, and means, except when the vessel weld heat affected zone metal shall is vented or as indicated in be installed inthe reactor vessel 3.6.A.4, cooldown following adjacent to the v.essel wall nuclear shutdown on low-level at the core midplane level. The physics tests, the reactor g
g g
g vessel temperatures shall be at or above the temperatures will be in accordance with GE of curve #2 of figure 3.6.1.
report NEDO-10ll5. The specimens shall meet the intent of ASTM E 185-70.
Samples chall 4.
The reactor vessel shell
' be withdrawn at one-fourth and temperatures during inservice hydrostatic or leak testing three-fourths service life.
shall be at or above the 4.
Neutron flux wires which were temperatures shown on curve
- 1 of figure 3.6-1.
The installed adjacent to the reactor applicability af this vessel wall at the core midplane curve to these tests is level were removed during the extended to non-nuclear first refueling outage and heatup and ambient loss tested. The results were used cooldown associated with to more accurately determine the these tests only if the neutron fluence in tl.e vessel heatup and cooldown rates beltline shell at a depth of one-do not exceed 15*F per fourth of the wall thickness.
hour.
TN se determined values of neutron 5.
The reactor vessel head bolting iluence and the methods in studs may be partially Regulat ory (;nide 1.99 were used tennioned (four sequences of to predict the changes in reference the neating pann) psovided the temperature, RTNDT, for each studs and flange materials are effective full power year. Figure above 70*F. Before loading the 3.6-1 shall be updated based on flanges any more, the vessel these predictions until the flange and head flange must be resu om tes & g the specimens greater than 100*F, and must remain above 100*F while under described in 4.6.A.3 are available.
full tension.
5.
When the reactor vessel head bolting 6.
The pump in an idle recircula-studs are tensioned and the reactor tion loop shall not be started is in a cold condition, the reactor unless the temperatures of the vessel shell temperature immediately coolant within the idle and below the head flange shall be per-operating recirculation loops are within 50*F of each other, manentJy recorded.
7.
The reactor recirculation pumps 6.
Prior to and during startuP of an shall not be started unless the idle recirc.ulation loop, the temper-coolant temperatures between the dome and the bottom head ature of the reactor coolant in the drain are within 145'F.
operating and idle loops shall be permanently logged.
7.
Prior to starting a recirculation pump, the reactor coolant tempera-tures in the dome and in the bottom head drain shall be compared and permanently logged.
175
LIMITINC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMINTS 3.6.C cool.s a r Le a k a vf 4.6.C Coolant te*%ste 3.
If the condition in 1 or 2 above cannot be met, an orderly shutdown shall be initiated and the teactor shall be shut.
D. Relief valves down in the Cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1.
At least one safety valve and approxicately one-half of all n.
Relief Valves relief valves shall be bench-checked or replaced with a 3
When more than one relief bench-checked valve each opera-valve or one or more safety ting cycle. All 13 valves (2 valves are known to be safety and 11 relief) will have failed, an orderly shutdown been ch=cked or replaced upcn shall be initiatad and the the conoletion of every accond reactor depressurized to cycle.
less than 105 peig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Once during each operating cycle, each relief valve shall be e.anually opened until thermocouples and acoustic monitors dosmstream of the valve indicate steam is flowing from the valve.
3,The integrity of the relief valve bellows shall be continously monitored when. valves incorporating the belloVs design are installed.
4.
At least one relief valve sh411 be disassembled and inspected each operating cycle.
E.
Jet Pumps E.
Jet rusos i
1.
Vhenever the reactor la in the 1.
Whenever there is recirculation atartue or run modes, all jet flow with the reactor in the pumpe ehell be operable. If startup or run modes with both it is determined that a jet recirculation pumps ru.ning.
pump to inoperable, or if two jet pump operability shall be or more jet pump flow instru-checked daily by verifying that ment f ailures occur and can-the following conditions do not not be corrected within 12 occur simultaneously
(
hourn, an orde:rly shutdovn
}
ahall be inattated and the a.
The two recirculation toeps
}
reactor shall be shutdown in have a flow inhalance of the Cold Condition within 24 15: or more when the pu:ps t.ours.
are operated at the same speed.
l 181 l
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DELETED FIGURE 3.6 2 CHANGE IN CHARPY V TRAfJSITION TEMPERATURE VERSUS NEUTRON EXPOSURE 195 l
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3.6/4.6 Rases 3,6. A/4.6. A Thermal and Pressurization I. imitations The vessel has been analyzed for stresses caused by ther-Al and pressure transients. heating and cooling transients throughout plant life at uniform rates of 1000 F per hour were considered in the temperature range of 100 to
$460 F and were shown to be within the requirements for stress intensity and fatigue litilts of Section III of the ASME Boiler and Pressure Vessel Code (65 Liition including Su:r.=er 1%6 addenda).
Operating limits on the reactor vessel pressure ai i temperature during norr.a1 heatup and cooldown, and during inservice hydrostatic testing, were established using Appendix G of the Stamer 1972 Addenda to Section III of the AS:C Boiler and Pressure Vessel Code, 1971 Edition, as a guide. These operating limits assure that a large postulated surface flaw, having a depth of one-quarter of the material thickness, can be safely accommodated in regions of the vessel shell remote from discontinuities. For the purpose of setting these operating limits the reference temperature, RTUDT, of the vessel caterial was estimated from impact test data taken in accordance with requirements of the Code to which this vessel was designed and n:anufactured (65 Edition to Su=mer 1966 addends.)
