ML20078B718

Proposed Tech Spec Changes Re pressure-temp Limit Curves

ML20078B718
Person / Time
Site:	Browns Ferry
Issue date:	09/22/1983
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20078B704	List: ML20078B703 ML20078B718
References
NUDOCS 8309270219
Download: ML20078B718 (24)

Text

?

9 ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION REVISIONS BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, A'ND 3 (TVA BFNP TS 191) i' 4:

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t 8309270219 830922 PDR ADOCK 05000259 PDR p .

-- ,,w, . . . , - . . . - -,--.,...-..%,m, - . . . , .wo---, , . - , .w-p,, , - - _. . --c-,-w,cy. ,,,- ,,,--~,.g,,_m, -,we - . ,.w7,,7,,,.,, -,--vw

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PROPOSED CHANGES UNIT 1 1

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. . . _-. _ . - - - _ _ _ . . _ . _ . _ ~ _ _ _ .. _ _ _ _ _ . - -, , . _ - - . - - _ _ . - _

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.A Therrial and Prer.surization .- 3..G.A ]),igrg,_neet,,1;, em:rt ret ty

, 1, int tat ions, I 8 "d'. L si'182,

" ~

3. Test specimens representing the 3.- During heatup by non- reactor vessel, base weld, and nuclear means,except .

weld heat affected zone metal shall when the vessel is' be installed inthe reactor vessel

, vented or as indicated adjacent to the vpssel wall in 3.6.A.4, during at the core midplane level. The cooldown following number and type of specimens nuclear shutdown, or will be in accordance with GE during low-level physics report NED0-10115. The tests, the reactor specimens shall meet the intent veiset temperature shall of ASTM E 185-70. Sampics shall be at or above the tem- , ' be withdrawn at one-fourth c.nd pciratures of curve #2 of. Figure 3.6.2 until three-fourths service life.

removing tension on the 4. Neutron flux wires which were

, head stud bolts as inntn11cd adjacent to the reactor specified lui 3.6.A.5. vessel wall at the core midplaae

4. The reactor vessel shell level were removed during the temperatures during inservice first refueling outage and hydrostatic or leak, testing tested. The results were used shall be at or above the to more accuratel'y determine the temperatures shown on curve & d

1. 1 of figure 3.6-1. The beltline shell at a depth of one-applicability of this fourth of the wall thickness.

curve to these tests is These determined values of neutron extended to non-nuclear fluen'ce and the methods in heatup and ambient loss Regulatory Guide 1.99 were used cooldown associated with to predict the changes in reference these tests only if the heatup and cooldown rates temperature, RTNDT. After testing do not exceed 15'F per the specimens described in 4.6.A.3, hour. Figure 3.6-1 shall be updated if the test results indicate an update is i

5. The reactor vessel head bolting necessary.

. studs may be partially tensioned (four serguences of itic ne.u tun pman) ps .vided the studu and flange materials are e, . When tin reactor v'ssel e head bolting above 70*F. Before loading the suds asu u.nsio M M N m m flanges any more, the vessel is in a culd condition, the reactor flange and head flange must-be vessel shell temperature immediate1 Y y.rcater than 100*F and must beim the head flange shall be per-remain above 100'F while under manently recorded.

full tensiun.

6. Prior to and during startup of an

6. The pump in an idle recircula-idle recirdulation loop, the temper-tion loop shall not be started .

unless the temperatures of the ature'of the reactor coolant in the coolant within the idle and operating and idle Inops shall be operating recirculation loops permanently longed.

are within 50*F of each other.

7. Prior to starting a recirculation

7. The reactor recirculation pumps shall not be started unless the pump, the reactor coolant tempera-coolant temperatures between t in & de M h m um the done and the bottom head head drain shall bu compared and drain are within,145*F. P" F"d'W 3Y 30830d*

