ML20079C364

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Proposed Tech Specs Figures 3.6.1,3.6.2 & 3.6.3 Re Reactor Pressure Vessel Thermal & Pressurization
ML20079C364
Person / Time
Site: Pilgrim
Issue date: 06/11/1991
From:
BOSTON EDISON CO.
To:
Shared Package
ML20079C363 List:
References
NUDOCS 9106190275
Download: ML20079C364 (21)
v • d • e


Text

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l ATTACHMENT B i and Pressurization Limits RPV Thermt Revised Technical Soetification Pagen

! 123 l 124 124A 125 128 128A l 128B 139 139A i

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9106190275 910611 PDR ADOCK 05000293 P PDR

LIMITfNG CONDITION FOR OPERATION SDRVEILLANCLRfg!REMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6. PRIMARY SYSTEM BOUNDARY Aeolicability: ADolicability:

Applies to the operating status of the Applies to the periodic examination and reactor coolant system, testing requirements for the reactor cooling system.

Obiective: Obiective:

To assure the integrity and safe To determine the condition of the operation of the reactor coolant system reactor coolant system and the operation of the safety devices related to it.

Soecification: Soecification:

A. Thermal and Pressurization A. Thermal and Pressurization Limitations limitations

1. The average rate of reactor 1. During heatups and cooldowns, with I coolant temperature change during the reactor vessel temperature less l normal heatup or cooldown shall than or equal to 450'F, the l not exceed 100*F/hr when averaged temperatures at the following I over a one-hour period except locations shall be permanently I when the vessel temperatures are logged at least every 15 minutes above 450'F. The reactor vessel I until the difference between any flange to adjacent reactor vessel l two readings at inoividual l shell temperature differential locations taken over a 45 minute I shall not exceed 145'F. period is less than 5'I:
a. Reactor vessel shell adjacent to reactor vessel flange l
b. Reactor vessel shell flange
c. Recirculation loops A and B
2. The reactor vessel shall not be 2. Reactor vessel shell temperatures, l pressurized for hydrostatic including reactor vessel bottom l and/or leakage tests, and head, and reactor coolant pressure I critical core operation shall not shall be permanently logged at be conducted unless the reactor least every 15 minutes whenever the vessel temperatures are above I shell temperature is below 220'F those defined by the appropriate and the reactor vessel is not curves on Figures 3.6.1, 3.6.2, vented.

and 3.6.3. (Linear interpolation l between curves is permitted). At l Test specimens of the reactor stated pressure, the reactor l vessel base, weld and heat affected vessel bottom head may be l zone metal subjected to the highest maintained at temperatures below I fluence of greater than 1 Hey those temperatures corresponding l neutrons shall be installed in the i to the adjacent reactor vessel l reactor vessel adjacent to the shell as shown in Figures 3.6.1 vessel wall at the core midplane l

and 3.6.2. I level. The specimens and sample program shall conform to the Amendment No. 82 123 l

l.1MITING CONDITION FOR OPERATf0N SURVEILLANCE RFJ)1)IREMENTS I

i 3.6.A Thermal and Pressurization 4.6.A Thermal and Pressurization Limitations (Cont'd) limitations (Cont'd)

In the event this requirement requirements c' ASTM E 185-66.

is not met, achieve stable Selected neutron flux specimens reactor conditions with shall be removed at the frequency reactor vessel temperature required by Table 4.6.3 and tested above that defined by the to experimentally verify appropriate curve and obtain adjustments to Figures 3.6.1, an engineering evaluation to 3.6.2, and 3.6.3 for predicted NOT I determine the appropriate temperature irradiation shifts. I course of action to take.

3. The reactor vessel head 3. When the reactor vessel head bolting studs shall not be bolting studs are tensioned and the unaer tension unless the reactor is in a Cold Condition, the temperature of the vessel reactor vessel shell temperature head flange and the head is immediately below the head flange greater than 55'F. shall be permanently recorded.
4. The pump in an idle 4. Prior to and during startup of an recirculation loop shall not idle recirculation loop, the be started unless the temperature of the reactor coolant temperatures of the coolant in the operating and idle loops within the idle and operating shall be permanently logged.

recirculation loops are within 50*F of each other.