The fracture toughness of all ferritic steels gradually and uniforrly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative to account for this in the operation of the P2V. Two types of information are needed in this analysis: 1) A relationship between the change in fracture toughness of the RPV steel and the neutron fluence (integrated neutron flux), and b) a measure of the neutron fluence at the point of interest in the RPV wall.
A relationship between neutron fluence and change in reference temperature, RTNOT, is provided in Regulatory Guide 1.99.
In turn, this change in refer-ence temperature can be related to a change in the temperature ordinate shown in Figure G-2110-1 in Appendix C of Section III of ASME Boiler and Pressure Vessel Code.
The change in reference temperature at any time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Regulatory Guide 1.99.
During the first fuel cycle, only calculated neutron fluence values were used. At the first refueling, neutron dosimeter wires which had been installed adjacent to the vessel wall at the core midplane level were' removed and tested to determine the neutron fluence. Three sets of mechanical test specimens representing the base metal, weld metal and weld heat affected zone have also been placed adjacent to the vessel wall at the core midplane level. These will be removed and tested as required by 10CFR50, Appendix 11.
Until such testing is performed, the changes in reference temperature, RT will be determined based on the results of the testing of the dosimNE r, wires and the methods in Regulatory Guide 1.99.
The operating pressure-temperature 215
+
' 3.6/h.6 BASEG 3.6.A/h.6.A limits shown in Figure 3.6-1 will be adjusted periodically to conservatively nccount for these determined changes in the reference temperature.
As described in paragraph L.2.5 or the carety analysis report, detailed streco annlyces have been made on the reactor vessel for both steady-ctate and transient conditions with respect to material ratigue. The results of these analyses are compared to allovable stress limits.
Requirin6 the coolant temperature in an idle recirculation loop to be 0
within 50 F of the operating loop temerature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.
~
The coolant in the bottom of the vessel is at a lover temperaure than that in the upper regione of the vessel when there is no recirculation flow.
This colder vater is forced up when recirculation pumps are started. This
~:1 vill not result in stresses which exceed ASME Boiler and Pressure Vessel Code,Section III limits when the temperature differential is not greater 0
than lh5 F.
The requirements for full tension boltup of the reactor vessel closure are based on the NDT tcoperature plus 60*F.
This is derived from the requirements of the ASME code to which the vessel was built. The NDT temperature of the closure flanges, adjacent head, and shell material is a maximum of 40*F and a maximum of 10*F for the stud material.
Therefore, the minimum temperature for full tension boltup is 40*F plus 60*F for a total of 100*F.
The partial boltup is restricted to the full loading of eight studs at 70*F, which is stud NDT temperature (10*F) plus 60*F.
The nyytron radiation fluence at the closure flanges is well below 10 avt 1 1 Hev; therefore, radia, tion effects will be minor and will not influence this temperature.
3.6.ll/h.6.5 Enolant Chemintev thu -3,1u in the prim-try cyntea are primarily 30h utainicss nteel and the Zircaluy cladding. The reactor water chemistry linits are establinhed to prevent danac,e to these materials. Limits are placed on cond :ctivlif and chloride measured concentrations. Conductivity lo limited beenuse it in continousP; and cives an indication of ab. normal conditions nr.d the presc::cc o.'
unusual Materials in the coolant. Chloride li;.1its are specified to.eveat f
stress corrosion cracking of stainlesc steel.
m i
216
LIMITING CONDITIC38 FC2 OPZILATION SURVEILLAUCZ REQUIREMENTS
.g 3.7 CONTAINMENT SYST2H8 4.7 CONTAIINENT SYSTM8 B.
Standbv Gao Treatment 8vetem B.
Stanch Gao Treatsient syntem 1.
Except as specified in Specification 1.
At least once per 3.7.B. 3 belcw, all year, the following three trains of the conditions shall be stancly gas treatment dcmonstrated, system a.
Pressure drop across the shall be combined HEPA operable at all times filters and when secondary charcoal containment integrity adsorber banks is required.
is less than 6 inches of water at a flow of 9000 cfm (1
\\
10f.).
b.
The inlet boaters cn each circuit are tento$ in accordance with AUDI M510-1975 and are capable of an output of at least 40 kW.
c.
Air dictribution is unifom within 20%
across HEPA filters and charcoal adsorbers.
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236
!!IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUI.SMENTS
.7 COvrAti! MENT SYSTEtiS 84. 7 COffrAINMENT SYSTEMS E.
Cont rol Room Emergency E.
Control Room Emeroency ventilation Ventilation 1.
Except as specified in specification 1'
At least once per 3.7.E.3 below, both operating cycle, not centrol room to exceed 18 months, emergency the p'ressure drop pressurization across the combined systems HEPA filters and charcoal adsorber banks shall be shall be demonstrated to be operable at all times less than 6 inches of when any reactor water at system vessel contains design ficw rate
(+
irradiated fuel.