175 e

p w e -~y--

8/24/83 Figure 3 6-1 1200 - 1 2 1 Curve #1 i ---

Minimum temperature

' '~~ '~~

i for pressure tests

' . L __ ,____

such as required by j -----

~'

~~~

Section XI. Minimum 1000 -

i f v

. temperature of 170oF is required for test l

l[

i pressure of 1,100 psig.

j.._,.___ Curve 02

-~~T ~~~ Minimum temperature for is!

i mechanical heatup or h 800 - cooldown following

= '

-ffr, ~-glI

n. ' ' nuclear shutdown.

g _ i j

! I i

Curve #3 3

g

', j j '

Minimum temperature for 2

', j r j core operation "g 600 - i .

(criticality) includes j--.tj MG

, i i  ! additional margin

!lL-4f r

l, i required by 10CFR50, d

i i Appendix 0, g / l l Par. IV.A.3 which N

--~~f.--y i became effective '

O 400 - i i July 26, 1983.

I t3

__, r' Notes a ..._ {_---'

These curves include

{-..

-l  !

sufficient margin to provide protection 200 - ,

/ / against feedwater nc% '-

I

! Y ~~I~/ degradation. The ctr.

- , j ,j i

allow for shifts in RTt'a-of the reactor vessel 4# ' '

ll' f~ N -

30El-%MD:PELUifiE l  ;

beltline materials to com-0- "+W '

i ' '

pensate for radiation 0 100 200 embrittlement for the 300 400 life of the plant.

MINDIUM TDIPERATURE

( F) 194 I

-- 6 0

0

• g e

e e*

e DELETED FIGURE 3.5 2 CHANGE IN CHARPY V TRArJSITION TEMPERATURE VERSUS NEUTRON EXPOSURE f

G G

e 195 O

I .

h i e

i l

1

( -- - -

e 3.6h.6 neEs

' 3.6.A/Is.6. A Thermal and Pressurtsation I.tattations The vessel has been anal'ysed for stessses caused by ther.-Al and pressure transients. Heating and cooling transients throughout plant life at unifor:s rates of 1000 F per hour were considered in the temperature range of 100 to

$6' Y and were shown to be within the requirc:nents for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code (65 Edition including Suazer 1966 addenda).

Operating lir.its on the reactor vessel pressure and te:aperature during nor:::a1 heatup and cooldown, and during inserrim hydrostatic testira, were establ.ished using Appendix G of the Summer 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 sett.irg these operating limits the reference temperature, RTi!DT, of the vessel caterir.1 was estimated from impact test data taken in accordance with require:ner.ts of the Code to which this vessel was designed and manufactured (65 Edition to Su=ner 1966 addenda.)

i The fracture toughness of al.1 territic steels gradually and uniformly decreases with exposure to fast neutrons above a threshold value, and it is pr.uient and conservative to account for this in the operation of the PJV. 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 relaLionship between neutron fluence and change in reference temperature, RTNDT, a pr e n egu a ry u t.H. n. turn, t s change in re h ence temperature enn be related to a chi.nge in the temperature ordinate shown in Figure C-2110-1 in Appendix G of Section III of ASHE Boiler and Pressure Vesset Code.

I The change in reference temperature at any time period can be determlued from the thermal power output of the plant and its relation to the neutron fluence nnd 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 IcVel. These will be removed and tested as, required by 10CFR50, Appendix !!. l!ntil such testing is performed, the changes in reference temperature, RT he will baned on the resulta of the testing of the dosimN,,,r, e wires

- and the methods in Regulatory Culdo 1 99. The operating pressure-temperature P

215

,.+.,.-y-- e-- . - - . . , ,m.7,~,y v, -.

w%-. -.r,w. . , - - , , sw-.-,-.,_----~+--,-.e c.. - - _ v+--, ----1.m+ --,--r. ,

. ' ' " . * 3.'6/h.6 Basts-3.6.A/h.6.A ,, , _ , _ , , , , , ,

limits shown in Figure 3.6-1 will be adjusted if necessary when the test results for the mechanical test specimens are available.

i As described in paragraph b.2.$ of the safety analysis report, detailed stress analyses have been made on the reactor vessel for both steady-state and transient conditions with respect to material rati6 ue. The results of these analyses are compared to allovable stress limits.

Requiring the coolant temperature in an idle recirculation loop to be within 50 F of the operating loop temerature before a recirculation pump is started assures that the chances in coolant temperature at

• the reactor vessel nozzles and bottom head region are acceptabic.

The coolant in the bottom of the vessel is at a lover temperaure than that in the upper regions of the vessel when there is no recirculation flow.

s This colder water is forced up when re:1rculation pumps are started. This