5. The reactor recirculation 5. Prior to starting a recirculation pumps shall not be started pump, the reactor coolant unless the coolant temperatures in the dome and in the temperatures between the dome bottom head drain shall be compared and the bottom head drain are and permanently logged, within 145'F.
6. Thermal-Hydraulic Stability Core thermal power shall not exceed 25% of rated thermal power without forced recirculation.

B. Coolant Chemistry B. Coolant Chemistry

1. The reactor coolant system 1. a. A reactor coolant sample shall radioactivity concentration be taken at least every 96 in water shall not exceed 20 hours and analyzed for microcuries of total iodine radioactivity content, per ml of water.
b. Isotopic analysis of a reactor coolant sample shall be made at least once per month.

l Amendment No. 82 124

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS

' 3.6.B Coolant Chemistry (Cont'd) 4.6.B Coolant Chemistry (Cont'd)

2. The reactor coolant water shall 2. During startups and at steaming not exceed the following limits rates less than 100,000 pounds with steaming rates less than per hour, a sample of reactor 100,000 pounds per hour, except coolant shall be taken every as specified in 3.6.B.3: four hours and analyzed for chloride content. l Conductivity . . 2 pmho/cm Chloride ion .. 0.1 ppm
3. For reactor startups and for the 3. a. With steaming rates of 100,000 first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after placing the pounds per hour or greater, a reactor in the power operating reactor coolant sample shall be condition, the following limits taken it least every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> shall not be exceeded. and analyzed for chloride ion content.

Conductivity. . 10 pmho/cm Chloride ion. . 0.1 ppm b. When all continuous conductivity monitors are

4. Except as specified in 3.6.B.3 inoperable, a reactor coolant above, the reactor coolant water sample shall be taken at least shall not exceed the following daily and analyzed for limits when operating with conductivity and chloride ion steaming rates greater than or content.

equal to 100,000 pounds per hour.

Conductivity. . 10 pmho/cm Chloride ion. . 1.0 ppm

5. If Specification 3.6.B cannot be met, an orderly shutdown shall be initiated and the reactor shall be in Hot Shutdown within 24 hrs. and Cold Shutdown within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C. Coolant Leakaae C. Coolant Leakaae

1. Any time irradiated fuel is in the 1. Reactor coolant system leakage reactor vessel and reactor coolant shall be checked by the sumn and temperature is above 212*F, air sampling system and recorded at reactor coolant leakage into the least once per day.

primary containment from unidentified sources shall not exceed 5 gpm. In addition, the total reactor coolant system leakage into the primary containment shall not exceed 25 gpm

2. Both the sump and air sampling systems shall be operable during reactor power operation. From and after the date that one of these systems is made or found to be inoperable fur any reason, reactor Amendment No. 42 125

l TABLE 4.6.3 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM HITHDRAHAL SCHEDULE Effective Full I Capsule Power Years l Number (EFPY) l l

1 1 4.17 l l

2 15 l (approx.) l l

3 32 l (End of Life) l l

Amendment No. 82 124A

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g o

o @ o 8 ^

n - s * = 8, r 8 8 n 8 y n 8 8 s -

- - - n -

(4sd) anssay l

l Amendment No. R88 l

[

L

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

Bases:

3.6.A and 4.6.A Thermal and Pressurization Limitations (Cont'd)

The reactor coolant system is a primary barrier against the release of fission products to the environs. In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.

Appendix G to 10CFR50 defines the temperature-pressurization restrictions for hydrostatic and leak tests, pressurization, and critical operation. These limits have been calculated for Pilgrim and are contained in Figures 3.6.1, 3.6.2, and 3.6.3. l The bottom head, defined as the spherical portion of the reactor vessel l located below the lower circumferential weld, was also evaluated. Reference I transition temperatures (RTyg7) were developed for the bottom head and the I resulting pressure vs. temperature curves plotted on Figures 3.6.1 and 3.6.2. I It has been determined that the bottom head temperatures are allowed to lag i the vessel shell temperatures,

Reference:

Teledyne Engineering Services (TES) l report TR-6051C-1, dated June 27, 1986. The referenced analysis utilizes the l stress results established in the Combustion Engineering Inc., Pilgrim Reactor l Vessel Design Report, No. CENC 1139, dated 1971, and combines the stress l analysis results, specific to the bottom head, with the pressurization i temperatures necessary to maintain fracture toughness requirements in i accordance with the ASME Boiler and Pressure Vessel Code,Section III, the l criteria of 10CFR Part 50, Appendix G, and the supplementary guidelines of l Reg. Guide 1.99, Rev. 2. l For Pilgrim pressure-temperature restrictions, two locations in the reactor vessel are limiting. The closure region controls at lower pressures and the beltline controls at higher pressures.

The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner.

Radiation exposure from fast neutrons (>l mev) above about 10" nyt may shift the NDT temperature of the vessel metal above the initial value. Impact tests l from the first material surveillance capsule removed at 4.17 EFPY indicated a l maximum RT yg7 shift of 55'F for the weld specimens. l Neutron flux wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall at the core midplane level. The wires and samples will be periodically removed and tested to experimentally verify the values used for Figures 3.6.1, 3.6.2, and 3.6.3. The withdrawal schedule of l Table 4.6.3 has been established as required by 10CFR50, Appendix H.

The RT y of the closure region is - 5'F. The initial RT yg for the beltline l weldan![basemetalare-50*Fand0*Frespectively. ThesekT yg7 temperatures l are based upon unirradiated test data, adjusted for specimen orientation in I accordance with USNRC Branch Technical Position MTEB 5-2. l l

Amendment No. 139 l

L

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

Bases:

3.6.A and 4.6.A Thermal and Pressurization Limitationi (Cont'd)

The closure and bottom head regions are not expo:eu to neutron fluence l

(> 1 Hev) over the vessel life sufficient to cause a shift in RT The i pressure-temperature limitations (Figures 3.6.1, 3.6.2, and 3.6.N.of the l closure and bottom head regions will therefore remain constant throughout i vessel life. Only the beltline region of the reactor vessel will experience a i shift in RT NOT with a resultant increase in Pressure-Temperature limits. l The curves apply to 100% bolt preload condition, but are conservative for lesser bolt preload conditions.

For critical core operation when the water level is within the normal range for power operation and the pressure is less than 20% of the preservice system hydrostatic test pressure (313 psi), the minimum permissible temperature of the highly stressed regions of the closure flange is RT uo + 60 - 55'F. A l conservative cutoff limit of 75'F was chosen as shown on figure 3.6.3 and as l permitted by 10CFR50 Appendix G, paragraph IV. A.3. This same cutoff is l .

included in the limits for hydrostatic and leak tests and for non-critical l I

operation, as shown on Figures 3.6.1 and 3.6.2 respectively, in order to be l l

consistant with the limits for critical operation. l The closure region is more limiting than the feedwater nozzle with regards to both stress intensity and RT yg7 Therefore the pressure-temperature limits of the closure are controlling.

The adjusted reference temperature shift is based on Regulatory Guide 1.99, i Revision 2, dated May 1988; the analytical results of General Electric Report i HDE 277-1285, Revision 1, dated January 21, 1985, regarding projected neutron I fluence; and Teledyne Engineering Services Reports, TR-60528-1, Revision 1, I dated June 26, 1986, as supplemented by TR-7487, dated April 16, 1991, for l RT y vs. fluence as a function of temperature and pressure, and TR-6052C-1, l dated June 27, 1986, for the RPV bottom head pressurization temperatures. l l

l l

Amendment No. 82 139A

)

1 i

1 1

ATTACHMENT C RPV Thermal and Pressurization Limits ,

Marked-Un Technical Specification Paaes !

I I

l 1

. , _ , , , _ _ _ _ _ _ _ _ . +T9e '

LIMITING CCNDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY Applicability: Applicability:

Applies to the operating status of the AppIles to the periodic examination and reactor coolant system. testing reautrements for the reactor cooling system.

Objective: Objective:

To assure the integrity and safe oper- To determine the condition of the reactor stion of the reactor coolant system. coolant system and the operation of the safety devices related to it.