10 %).
2.
a.
The results of' 2.
a.
The tests and the in-place sample analysis cold COP and of Specification halogenated 3.7.E.2 shall be hydrocarbon performed at tests at design least once per flows on HEPA operating cycle filters and or once every 18 charcoal
- months, adsorter banks whichever occurs shall show 299%
first for DOP removal and standby service 299% halogenated or af ter every hydrocarbon 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of removal when system operation tested in and following accordance with significant ANSI N510-197 5.
painting, fire or chemical b.
The results of release in any laboratory ventilation zone carbon sample commu nica ting analysis shall with the system.
show 190%
radioactive b.
Cold DOP testing methyl iodide shall be removal at a performed after velocity when each complete or tested in partial accordance with replacement of ANSI N510-1975 the HZPA filter (130*C, 95%
bank or after R. H. ).
any structural maintenance on 244 the system housing.
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TABLE 4.1.1 REACTot F13TECT10W SYSTIN (SGAM) INSTRUKENTATION FmICTIONAL TESTS NIMIE'M TUNCTIONAL TEST TELQUENCIES FOR SAFETT INSTR. AND CONTROL CIRCUITS Croue (2)
Funettonal Test Minimus Frequency (3) l bde Switch in Shutdown A
Place Mode Switch ir. Shutdown Each Refueltog Outage Mannel Scram A
Trip Channel and Alaru Every 3 Months DK High Flux C
Trip Channel and Alaru (4)
Once Fer Week During Refuella and Ecfore Each Startup Inoperative C
Trip Channel and Alaru (4)
Once Fer Week During lefuelta and Eefore Each Startup l
-4 Eigh Flur (15: scram)
C Trip Output Relays (4)
Before Each Startup and Week 1 w
When Required to be Operable i
Eigh Flux B
Trip output Relays (4)
Once/Veck inoperative 2
Trip output Relays (4)
Ouce/Veek Downscale 5
Trip output Relays (4)
Once/Veck Flow Blas B
(6)
(6)
Eigh Reactor Pressure A
Trip Channel and Alara
'Occe/ Month (1)
Eigh Drywell Fressure A
Trip Channel and Alarm once/ Month (1)
Trip bhannel and Alarm Once/Menth (1)
Reactor low Water level A
High Water Level in Scras Discharge Tani A
Trip Channel and Alarm Once/ronth Turbine Condensar low Vacuum A
Trip Channel and Alara Once/Honth (1)
Nata Steam Line Righ Radiation 3
Trip Channel and Alarm (4)
Once/veek
!!OTr$ F_0R TABLE 4.1. A 1.
Initially the minirum frequency for the indicated tests shall be once pe r rnonth.
2.
A description of the three groups is included in the Bases of this specificat.an.
3.
Tunctional tests are not required when the systems are not recuired to be operable or are operating (i.e., already tripped). If tests are missed, they shall be performed prior to returninr, the systems to an operable status, o
I.. This instrumentation is exempted from the instrument channel test definition. This instrument channel functional test will consist of injecting a sinulated electrical signal into the mecsurement channels.
5.
Deleted 6.
The functional test of the flow bias network is perfortned in accordance with Table 4.2.C.
39
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LIMITING COND{TIONS FOR OPERATIONi; SURVEILLANCE REQUIREMENTS 1
bl). A 1he rm t eind Ps e'twr17nt iu s
?
3.6.A h rmn) and Pressuri tation g
t,imit e t a nn:
J,j mi t ra t i ons l
.3.
Test specfc. ens representing the t
3 During heatup by non-nuclear reactor ve'ssel, bane weld, and means, except when the vessel welCheat af fected' zone metal shall is vented or as indicated in be installel intie reactor vessel 3.6.A.4, cooldown following nuclear shutdown on low-level adjacept to the vesse,1 vall physics tests, the reactor at the core midplane level. The vessel temperatures shall be number and type of
'eci'nens at or above the temperatures.
will be in accordan twith/GE of curve #2 of figure 3.6.1.
report NEDO-10115. Jhe specimens shall meet the intent 4.
The reactor vessel shell of ASTM E 185-70. Samples shall temperatures during inservice be withdrawn at one-fourth and
-hydt'ostatic or leak testing three-fourths service life.
shall be at or above the' temperatures shown on curie 4.
Neutron flux wire hich were sw
- L of figure 3.6-1.
The installed adjacent to the reactor applicability of this vessel wall at the core midplane curve to these tests is level were removed during the extended to non-nuclear first refueling cutage and heatup and ambient loss tested. The results were used j
cooldown associated with to more accurately determine the I
these tests only if the neutron fluence in the vessel hentup and cooldown rates j
beltline shell at a depth of one-do nat exceed 15*F per fourth of the wall thickness.
hour.
These determined values of neutron 5.
The re' actor vessel head bolting fluence OAd the metho(a in i
j studo may be, partially i-Regulatory Guide 1.99 were used~
tensianed (four sequences of to pre $ $t the changes in reference the neat any, pann) provided the t empe r9 t ufe. RTNDT, for each I
studu and flange materials are errectjlre full power year. Figure
~
above 70*F. Before loading the
~3.6-1 alIall be urdated based on flanges any more, the vessel flange and head flanr,c must be thgse pqedictions until the from testing We specimeur
&nu a greater than 100*F, and must remain above 100*F while under
'h serihed }n 4.6. A.3 are availableg" 4
< full tension.