1 vill not result in stresses' which exceed ASME Boiler and Pressurc Vessel Code,Section III limits when the temperature differential is not greater than Ib50F.

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 requitcments of the ASHE 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 ainLmum temperature for full tension boltup is 40*F plud 60*F for a total of 100*F. The partial boltup is restricted to thn full loading of eight studs at 70*F, which is stud NDT temperature (10*F) plus 60*F. Thengytronradiationfluenceattheclosure flanges is well below 10 nyt > 1 Hev; therefore, radiscion effects

. will be minor and will not influence this temperature.

3. 6. h/Is . 6. 5 @nl a n t,,Cht;mi nt. r,v, lhte:.iile in the prim.ary systea are primarily 30I stain 1 css r.tect and the

~

Zirenluy cladding. The reactor water chemistry linits are establi.-hed to prevent danar.c to these materials. Limits are placed on condu:ti'/Ity and chloride concentrations. Conductivity 16 limited beenuse it is contin.Susly mensured,and cives nn indication of nhnermal conditione nr.d the presence c.'

unusuni natorints in the coolant. Chloride limits are specified to preveat stresu corrosion cracking of stainicas stec1. .

l

. . . . 216

_.L-9 0 -

PROPOSED CHANGES UNIT 2 i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.A Therrial and Prer.surization . Is.6.A 1heaval ar=1 P essnriratto.. _

l. int ta tions . 3lautaL8"n:

3. Test specimens representing the

3. During heatup by non- reactor vessel, base weld, and nuclear means,except . weld heat affected sone metal shall when the vessel is'

~

be installed inthe reactor vessel vented or as indicated adjacent to the v.essel wall in 3.6.A.4, during at the core midplane level. The  ;

cooldown following number and type of specimens nuclear shutdown, or

will be in accordance with CE l

, during low-level physics report NEDO-10115. The '

tests, the reactor specimens shall meet the intent vessel temperature shall . of ASTM E 185-70. Samples shall I be at or above the tem- , ' ba withdrawn at one-fourth and peratures of curve #2 three-fourths service life.

of Figure 3.6.1 until ,

removing tension on the 4. Neutron flux wires which were

• l

. head stud bolts as installed adjacent to the reactor

. specified in 3.6.A.S. vessel wall at the core midplanc 1 3cyc1 were removed during the '

4. The reactor vessel shall first refueling outage and temperatures during inservice hydrostatic or leak testing tested. The results were used shall be at or above the to more accuratel'y determine the I temperatures shown on curve neutron fluence in the vessel )

1. 1 of figure 3.6-1. The beltline shell at a depth of one-applicability of this fourth of the wall thickness. l curve to these tests is These determined values of neutron extended to non-nuclear fluence and the methods in

• i heatup and ambient loss Regulatory Guide 1.99 were used cooldown associated with to predict the changes in reference these tests only if the temperature, RTNDT. After testing heatup and cooldown rates the specimens described in 4.6.A.3, do not exceed 15'F per

, Figure 3.6-1 shall be updated if the

, hour.

test results indicate an update is

5. The reactor vessel head bolting necessary.

. studs may be partially tensicaed (four sequences of i t;. en*.it lur. Pmm) Ps e.vided the 5. When the reactor vessel head boltiny u.udu and flange materials are above 70 F. Befure loading the studs are tensioned and the reactor flanges any more, the vessel is in a cold condition, the reactor flange and head flange must=be vessel shell temperature immediately greater than 100*F, and must below the head flange shall be per-remain above 100*F whLle under manently recorded, full tension.

6. TSs pump in an idle recircula- 6. Prior to and during startup of an tion loop shall not be started ,

idle recirdulation loop, the temper-untess the temperatures of the ature of the reactor coolant in the i crolant within the idle ar.d operating and idle Inops shall be '

eperating recirculation loops pera:nient ly lonned.

.are within 50*F of each other.

7. Pri r to starting a recirculation

7. The reactor recirculation puisps shall not be started unless the pump, the reactor coolant tempera-coolant temperatures between tures in the dome and in the bottom

,the done and the bottom head head drain shall be compared and drain are within.145'F. pearmanently logged.

, 175

8/24/83 Figure 3.6-1 1200 1  ? 1 Curve #1 L+ , ----

Minimum temperature f i i

' ~