Specification: Specification:

A. Thermal and Pressurization A. Thermal and Pressurization _

Limitations ( w, ts n . , . e < r. - nis ,

t i mi t a t i on_s_

1. The average rate of reactor coolant 1. vuringl ea ups and cooldownsythe[u g .f, temperature change during normal NNWRy ' emperaturemnall be heatup or cooldown shall not exceed s permane t y logged at least every' i

- nutes until the difference #1 M' 100*F/hr when averaged over a one-hour period escept when the vessel between any two readings (taken ove temperatures are above 450*F. a 45 mi nute-ped.Qd i s l es t than 5'F .^#* #"e Tne tht4.1 flarge tojshel

  • tempera- 6 ,16./ + / t+ f, . . e ture differential /Shal' noDtneed._ a. Reactor veTIC57i adjacent 7[,I.3,, i 145*F. $ffc T m.c,-/ t # 61 f1ange m v e o.e-tu.- v *.a sf-)

j

} b. Reactor vessel shell flange

c. Recirculation loops A and R p y c. 9 ,
2. The reactor vessel shall not be 2. Reactor vessel shell temperaturePand pressurized for hydrostatic and/or reactor coolant pressure shall be l #

leakage tests, and critical core permanently logged at least every

[ operation shall not be conducted T 15 minutes whenever the shell tem-f perature is below 220'F and the AN ess the reactor vessel temper tureS )

reactor vessel is not vented.

a propriate curves on Figures 3.6.Tf a n %.above that defined by the and 3.6.2.A In the event this Test specimens of the reactor vessel

/ Te~iiiifrement is not met, achieve base, weld and heat affected zone stable reactor conditions with reactor metal subjected to the highest flu-t r vessel temperature above that defined ente of greater than 1 Mev neutrons

! I by the appropriate curve and obtain shall be installed in the reactor

/ ,

an engineering evaluation to determine vessel adjacent to the vessel wall the appropriate course of action at the core midplane level. The

/^, to take, specimens and sample program shall 7 conform to the requirements of

/,,s. P~ ,M ,, I, . g ASTM E 185-66. Selected 123 AmendmentNo.g x

LIMITING CONDITION FOR OPERATION SURVED LANCE REQUIREMENTS 3.6.A Thermal and Pressurization 4.6.A Thermal and Pressurization Limitations (Cont'd) Limitations (Cont'd) neutron flux specimens shall be -

! removed at the frequency required /

by Table 4.6.3 and tested to /: ;.

experimentally verify adjustments /

)

ju adirleJ11DW, i r r adi a tionto Figures 3.6.

to r~.*-?~*>F 9 d J.C.

3. The factor .essel head bolting 3. W en V e T actor vessel hea s'uds shall ot ce under tension ing studs are tensioned and the

.- yi t ss th ' N.r cature of the reactor is in a Cold Condition, ve. u l h ,.? f 7 and the head the reactor vessel shell i is greae.r .h t n* F . tempe.ature immediately below the

~

head flange shall be permanently recorded.

4. Tht c. to 4 1 ;n idle recirculation 4. Prior to and during startup of an loop A all not ce started unless idle recirculation loop, the tem-the temperatures of the coolant perature of the reactor coolant within the idle and operating re- in the operating and idle loops circulation '300s are within 50*F shall be permanently logged.

of each other.

5. The reactor recirculation pumps 5. Prior to starting a recirculation shall not be started unless the pump, the reactor coolant temper-coolant temperatures between the atures in the dome and in the dome and the bottom head drain bottom head drain shall be are within 145'F. compared and permanently logged.
6. Thermal-Hydraulic Stability Core thermal power shall not exceed 25% of rated thermal power without forced recirculation.

B. Coolant Chemistry B. Coolant Chemistry

1. The reactor coolant system radio- 1. a. A reactor coolant sample shall activity concentration in water be taken at least every 96 shall not exceed 20 microcuries hours and analyzed for radio-of total iodine per ml of water. activity content.
b. Isotoptr analysis of a reactor coolant sample shall be made at leaft once per month,
2. The reactor coolant water shall 2. During sta-tups and at steaming not exceed the following limits rates less than 100,000 pounds per with steaming rates less than hour, a sample of reactor coolant 100,000 pounds per hour, except shall be taken every four hours as specified in 3.6.B.3: and analyzed for D D gde f content.