./
- s. Wheri the ri sctor vessel head bolting" 6.
The pump in an idle recircula-studs are tensioned and the reactor tion loop shall not be started rjlessthetemperaturesofthe cold condition, the reactor coolent within the idle and vessel shell temperature immediately operating recirculation loops below the head flange shall be per-are within 50*F of each other.
i*anently recorded, t
7.
The reactor recirculation pumps Prio5ftoandduringstartupofan 6.
shall not be started unless the coolant temperatures between idle recti,dulation loop, the temper-the dome and the bottom head tum f the reactor coolant in the drain are within 145'F..
~
P"F"LI"M ""d IdI" 100Ps shall be permanently logned.
~f Prior to starting a recirculaiion pump, the reactor coolant tempera -
tures in the dome and in the bottom head drain shall be compared and V
permanently logged.
175
-4
/
')
LIMITING CONDITIONS F0n OP Q TION SURVEILLANCE REQUIREMENTS 3.6.C cool $at teskava 4.6.C Coolant tenkate 3.
If the condition in 1 or 2 above cannot be cet, an orderly mhutdown shall be initiated and the t'eactor shall be shut.
D. Relief Valves i #"
down in the Cold Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1.
At least one safety valve and
~,
approxiestely one-half of all D.
Relief Valves relief valves shall be bench-checked or re'placed with a
+
1 When more than one relief bench-checked valve each opera-valve or one or more safety ting cycle. All 13 valves (2
'~
valves are kaovn to be safety and'11 relief) vill have
-[~
failed, an orderly shutdown been checked or replaced upon shall be initiatad and the the conolecion of every second e~
reactor depressurized to cycle.
less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Once during each operatin; cycle, each relief valve shall be c.anually opened until thernocouples and acoustic nonitors dovnstrean of the valve indicate steam is flowing from the valve.
3.
The integrity of the relief valve bellows shall be continously cionitored when. valves incorporating the bellous design are installed.
4.
At least one relief valve shall be disassembled and inspected each operating cycle.
t.
Jee'Punos E.
Jet Pumos 1.
Vhenever the reactor is in the 1.
Whenever there is recirculation startup or run nodes, all jet flow with the reactor in the pumps shall be operable. If startup or run modes with both it is de t e rmined that a jet recirculation punps running, pump te (noperable, or if two jet pump operability shall be or more jet purip flow instru-checked daily by verifying that ment failures occur and can-the following conditions co not cot be corrected within 12 occur sicultaneously:
hours, an ora rty shutdown shall be initiatet and the a.
The two recirculation loeps reactor-shall be shutdovn in have a flov te. balance of the Cold Condition within 26 l$I or acre when the yu ps hours.
are operated at the sane speed.
181 l
l L
l DELETED FIGURE 3.5-2 CHANGE IN CHARPY V TRAtJSITION TEMPERATURE VERSUS NEUTR0tJ EXPOSURE 195
3.6/t4.6 HASEs 3.6.A/l.6.A Thermal and Pressurization himitations 4
g The vessel has been analyzed for stresses caused by ther-a.1 and pressure transients. Heating and cooling transients throughout plant life at uniform rates of 1000 F per hour were considered in the temperature range of 100 to
$6" r and were shown to be within the requirements for stress intensity and -
fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (65 Edition including Su=er 1966 addenda).
Operating limits on the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservi m hydrostatic testir_g, were established using Appendix G of the Summer 1972 Addenda to Section III of the AS2 Boiler and Pressure Vessel Code, 1971 Edition, as a guide. These operating limits assure that a large postulated surface flaw, having a depth of one-quarter of the material thickness, can be safely acco=modated in regions of the vessel shell remote from discontinuities. For the purpose of setting these operating linits the reference temperature, RT!!DT, of the vessel caterial was estimated from impact test data taken in accordance witn requirements of the Code to which this vessel was designed and manufactured (65 Edition to Su=mer 1966 addenda.)
The fracture toughness of all ferritic steels gradually and unifor=ly decreases with exposure to fast neutrons above a threshold value, and it is pmdent and conservative to account for this in the operation of the PJV. Two types of information are needed in this ana. lysis: 1) A relationship between the change in fracture toughness of the RPV steel and the neutron fluence (integrated neutron flux), and b) a measure of the neutron fluence at the point of interest in the RFV wall.
A relationship between neutron fluence and change in reference temperature, RTNDT' '" EI "0" "
Y "
- # #"N ence temperature can be related to a change in the temperature ordinate shown in Figure G-2110-1 in Appendix G of Section Ill of ASMh Boiler and Pressure Vessel Code.
The change in reference temperature at any time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Regulatory Guide 1.99.
During the f.Irst fuci cycle, only calculated neutron fluence values were used. At the first refueling, neutron dosimeter wires which had been installed adjacent to the vessel wall at the core midplane level were' removed and tested to determine the neutron fluence. Three sets of mechanical test specimens representing the base metal, weld metal and-weld heat affected zone have also been placed adjacent to the vessel wall at the core midplane level. These will be removed and tested as, required by 10CFR50, Appendix H.