~~~'~~

for pressure tests

' L such as required by j __]___, ----

~~~' Section XI. Minimum T.

1000 -

1 f . temperature of 1700F is required for test pressure of 1,100 psig.

' j. _.___. Curve #2 i

ist Minimum temperature for

~

mechanical heatup or h 800 - '

' J! couldown following A

.l ---

nuclear shutdown.

o  !  !  !' t a iJ Curve #3

[

3 .j Minimum temperature for g 1 i core operation 2

~g 600 - i '

(criticality) includes

,' ij l jl , additional margir.

pG  !

ift--Jf required by 10CFR50,

@$ _ i i Appendix G, y / l 2 I Par. IV.A.3 which

'l became effective '

a: 400 -

~ -

/ i July 26, 1983.

U ,

if____ rl Notes h - f These curves include

~~ ~

..g_-Z-['

1-.-

+- "-

sufficient margin to r

' provide protection 200 - ,

against feedwater nozzle 1 '

/ degradation. The curves I

l, j , .j], l allow for shifts in RTNDT JI!

i i

i

' of the reactor vessel

!,,' y[- _ , JOL'f-UPAc r wu ihu. .'

beltline materials to com-

' ' pensate for radiation O- - * - + - -! r I- *44 -+  !' '!' '

0

. t embrittlement for the 100 200 300 400 life of the plant.

MIND!UM TDIPERATURE

( F) 194

~

..-_ - . -_. __ = - - - - . . _ _ _ - . . . - . . . . . . . - . . -- _,

• * . s l g

i l

l l

DELETED

I I

i i l

I 1

I i

FIGURE 3.6 2 ,

CHANGE IN CHARPY V TRANSITION TEMPERATURE VERSUS 1 NEUTRON EXPOSURE ,

i l

l i .

l I e e

e 195 D

. I l

- ' ' ' ' * * = ----w,_.-..._ , _ _ . , , _ _

. ~ . . - - - _ . - . ._

. l 3.6/4.6 BASES 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 546' F 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 Sun:mer 1966 addenda).

Operating limits on the reactor vessel pressure and temperature during normal heatup and cooldown, and during inservice hydrostatic testir.g, were established using Appendix G of the Summer 1972 Addenda to Section III of the ASME Boiler and Pressure Yessel Code, 1971 Edition, as a guide. These operating limits assure that a large postulated surface flaw, having a depth of one-qusrter of the material thickness, can be safely accommodated in regions of the vessel ,

shell remote from discontinuities. For the purpose of settirg these operating limits the referer.ce temperature, RTilDT, of the vessel caterial was estinated from impact test data taken in accordance with requirements of the Code to which this vessel was designed and manufactured (65 Editf*on to Surmer 1966

- addenda.)

The fracture toughness of all ferritic steels gradual.1,y and uniformly decreases with exposure to fast neutrons above a threshold value, and it is pr.: dent and conservative to account for this in the operation of the RFV. Two types of information are needed in this analysis: 1) A relationship between the change in fracture toughr.ess 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.

RTNirf' is provided in Regulatory Guide 1.99. In turn, this change in refer-ence temperature enn be related to a change in the temperature ordinate shown in Figure C-2110-1 in Appendix C of Section III of ASME Boiler and Pressure t 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 f3uence values were used. At the first refueling, neutron dosimeter wires which had bet.n installed adjacent to the vessel wall at the core midplane-level were' removed and tested to determine the neutron fluence. Three sets of mech:.nical 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 icvel. These will be 1

~

removed and tested as required by 10CFR50, Appendix,11. Until such w

testing is performd, the changes in reference temperature, RT be basedontheresultsofthetestingofthedosimeW.7,will er wires and the methods in deguintory Guide 1.99. The operating pressure-temperature 215 l

- e- 43- --. ++ygpy- - gy-- y -ap,

-~ -. --- - . .. _ ._

s

,e y, .

• 3.6/h.6 BASES ,

3.6.A/k.6.A _,,,,,~.~.~..-...,..,

H limits shown in Figure 3.6-1 will be adjusted if necessary when the test results for the mechanical test specimens are available.

As described in paragraph h.2 5 of the safety analysis report, detailed strecs annlyses have been .,ade on the reactor vesset for both steady-state and transient conditions with respect to materia 1 fati6ue. The results of these analyses are compared to allowable stress limits.

Requirin5 the coolant temperature in an idle recirculation loop to be 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 acceptabic.

, 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 is forced up when recirculation pumps are started. This

1 vill not result in stresses which exceed ASME Boiler and Pressure Vessei Code,Section III limits when the temperature differential is not greater than 145*F.

The requirer,nts for full tens!.on boltup of the reactor vessel closure are based te the NDT temperature plus 60*F. This is derived from the requicements 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*T and a maximum of 10*F foe the stud esterial.