'~~

Conductivity . 2 pmho/cm Chloride ton . . 0.1 ppm Amendment N 124

TABLE 4.6.3 P REACTOR VESSEL MATERIAL ,I .

SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE g-- - ,-\

Effective Full luer e RT .1 S Capsule Power iears ( n / cm,- ) (weldmetal)k

( F) j Number (EFPY) (1/ 4 /T) _

1 4.17 1.8/1

/ 55 c

2 15 j 6. x 1,0 9) f (approx.) ( pprpx.) j' / '/

s 3

32.

(End of Life)

/

1.4 y !0

/

136 e 7 /

l (agprox.)

_ _ _ .s Amendment No ') 124A

. ..w.

n. . . . . e. . ., . . 3. . n. ;. . v ,. r ; z . .m

~. m. ...w ,

... a. 5 *

  • F. V

.. r * * 'a' '.' '.* " *r'"..W.._._'...e r__ .

2.s.I *oelan: Cherist v (Con:'d) 4.6.1 Ce:lant Cheris::v (C .:'d)

3. 7e: reat::: star:ups and for :he 3. a.
  • tith stea:ing rates Of 100,000 firs: 2 !. hours after pla:ing the pounds per hour or greater, a rea:::: in the power operating reacter :ociant sample shall be tendi:12, the f:11oving limits taken at leas: every 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> shall to: be ex eeded, and analyzed for chieride ion content.

Condu::ivi:y . . 10 '=.ho/::

b. **nen

= all continuous condue:1v1:y Chloride 1::.. . 0.1 pp: conitors are inoperable, a reactor coolan: sa:ple shall be taken at leas: daily and ar.alyzed f c:

conductivity and chieride 1:n conten:,

4 Excep as specified in 3.6.E.3 above, the reae::: coelan vater shall met ex:eed the following lir.1:s vnen cperating with s:ca=-

ing rates greate: than or equal to 100,000 pounds per hour.

C: du::ivi:y.. 10 u=he/ :

Chieride ics.. 1.0 pp:

5. If Specifica:10: 3.6.3 cannot be

=en, as c:derly shutdown shall be 1:.1:ia:ed and the reactor shall be in Ec: Shutdown within 2t. hrs.

and Cold Shutdown vi hin the ner:

E hcc:s.

  • C. Ceclan: Leakane -

C. Cocian: Leaksee -

1. Any :ine 1::adia:ed fuel is in the 1. Rea::c cociant syste leakage shall rea:::: vessel and rea :o: coolant be checked by the sump and at te=pera:ure is above 21:07, rea::c: sa pling sys:e= and re:: ded a coolan: leakage into the primary least once per day.

con:ainment f: unidentified sour:es shall not exceed 5 gym. In addi:icn, the total reacter coolant sys:e= leakage into the pri=a y con:ain=en: shall ac: exceed 25 sp .

1. Sc:h the su=o.and -

r.1: sampling sys-te=s shall be cperable during rea:-

power =peration. ?:c: and af:e :he date tha: o. of these systens is =ade or found to be i=cp-e able fc any :tason, react

..p M

A endment S 4h/

i 125

. FIGURE 3. 6.1 PILGRIM REACTOR VESSEL

~

PRESSURE - TEMPERWTVRE CiidTT HYDROSTANC AND LEAK TESTS

\

44 6 W rk 3-f.l 1

[q #

(

ned lui ;

w i ii t i ui i l l

1 1

3 i t, i i

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1200

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x l h u

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sm

>>1 , , , , >: , , , ,

-1000 , nz i O

-g --++F i

j += -

4 --

ll g 4._._g_-_._ ,

i ge_ \;;  ; j.__

~

i H

,, / i ,

9 ,

i I

800 , .: ',. 'Nx

-v a . si i i i __

/  ! 1 i \. i .I

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./ --,.

..,- N , d.- 4 .