Until such testing is performed, the changes in reference temperature, RT testingofthedosimbr,will be determined based on the resultu of the wires and the methods in Regulatory Guide 1.99.
The operating pressure-temperature 215 9
~
3.6/h.6 BASES i
I.
3.6.A/h.6.A l
l 6
limits shown in Figure 3.6-1 will be adjusted periodically to conservatively account for these determined changes in the reference temperature.
l l
As described in paragraph h.2.5 of the safety analysis report, detailed streGG analyses have been made on the reactor vessel for both steady-state and transient conditions with respect to material fati ue.
The 6
resulta of these analyses are compared to allowable stress limits.
Requiring the coolant temperature in an idle recirculation loop to be within 50 F of the operating loop temerature before a rceirculation pump is started assures that the changes in coolunt temperature at the reactor vessel nozzles and bottom head region are acceptable.
The coolant in the bottom of the vessel is at a lower temperaure than that in the upper regione of the vessel when there is no recirculation flow.
This colder water 13 forced up when recirculation pumps are started. This
-s) vill not result in stresses which exceed ASKE Boiler and Pressurc Vessel Code,Section III limits when the temperature differential is not greater than lh5 F.
0 The requirements for full tension boltup of the reactor vessel closure are based on the liDT temperature plus 60*F.
This is derived from the
~
i requirements of the 'ASME code to which the vessel was built. The NDT temperature of the closure flanges, adjacent head, and shell material is a maximum of 40*F and a maximum of 10*F for the stud material.
Theref ore, the minimum temperature f or f ull tension boltup is 40*F plus 60*F for a total of 100*F. The partial boltup is restricted to the full loading of eight studs at 70*F, which is stud NDT temperature (10*F) plus 60*F.
The nyytron radiation fluence at the closure flanges is well belov 10 nyt 3,1 Mev; therefore, radiation effects will be minor and will not influence this temperature.
- 3. 6.11/ h. 6. B Coolant Chemiutry 1/at.e "i,ls in the primary system are primarily 30h utainicss r.tcel and the Zircaluy clnddinn,. The reactor water chenist'ry limits are establi.h*.d to prevent d2.ar:e to thche materials. Limits are placed on conductivity and chlorlde concentrations. Conductivity Ir. limited beenuse it is contimu,1; menr,ured and c,ives an indication of abnormal conditior.c nr.d the prescucc o.'
unusual naterials in the coolant. Chloride limits are specified to ;.Nye.it stress corrosion cracking of stainless steel.
i 216 4
LIMITING CONDITIC33 FCR OP21MTICN SURVEILLA* CZ 3L",QUIRZIEITS J
\\
i l
3.7 CONTAINMEFF SYSTEMS 4.7 COMPAIUMENT SYSTK48 B.
Standby Gao Treatment 8voteta B.
Stantby Gao Treatment a
System 1.
Except as specified in Specification 1.
At least once per 3.7.B.3 belcw, all year, the following three trains of the conditions shall be standby gas treatment demonstrated.
system a.
Pressure drop acrosa the shall be combined HEPA operable at all times filters and when secondary charcoal containment integrity adsorber banks is required.
is less than 6 inches of water i
at a flow of l
9000 cfm
(~+
\\
l 10%).
b.
The inlet boaters en each circuit are tecto$ in ac::ordance with AUSI N510-1975 and are capable of an output of at lenot 40 kw.
c.
Air diotrit:ution is uniforn within 20".
across HEPA filters and charcoal adsorbers.
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236
L'IMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUI.JMENTS 7 CONTAlilMENT_ SYSTEtiS 4.7 COtTTAINME!1T SYSTEMS E.
{iont rol Room Emergency E.
Control Room Emergency ventilation Ventilation 1.
Except as specified
~
in specification l '.
At least once per 3.7. E. 3 below, both operating cycle, not centrol room to exceed 18 months, emergency the p'ressure drop pressurization
.across the combined systems llEPA filters and charcoal adsorber banks shall be shall be demonstrated to be operable at all times less than 6 inches of when any reactor water at system vessel contains design ficw rate
(+
irradiated fuel.
10 %).
2.
a.
The results of' 2.
a.
The tests and the in-place sample analysis cold COP and of Specification halogenated 3.7.E.2 shall be hydrocarbon performed at tests at design least once per flows on llEPA operating cycle filters and or once every 18 charcoal
- months, adsorter banks whichever occurs shall show 299%
first for DOF removal and standby service 299% halogenated or af ter every hydrocarbon 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of removal when system operation tested in and following accordance with significant ANSI N$10_-1975.
painting, fire or chemical b.
The results of release in any laboratory ventilation zone carbon sample communicating analysis shall with the system.
show 290%
radioactive b.