Therefore, the minimum temperature for full cension boltup is 40*F plud 60*F for a total of 100*T. The partial boltup is restricted *to the full loading of eight studs at 70*T, which is stud NDT temperature (10*F) plus 60*F. The ngytron radiation fluence at the closure flanges is well below 10 nyt 11 Hev; therefore, radia,cion effects

. vill be minor and will not influence this temperature.

3.6.D/h.6.5 gnian t,,cig<pi nt e y, VnL*::*htu in the prinary system are primarily 30h utainicss r.Lec1 and the

'41renluy cladding. The reactor water chemistry linits are estnblinhed-to prevent danar.c to these materials. Limits are pieced on conductiv'L/ nnd chloride concentrations. Conductivity lu limited beenuse it is contin.1u-1; mensured,and cives an indication of nh.mormal conditior.o ar.d the presence o.'

unusual nuterials in the coolant. Chloride limits are specified to preveat ,

stresu corrosion cracking of stainicss steel. .

I

. . . . 21c I

l

_ - . . . - - ~ _ - ~ . - _ . -

& ,. - -- - - E - --- 4 a O

O PROPOSED CHANGES UNIT 3

}

1 e

t l

l

- _ - . . _ _ ___ . . _ _ . __ ._ _._. . ~ .. _. _ _ _ _ _ _.__ _ . . _ - _ _ . _ . _ _ _ . _ -

f.lMITING CONN 1710NS FOR Ol'ERATION SURVEILIRICE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM 8CUNDARY

i. . 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 resulta were used to more accurate-temperatures shown on ly determine the neutron fluence curve Number 1 of figure 3.6-1. The in the vessel beltline shell at a depth of one-fourth of the wall applicability of this curve to these tests thickness. These determined values is extended to non- of neutron fluence and the methods nuclear heatup and in Reguintury Guide 1.99 were used ambient loss cool- to predict the changes in reference down associated temperacure. RTNDT .

• Af ter

• testing the specimens described in 4.6.A.3, 1 i te e tup Figure 3.6-1 shall be updated if the<

and cooldown rates do test results indicate an update is i not exceed 15*F per necessary. j hour.

5. The reactor vessel head rei ly ens n 5. When the reactor vessel head bolting (four sequences of the studs are tensioned and the reactor 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. Prior to and during startup of flange must be greater than 100*F. and must an Idle recirrailntion loop, the g:mpernLure of the renetor coolant remain above 100*F while under full tension. in the operating and idle loops shall be permanently logged.

6.

• ie at$on1 7. Prior to starting a recirculation

' shall not be started Pump, the reactor coolant unless the temperatures in the dome and in s

temperatures of the

' the bottom head drain shall be

$*Nd recirculation loope pe t g compared and permanently logged.

are within 50*F of each other.

7 The reactor recirculation pumps shall not be started unless the coolant temperatures between the done and bottom head drain are within t e 5'F.

, 186

• I 1

. , . _ . _ _- _ _ . , . _ _ . _ . _ _ _ _ . _ _ _ . . _ . . - . _ _ _ . - _ . _ _ . _ . . . _ _ _ - ~ _ - . _ . .

8/24/83 Figure 3.6-1 1200 - 1 2 3 Curve #1

+ ,

Minimum temperature i,

~

~I~~~ ~~

, for pressure tests

,,, i i J_

j _,_l _-_-_- such as required by

~-~_, Section XI. Minimum T.

f___--

1000 -

temperature of 1700F 1

is required for test

'l

11 pressure of 1,100 ps.ig.

E'

_ Curve #2

,f.,_u__.

i isi Minimum temperature for l ' ' mechanical heatup or 800 - i "---

h l cooldown following 2 , +-ffrr 'j '

a. I nuclear shutdown.

o

_ i l  ! l 1 i Curve #3 3

g i

, j [ Minimum temperature for

, j f  ;' core operation z 600 - i Hg r

(criticality) includes ij j additional margin yG ,

t- >r .]

, i i

@$ ', required by 10CFR50, 1 #

y Appendix G, g 8 l l .

Par. IV.A.3 which N ---~/ l became effective '

O 400 - i July 26, 1983.

I I  !

U h --- h.,___ d r' Notes These curves include

-f  !

sufficient margin to

, ,r  ; provide protection 200 - . i i against feedwater nozzle f! ' '

9  !  ; degradation. The curves

]/ ~/

. , f l , allow for shifts in RTNDT of the reactor vessel 1

i >

• i ';'

,i!' f-y , JOL1-lTPNC22Lu utu.

l i l beltline materials to com-0-

- A +W '

r . W -+ '

. i pensate for radiation 0 100 200 embrittlement for the 300 400 life of the plant.

MINIMUM TDIPERATURE

( F) i 207 i

L

_g g n J-, b - L-_ ._ __ u- -. -- .m ---_A -m h w u,- 4

(

O e

J

• e i

DELETED O

FIGURE 3.6 2 .

I CHANGE IN CHARPY V TRANSITION TEMPERATURE . -

VERSUS -

NEdTRON EXPOSURE 208 l

1

4 1.6/4.6 BASES 3.6.A/4.6.A Thermal and Pressurization Limitation 1s The pressurevessel h a been analyzed for stresses caused by thermal and transients.