T q_.h-_.. _.-- .__.

j l D

m 600 i

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m

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E FPY* FLU CE (dT) '

+--

2.8 x 1 17 n/cm2 d F

400 i 8.o i '

/ I

!.' l :i- 10.0 3.4 x 10 n/cm {

/. , !i b -i

/ .

- .4L i a ili '

i t.. ---

/j. / ,

> i

, + .---

200 .

/

i iI

-/ ~~

- , l'  !  ! ~Y *PNPS EFFECTIVE FULL POWER

"~~~

'l !j  !

YEARS (EFPY) A5 0F 12/83 -k -

! END OF C YCLE 6 IS 6.68. K ll,

~

l ltt4MtH++HH+H i l

\

c 40 80 120 16 0 200 240 TE MPER ATURE (*F)

\ '

l Revision 95 l28 l

FIGURE 3.6.2 PILGRIM REACTOR VESSEL PRESSURE - JEMPER ATUREM SUBCRITICAL/CRITd HEAT UP a COOL DOWN

~

M' 3, f , V l pc &

  1. J C .

a's 4

/

NN l f

w . ,

N l c: '

/ a ,

D m

600 ,' s SUSC ' IC AL- -- 1 x s

m -

W c:

HEAT a COOL \x  !

DO '

rw '

C. l i, ' CRITICAL 400 j l ,

s CORE ---

l' N OPERATION  :::

, l 'l , ,

- l  ;  : N

, / f; 't 2OO p  ! j '

/ i a s l ,

l s x

I f I N O i  ! '

I

~'

40 80 120 I60 200 240 TE M PER ATURE (*F)

Bases:

3.6.A and 4.6.A Thermal and Pressurization 1. imitations (Cont'd)

The reactor coolant system is a primary barrier against the release of fission

- products to the environs. In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.

Appendix G to 10CFR50 defines the temperature-pressurization restrictions for n Thesi

('\~'; \

hydrostatic and leak tests, pressurization, and critical operation.

l i mi t s,bavCbe e h~c ahul a ted for P i l gr i m and a re con t a i ne d i n F i gure s N.~o'.1, and '\

4/

+ 3.6. .

3, t. f) '

- wj Q"'" yFor Pilgrim pres ,sure-temperature restrictions, two locations in the reactor vessel are limiting. The closure region controls at lower pressures and the

) j ,

beltline controls at higher pressures.

The nil-ductility transition (NDT) temperature is defined as the temperature below which ferritic steel breaks in a brittle rather than ductile manner.

Radiation exposure from fast neutrons (>l mev) above about 10 nvt may -

the vessel metal Ab_ov_e t,he initial value. Im ct

,5_hlft_the. RDL_tempe t sts from the first materia ra ture of_l sTJ rve i l l anCe C aps u l e remove, d "f rom the reac tor shift for veselhave\establishedthemagnitudeefstheRT,,orst The\ shift, w ich is grea for theelow weldfor\ metal, is tabulated {he b various flue levels an FPY of operati

\

x RPV Wall f f .f. N F4uence (1/4T) \ RT,, o r

., 9 n

\ K \

hen

'7 2.8 x Ol' n/cm 2

\ 61*F _ k 3.4 x 10 n/cm 2 65'F

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

w_

Neutron flux wires and samples of vessel material are installed in the reactor vessel adjacent to the vessel wall A the corp midplane level. The wires and samples will be periodicall removed and tested-to-experimentally verify the values used for Figures 3. 'a & 3.6.2. The withdr6a1 schedule of Table

.4.6.3 has been established (As_\ required bn10CFR50, d 2,. t.] ) App}ndixH. _,,

-~r ,

T essu emp ature(imitatt of KLguSL-eb .1 and 3 6. 2 icAble o

/gs f j

-- ~-__

( . -

" N J

,E, w,

)(v{Qdq'M 139

cases:

3.6.A and 4.6.A Thermal and Pressurization Limitations (Cont'd) on irrad\atedteltcata\3cjustedfor\scetimencrhntat+on :(accordancewith g/

45NRt Branch technieel Posi' tion MTES 5-h \

/ o eyse emc e o t i\ilm M .6., 5.2 obicaoleto th closare re @ n reflec q RT., q C of

-5'F, acjgsted ftr soec on orientation,f The curvesalso 3,

calto1Lo:Ltest_dala__.