Cold DOP testing methyl iodide shall be removal at a performed after velocity when each complete or i
tested in partial accordance with replacement of AtlSI N510-1975 the H2PA filter (1300C, 95%
bank or after R.H.).
any structural maintenance on 244 the system housing.
i i
l
9 9
9 9
I UNIT 3
TABLE e.1.A-REACT 0E PICTECTI0tf SYSTEM (SCRAM) INST 1tUMEt." RATION FUNCTIONAL TESTS MINIMUM FUNCTIChAl TEST FEEQUENCIES FOR SAFETY INSTR. At;O CONTROL C1.9 CULTS Group (2)
Functional Test Minimum Frequency (3)
Made Switch in Shutdown A,
Place Mode Switch in Shutdown Each Eefueling Outage Maniaal' Scram A
Trip Channel and Alarm Every 3 Months IRM Htqh Flux C
Trip Channel and Alarm (e)
Once Per Week During Refueling and Before Each Startup Inoperative C
Trip Channel and Alarm (e)
Once Per Week During aefueling and Before Each Startup APRM High Flux (155 scram)
C Trip Output Pelays (4)
Before Each Startup and Weekly Wher. Fuquared to be Operable Nigh Flux B
Trip output Relays (4)
Cnce/ Week,
g Inope ra tive 8
Trip output Relays (e)
Once/ Week Downscale C
Trip Output Relays (e)
Oncerweek Flow Bias 8
(6)
( 6) i High Peactor Pressare A
Trip Channel and Alarm once/ Month (1)
High Drywell Pressure A
Trip Channel and Alarm once/Manth (1)
^
Feactor Low Wat er Level A
Trip Channel and Alarm Once/ Month (1) fligh Water level in Scram Discharge Tank A
Trip Channel and Alarm Once/!fonth T*Jr bine COndenMer low Vacuum A
. Trip Channel and Alarm Once/ Mon th ( 1)
I i
l I
e o
ty n.'E S FOR TABLE is.1.A Initially the minimum f requency for the indicated tests shall 1.
be once per month.
m A description of the three groups le included in the Bases of 2.
this specification.
Functional tests are not required when the systems are not 3.
(i. e., already required to be operable or are operatingthey shall be performed If tests are missed, tripped).
prior to returning the systems to an operable status.
J This instrumentation is exempted from the' instrument channel 84.
This instrument channel functional test test definition.
will consist of injecting a simulated electrical signal into the measurement channels.
5.
(deleted)
The f unctional test of the flow bias network is performed in 6.
accordance with Table 4.2.c.
e 38
)
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l A-._.....-
LIMITING CONDITIONS FOR 01'ERATION f
SURVEILIANCE REQUIREMENTS
- 3. 6 PRIMARY SYSTEM BOUNDARY 18. 6 PRIMARY SYSTEM BCUNDARY 4
The reactor vessel 4.
Neutron flux wires which were shell temperatures installed adjacent to the reactor during inservice vessel wall at the core midplane hydrostatic or leak level were removed during the first testing shall be at refueling outage and tested. The or above the results were used to more accurate-temperatures shown on iy determine the neutron fluence curve Number 1 of in the vessel beltline shell at a figure 3.6-1.
The depth of one-fourth of the wall applicability of this thickness. These determined values curve to these tests of neutron fluence and the methods is extended to non-in Regulatory Guide 1.99 were used nuclear heatup and ambient loss cool-to predict the changes in reference down associated temperature, RTNDT, for each with these tests ef fective full power year.
only if the heatup Figure 3.6-1 shall be updated and cooldown rates do based on Gese pMWms mU not exceed 15*F per the results from testing the hour.
specimens described in 4.6.A.3 are 5.
The reactor vesset head available, bolting studs may be 5.
When the reactor vessel head bolting partially tensioned studs are tensioned and the reactor (four sequences of the seating pass) provided is in a Cold Condition, the reactor the studs and flange vessel shell temperature immediately materials are above below the head flange shall be 70*F. Before loading the permanently recorded.
flanges any more, the vessel flange and head.
6.
Prio to and during startup of flange must he r,reator an idle recirculation loop, the than 200*F. and mu;t t empe r.it ure 01 the reactor coolant remain above 100*F while under full tension.
in tin' operat ing and idle loops shall be permanently logged.
6.
The purnp in an idle recirculation loop 7.
Prior to starting a recirculation shall not *e started pump, the reactor coolant unicou ti temperatures in the dome and in the
$1 the bottom head drain shall be ar t th idle and operating c mpared and permanently recirculation loopo logged.
are within 50*r of each other,.
7 The reactor recirculation pompa chall not be started unless the coolant t ':m pe ra t u r e r. between the dome and bot tom head drain are within Ua 5
- r.
186
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIRDENTS
- 1. <. Ep 1 M Ave y systrM RcVNDARY
- 4. 6 PRfMARY SYSTrM BOUNDARY o.
Relief Valves D.
Relief Valves 1.
1 When more than one relief valve approximately one-is known to -
half of all relief be f ailed, an orderly valves shall te shutdown shall be bench-checked or initiated and the replaced with a reactor depressurized bench-checked valve to less than 105 psig each operating cycle.
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
All 13 valves will have been checked or replaced upon the completion of every second cycle.
2.
Once durine, each operating cycle, each relief valve shall be manually opened until thermocouples and acoustic monitors downstream of the valve indicate steam is flowing from the valve.
3.
The integrity of the relief valve bellows shall be continously monitored when. valves incorporating the belloks design are installed.