Heating and cooling transients throughout plant life et uniform rates of 100*r per hour were considered in the temperature range of 100 to 546*P and were shown to be within the requirements for stress intensity and fatique limits of Section Edition including III of theSummer ASME1966 Boiler and Pressure addenda) . Vessel Code (65 Operating limits on the reactor vessel pressure and temperature during hydrostatic normal heatup and cooldown, an1 during inservice testing, were established using Appendix G of the Summer 1972 Pressure vessel Code, 1971 Addenda to Section III ofasthe Edition, ASME Boiler and a guide.

limits assure that a large postulated surface flawThese operating depth of one quarter of the material thickneca, can, having a accommodated in regions of the vessi shell remote frombe safely discontinuities. For the purpose of setting these operating limits the reference temperature, RTHDT, of the vessel material was estimated from impact test data taken in accordance with requirements of 'the code to which this vessel was designed and manufactured (65 Edition to Summer 1966 addenda.).

' The fracture toughness of all ferritic steels gradually and uniformly decreases with exposure to fast neutrons.above a threshold for this in value, and it isof the operation prudent the RPV.and conservative to account are needed in this analysis: 1) A relationship Two types of information between the i 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, RTNDT, is provided in Regulatory Culde 1.99. In turn, this channe in reference temperature can be related to a change in

, the temperature ordinate shown in F1p,ure C-2110-I In Appendix C of Sectina 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 l 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 dositacter wires which had been installed adjacent to the vessel wall at the core midplane 1cvel were removed and tested to determine.the neutron fluence. Three sets of mechanical test specimens representing the base metal, weld metal and weld heat j affected zone have also been placed adjacent to the vessel wall at the core midplane icvel.

10 CFR 50, Appendix H.These will be removed and tested as required by Until such testing is performed, the changes in 220 4

r---t-- c+--* e- + - - - -- = v~ -* *-s--y- - - ' - . - - ~ - - - - - --- - > - +------a- - - * * - - - - - - - --m-- - - - + - - - * - - - - - - - -

. . l 3.E/4.6 E.AgIf reference temperature, RTNDT, will be based on the results of the testing of the dosimeter wires and the methods .'n Regulatory Guide 1.99. The operating pressure-temperature limite. shown in Figure 3.6-1 will be adjusted if necessary when the test results for the mechanical test specimens are available.

. As described in paragraph 4.2.5 of the safety 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 lown.c temperature

. than that in the upper regions of the vessel when there is no i' t. . . recirculation flow. This colder water is forced up when

- recirculation pumps are started. This will not result in stresses which exceed AsME Boiler and Pressure vessel Code,Section III limits when the temperature differential is not greater than 145'F.

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 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 ny tron radiation flue.nce at the closure flanges is well below 10 nyt y,1 Mev; therefore, radiation effects will be minor and will not influence this temperature, 221

• ENCLOSURE 2 DESCRIPTION OF PROPOSED CHANGE TO TECHNICAL SPECIFICATIONS FOR BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 (TVA BFNP TS 191)

Page 175 - Units 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 o

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 RT NDT*

No additional justification is needed.

Page 195 - Units 1 and 2 .

Page 208 - Unit 3 Description Figure 3.6-2 " Change in charpy V Transition Temperatur-e 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.

Pages 215 and 216 - Units 1 and 2 Pages 220 and 221 - Unit 3 Description l

These proposed changes revise the BASES to more accurately reflect current industry reference practices temperature RY regarding determination of changes in BASES to reflect what ha! be. The change updates the specification en done with neutron dosimeter wires that were installed adjacent to the reactor vessel wall. It also describes what will be done with mechanical test specimens. It describes temperatureour RT future plans for determining changes in reference NDT*

0

Description and Justification (cont.)

Page 194 - Units 1 and 2 Page 207 - Unit 3 Description Revise figure 3.6-1 as necessary to reflect more realistic, but conservative, values of beltline material RTNDT based on material analyses and testing. i Reason O

Provide lower, but conservative, pressure-temperature limits which will eliminate sealing containment prior to pressure testing.