analy 1007. ocit creload4 (<

x condition, out are conservative for lesser colt preload ccnditions. )

For critical core ODeration unen the water level is within the normal range C,' ,) for power operation and tne cressure is less tnan 207. of the creservice system (}<

'[,

j

! nydrostatic test oressure (313 psi), the minimum permissiole temperature of the hignly stressed regions of the closure flange is RT..or + 60 55'F. j

(

)

/ *

(

c

\ "F // The closure region is more limiting than the feedwater nozzle with regards to

/ ' toth stress intensity and RT.,3,. Therefore the cressure-temperature limits

(' b M ,' of the closure are controlling.

s ..-v' N_ -

,/ ==

I f

/ I

(

( (1 J &

f(' l

\~ l l iI l ,

5. {1 4 - & ,r/ j y /

l Amencment No. D 139A

  1. Proposed Technical Soecification Chanaes:

Insert "A":

, and 3.6.3. (Linear interpolation between curves is permitted.)

At stated pressure, the reactor vessel bottom head may be maintained at temperatures below those temperatures corresponding to the adjacent reactor vessel shell as shown in Figures 3.6.1 and 3.6.2.

lasert "B":

including reactor vessel bottom head, Insert "C":

The bottom head, defined as the spherical portion of the reactor vessel I located below the lower circumferential weld, was also evaluated. I Reference transition temperatures (RTgg7) were developed for the bottom l head and the resulting pressure vs. temperature curves plotted on 1 Figures 3.6.1 and 3.6.2. It has been determined that the bottom head I temperatures are allowed to lag the vessel shell temperatures, i

Reference:

Teledyne Engineering Services (TES) report TR-6051C-1, dated i June 27, 1986. The referenced analysis utilizes the stress results l established in the Combustion Engineering Inc., Pilgrim Reactor Vessel l Design Report, No. CENC 1139, dated 1971, and combines the stress l analysis results, specific to the bottom head, with the pressurization l temperatures necessary to maintain fracture toughness requirements in l accordance with the ASME Boiler and Pressure Vessel Code,Section III, I the criteria of 10CFR Part 50, Appendix G, and the supplementary l guidelines of Reg. Guide 1.99, Rev. 2. I Insert "0":

Impact tests from the first material surveillance capsule removed at l 4.17 EFPY indicated a maximum RT ggr shift of 55*F for the weld specimens.l Insert "E":

The RT yg7 of the closure region is - 5'F. The initial RTngy for the l beltline weld and basemetal are -50*F and 0*F respectivel'y, These RT yg7 l temperatures are based upon unirradiated test data, adjusted for I specimen orientation in accordance with USNRC Branch Technical Position l HTEB 5-2. The closure and bottom head regions are not exposed to I neutron fluence (> 1 Hev) over the vessel life sufficient to cause a i shift in RT . The pressure-temperature limitations (Figures 3.6.1, l 3.6.2,andY.6.3)oftheclosureandbottomheadregionswilltherefore l remain constant throughout vessel life. Only the beltline region of the I reactor vessel will experience a shift in RTgg7 with a resultant i increase in Pressure-Temperature limits. 1 I

, ' Insert "F":

.' A conservative cutoff limit of 75'F was chosen as shown on Figure 3.6.3 l and as permitted by 10CFR50 Appendix G, paragraph IV. A.3. This same I cutoff is included in the litrits for hydrostatic and leak tests and for l non-critical operation, as shown on Figures 3.6.1 and 3.6.2 l respectively, in order to be consistant with the limits for critical l operation. l Insert "G":

The adjusted reference temperature shift is based on Regulatory Guide i 1.99, Revision 2, dated May 1988; the analytical results of General l Electric Report HDE 277-1285, Revision 1, dated January 21, 1985, I regarding projected neutron fluence; and Teledyne Engineering Services l Reports, TR-6052B-1, Revision 1, dated June 26, 1986, as supplemented by l TR-7487, dated April 16, 1991, for RT vs. fluence as a function of I temperatureandpressure,andTR-6052fl,datedJune 27, 1986, for the l RPV bottom head pressurization temperatures. l

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