4 At least one relief v al v e shall be disassembled and inspected each operating cycle.
192
e 9
e DELETED FIGURE 3.6-2 CHANGE IN CHARPY V TR ANSITION TEMPERATURE VERSUS NEdTRON EXPOSURE 208
e-1.6/4.6 BASES 3.6.A/4.6.A Thermal and Pressurization Limitations The vessel has been analyzed for stresses caused by thermal and pressure transients.
Heating and cooling transients throughout plant life at uniform rates of 1000F per hour were considered in the temperature range of 100 to 5460F and were shown to be within the requirements for stress intensity and fatique limits of Section III of the ASME Boiler and Pressure vessel Code (65 Edition including Summer 1966 addenda).
Operating limits on the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservice hydrostatic testing, were established using Appendix G of the Summer 1972 Addenda to Section III of the ASME Boiler and Pressure Vessel Code, 1971 Edition, as a guide.
These operating limits assure that a large postu~ated surface flaw, having a depth of one quarter of the material thickness, can be safely accommodated in regions of the vessi shell remote from discontinuities.
For the purpose of setting these operating limits the reference temperature, RTN0T, of the vessel material was estimated from impact test data taken in accordance with requirements of the Code to which this ressel was designed and manufactured (65 Edition to Summer 1966 addex.da. ).
The f racture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative to account for this in the operation of the RPV.
Two types of information are needed in this analysis:
- 1) A relationship between the change in fracture toughness of the RPV steel and the neutron fluence (integrated neutron flux), and b) a measure of the neutron fluence at the point of interest in the RPV wall.
i A relationship heLween neutron fluence and change in reference i
temperature, RTNDT, is provided in Regulatory Guide 1.99.
In turn, this change in reference temperature can be related to a change in the temperature ordinate shown in Figure G-2110-1 in Appendix G of Section III of the ASME Boiler and Pressure Vessel Code.
The change in reference temperature at any time period can be determined from the thermal power output of the plant and its relation to the neutron fluence and from Regulatory Guide 1.99.
During the first fuel cycle, only calculated neutron fluence values were used. At I
the first refueling, neutron dosimeter wires which had been installed l
adjacent to the vessel wall at the core midplane level were removed and tested to determine the neutron fluence. Three sets of mechanical test specimens representing the base metal, weld metal and weld heat affected zone have also been placed adjacent to the vessel wall at the core midplano level. These will be removed and tested as required by 10 CFR 50, Appendix 11.
Until such testing is performed, the changes in 220 i
h
3.6/4.6 BASES reference temperature, RTNDT, will be determined based on the results of the testing of the dosimeter wires and the methods in Regulatory Guide 1.99.
The operating pressure-temperature limits shown in Figure 3.6-1 will be adjusted periodically to conservatively account for these determined changes in the reference temperature.
As described in paragraph 4.2.5 of the saf ety analysis report, detailed stress analyses have been made on the reactor vessel for both steady-state and transient conditions with respect to material fatigue. The results of these analyses are compared to allowable stress limits.
Requiring the coolant temperature in an idle recirculation loop to be within 50*F of the operating loop temperature before a recirculation pump is started assures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.
The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow.
This colder water is forced up when recirculation pumps are started.
This will not result in stresses which exceed ASME Doller and Pressure Vessel Code,Section III limits when the temperature dif f erential is not greater than 1450P.
The requirements for full tension boltup of the reactor vessel closure are based on the NDT temperature plus 60*F.
This is derived from the requirements of the ASME code to which the vessel was built. The NDT t uperature of the closure flanges, adjacent head, and shell material is a ma.ximum of 40*F and a maximum of 10*F for the stud material.
Therefore, the minimum temperature for full tension boltup is 40*F plus 60*F for a total of 100*F.
The partial boltup is restricted to the full loading of eight studs at 70*F, which is stud NDT temperature the closure (10*F) plus 60'F.
The n g tron radiation fluence at flanges is well below 10 nyt > 1 Mev; therefore, radiation effects will be minor and will not influence this temperature.
221 l
L.
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
.s
- 3. 7 CONTAINMENT SYSTEg 4.7 CONTAINMANT SYSTM8 B.
Standby Gas Treatment System B.
Standby Gas Treatment System 1.
Except as specified in Specification 1.
At least once per 3.7.B.3 below, all year, the following three trains of the conditions shall be standby gas treatment demonstrated, system a.
Pressure drop across the shall be combined HEPA operable at all times filters and when secondary charcoal containment integrity adsorber banks is required.
is less than 6 inches of water at a flow of 9000 cfm fi 105).
\\
b.
The inlet heaters on each circuit are testes in accordance with An8I H510-1975 and are capable of an 5
output of at least 40 kW.
c.
Air distribution is uniform within 20%
across HEPA filters and charcoal adsorbers.
247
1 I
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS E.
Control Foom Emergency E.
Control Room Emergency Ventilation Ventilati,gn 1.
Except as specified in specification 1.
At least once per 3.7.E.3 below, both operating cycle, not centrol room to exceed 18 months, the pressure drop emergency pressurization across the combined systems HEPA filters and charcoal adsorber banks shall be shall be demonstrated to be operable at all times less than 6 inches of when any reactor water at system vessel contains design flow rate (1 irradiated fuel.