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ENCLOSURE.3 JUSTIFICATION AND SAFETY ANALYSIS (TVA BFNP TS 191)

! The. pressure-temperature limit curves which are presently implemented at Browns Ferry Nuclear Plant (BFN) are based on a conservative baseline RTNDT in the vessel beltline region to compensate for embrittlement l oaused by 6.0 effective full power years (EFPY) of accumulated neutron fluence. Thus, the presently used curves are based on a maximum RTNDT Of 40 + 36 = 760F in the vessel beltline region and a maximum RTNDT Of 400F in the vessel closure flange region and the feedwater nozzle.

Combustion Engineering, Inc. (CE), performed material analysis to determine

• the actual copper content for the limiting vessel beltline materials in  !

Examination of the results showed that the maximum units 1, 2, and 3 copper content is in the unit 2 weld material and is equal to 0.20 percent.  !

. Charpy impact testing has been performed at Singleton Materials Engineering Laboratory to determine the baseline RTNDT for the beltline materials in units 1, 2, and 3 The combined end results of the CE analysis and the Singleton testing show that the maximum final RTNDT is in the unit 2 ,

weld material and is equal to 580F. This value of 580F is less than the value of 760F on which the presently used curves are based and therefore.provides a basis for relaxing the pressure-temperature limits.

- The proposed curves are based on a beltline region 40 EFPY RTNDT CI 580F which is based on actual material testing. The curves also provide relief on minimum temperature for core operation (criticality) as allowed

for boiling water reactors by revised 10 CFR 50 appendix G which became i j effective July 26, 1983.- They also include temperature limits as recommended in General Electric Company Service Information Letter 207, *

November 1979, to protect against further degradation of the feedwater nozzles; therefore, the safety of the plant is not degraded due to the

proposed curves.

The revisions proposed for section 4.6.A and bases are being made to i

refloot surveillance in accordance with Regulatory Guide 1.99. The neutron flux wires were installed for the purpose of exoerimentally verifying the calculated values of neutron fluence. This proposed change reflects that

{ the wires were removed and tested during the first refueling outage and the

- results were used to determine the neutron fluence. The proposed l

specification also outlines planned actions regarding future revisions to the technical specifications concerning reference temperature RTNDT. Figure 3 5-2 as shown in the technical specifications is not consistent with Regulatory Guide 1 99. We do not consider it appropriate l~ 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.

i This in no way affects the safety of the plant.

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. ENCLOSURE 4 BROWNS FERRY NUCLEAR PLANT SIGNIFICANT HAZARDS CONSIDERATION FOR ,

PROPOSED TECHNICAL SPECIFICATION CHANGES Fiours 3 L ~l-

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The revision reflects more realistic, but conservative values of RT NDT f r the reactor vessel beltline region and provides a

• margin of safety which complies with the fracture toughness

. requirements in 10CRF50, appendix G; therefore, this revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Does the proposed amendment create the proability of a new or different kind of accident from any accident previously evaluated?

No. The new pressure-temperature limit curves provide the required margin of safety and will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3 Does the proposed amendment involve a significant reduction in a margin of safety?

No. The revision provides a margin of safety which complies with the fracture toughness requirements in 10CFR50, appendix G, and, therefore, does not involve a reduction in a margin of safety.

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BROWNS FERRY NUCLEAR PLANT SIGNIFICANT HAZARDS CONSIDERATION FOR ,

PROPOSED TECHNICAL SPECIFICATION CHANGES T.S. 4.I,. A ed bsm

1. Does the proposed amandment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The revisions are being made to reflect the present status of the flux wires, planned actions regarding testing of the flux wires, and updating the technical specifications with respect to Regulatory

. . Guide 1.99. It does not involve an increase in the probability of any accident.

2. Does'the proposed amendment create the proability of a new or different kind of accident from any accident previously evaluated?

See #1.

3 Does the proposed amendment involve a significant reduction in.a margin of safety?

. See #1.

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