10%).
2.
a.
The results of 2.
a.
The tests and the in-place sample analysis cold DOP and of Specification halogenated 3.7.E.2 shall be hydrocarbon performed at tests at design least once per flows on HEPA operating cycle filters and or once every 18 charcoal
- months, adsorter banks whichever occurs shall show 299%
first for DOF removal and standby service 299% halogenated or after every hydrocarbon 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of removal when system operation tested in and following accordance with significant ANSI N510-197 5.
painting, fire or chemical b.
The results of release in any laboratory ventilation zone carbon sample communicating analysis shall with the system.
show 290%
radioactive b.
Cold DOP testing methyl iodide shall be removal at a performed after velocity when each complete or tested in partial accordance with replacement of ANSI N510-1975 the HEPA filter (1300C, 951 bank or after R.H.).
any structural maintenance on 256 the system housing.
l
A ENCLOSURE 2 DESCRIPTION AND JUSTIFICATION (TVA BFNP TS 715)
BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 1.
Pages 35 and 37 - Units 1 and 2 Pages 36 and 38 - Unit 3 Description The proposed changes remove the requirement to perturb the water level in the reactor vessel and monitor the water level indicator changes after performing the monthly functional test. The functional test will continue to be performed monthly as required by technical specification Table 4.1.A.
Reason /Justriciation The water level instrumentation is valved out during performance of the monthly functional test. After completion of that test the level instrument is valved back into service. That instrument then indicates the reactor vessel water level. This indication can be compared with the numerous other water level instruments for verification that the instrument has indeed been returned to service.
Perturbing the reactor water level is an operational inconvenience to the plant staff. We are not aware of any regulatory requirement or recommendation to perturb the water level. The BWR Standard Technical Specifications do not require it.
The FSAR states that for any sensor that is valved-out or otherwise removed from service during testing, positive indication is obtained that the sensor has been returned to service and will see changes in the process variable. Indication of the reactor vessel water level after the instrument is valved back into service and general agreement with the other instrumentation as discussed above demonstrates compliance with the FSAR statement. Additionally, removal of this requirement will not adversely affect the operation, safety margins, accident analysis, or overall safety of the plant.
, 2.
Page 175 - Unita 1 and 2 Page 186 - Unit 3 Description This change only updates technical specification 4.6.A.4 to reflect the present status of the neutron flux wires.
Reason These wires were installed for purposes of experimentally verifying the calculated values of neutron fluence. This proposed change reflects that the wires were removed and the results used to determine the neutron fluence. The new proposed specification outlines planned actions regarding future revisions to the technical specifications concerning reference temperature RTNDT*
No additional justification is needed.
3 Page 181 - Units 1 and 2 Page 192 - Unit 3 Description This proposed change revises the surveillance requirement concerning monitoring of relief valve bellows to clarify that the bellows will be monitored when valves incorporating the bellows design are installed. For unit 3 this surveillance requirement had been removed but is now proposed to be added back to have all three units consistent in technical specification requirements.
Reason / Justification All bellows type relief and safety valves have been changed out to nonbellows two-stage Target Rock valves. This proposed change only clarifies requirement for testing the integrity of bellows if spare valves with bellows are installed. The test requirement is not applicable unless a bellows type valve were to be reinstalled.
(9
. 4.
Page 195 - Units 1 and 2 Page 208 - Unit 3 Description Figure 3.6-2 " Change in Charpy V Transition Temperature versus Neutron Exposure" is to be deleted.
Reason / Justification Figure 3.6-2 as shown in the technical specifications is not consistent with Regulatory Guide 1 99. We do not consider it appropriate to have this figure in the technical specifications. As necessary in the future, the regulatory guide can be used in lieu of this Figure 3 6-2.
5.
Pages 215 and 216 - Units 1 and 2 Pages 220 and 221 - Unit 3 Description These proposed changes revise the BASES to more accurately reflect current industry practices regarding determination of changes in reference temperature RT The change updates the specification BASES to reflect what has be.en done with neutron dosimeter wires ND that were installed adjacent to the reactor vessel wall. It also describes what will be done with mechanical test specimens. It describes our future plans for determining changes in reference temperature RTNDT*
i
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6 Page s236 Page and
, 4_
s 247 244 ad n
256 Uni
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J 6.
Pages 236 and 244 - Units 1 and 2 Pages 247 and 256 - Unit 3 Description l
The proposed changes remove the specific references to the diesel generators required for operation of the Standby Gas Treatment System (SGTS) and the Control Room Emergency Ventilation from these sections.
Reason / Justification These proposed changes are for clarity only. Rather than addressing diesels in the section Limiting Condition for Operation, it has been placed in the Definitions section per NRC request.
(Reference letter from D. G. Eisenhut to All Power Reactor Licensees dated April 10, 1980.) The requirement for operability of the backup power supply to the SGTS and Control Room Emergency Ventilation is now addressed by technical specification definitions 1.C.2 and 1.E.
Therefore, these requirements do not need to be addressed separately in the LOO. This proposed change does not affect the safe operation of the plant.
-