ML20217G600
| ML20217G600 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 10/07/1997 |
| From: | Lyons J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217G603 | List: |
| References | |
| NUDOCS 9710140050 | |
| Download: ML20217G600 (22) | |
Text
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UNITED STATES p-y NUCLEAR REGULATORY COMMISSION WA8HING1ON, D.C. 30e66-0001
\\=...+
CAROLINA POWER & LIGHT COMPANY. et al.
DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.189 License No. OPR-71 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by Carolina Power & Light-Com any (the licensee). dated January 7, 1997 as supplemented on Jul 25, 1997. August 27, 1997, and September 15, 1997, complies wit the standards and requirements of the Atomic Energy Act of 1954. as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I:
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission:
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations:
D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No.
DPR-71 is hereby amended to read at follows:
9710140050 971007 DR ADOCK 05000324 PDR
__.-_-__..___._.m.
p.s 4
Y 2
-(2)
Technical Soecifications -
The Technical Specifications contained in A)pendices A and B. as 4
revised through Amendment-No.189. are here)y incorporated in the license. Carolina Power & Light Company shall operate the j
facility in accordance with the Technical Specifications.
9' 3,
Thic license amendment is effective as of the date of its issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGUIATORY COMMISSION i
.t ame; E. L (lirector i
roject Direct ote 11-1 4
' vision of Reac or Projects - 1/il-
]'
Of 'ce of Nuclear Reactor Regulation
Attachment:
Changes to the Technical j
Specifications Date of Issuance:
October 7, 1997 d
J 3.
f 1
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ATTACHMENT'TO LICENSE AMENDMENT NO.189 FACILITY OPERATING LICENSE NO. DPR DOCKET NO. 50-325' Replace the following ) ages of the Appendix A-Technical Specifications with-the enclosed pages.
T1e revised areas are indicated by marginal lines.
-Remove Paaes Insert Paaes 3/4 4 13-3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15
^
3/4-4-16 3/4 4-16 3/4 4 17 3/4 4 3/4 4 18 3/4 4-18
-3/4 4-19 3/4 4-19 B3/4 4-4 B3/4 4-4
REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SY: TEM LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on (1) Figure 3.4.6.1-1 for heatup by non-nuclear means, cooldown following a nuclear shutdown, and low power PHYSICS TESTS: (2) Figure 3.4.6.1-2 for operations with a critical core other than low power PHYSICS TESTS or when the reactor vessel is vented: and (3) Figures 3.4.6.1-3a or 3.4.6.1-3b, as applicable for inservice hydrostatic l
or leak testing, with:
a.
A maximum heatup of 100*F in any one-hour aeriod, except for inservice hydrostatic or leak testing at w11ch time the maximum heatup shall not exceed 30*F in any one-hour period.
b.
A maximum cooldown of 100*F in any one-hour period except for inservice hydrostatic or leak testing at which time maximum cooldown shall not exceed 30*F in any one-hour period.
A maximum temperature change limited to 10'F in any one hour period c.
during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.
The reactor vessel flange and head flange temperatures greater than or equal to 70*F when reactor vessel head bolting studs are under tension.
APPLICABILITY: At all times.
ACT10el:
With any of the above 1imits exceeded, restore the temperature and/or pressure-to within the limits within 30 minutes: perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the reactor coolant system: determine that the system remains acceptable for continued t,perations, or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.6.1.1 The reactor coolant system temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
BRUNSWICK - UNIT 1 3/4 4-13 Amendment No. 189 I
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-2 within 15 minutes prior to the withdrawal of control. rods to bring the reactor to criticality.
-4.4.6.1.3 The reactor material irradiation surveillance specimens shall be removed and examined to determine changes in material properties at the intervals shown in Table 4.4.6.1.3 1.
The results of these examinations shall be used to update Figures 3.4.6.1-1. 3.4.6.1-2. 3.4.6.1-3a and 3.4.6.1 3b. as I applicable. The cumulative effective full power years shall be determined at least once per 18 months.
1
=
BRUNSWICK - UNIT 1 3/4 4-14 Amendment No. 189 l
y FIGURE 3.4.6.1-1 PRESSURE TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPERATION WITH CORE NOT CRITICAL 1200_
t__
f M
1100
- j I
1 1000 I
t r
900 A
v l
800-j Y
700 i
S I/
NORMAL OPERATION w"
CORE NOT CRITICAL - UNIT 1 600 r
1 HEATUP / COOLDOWN y
550 psig -I A
LIMmNG CURVE 500 1
I OPERATE TO RIGHT AND/OR BELOW 400
/
uumNG UNE 1
I/
aut
[
- 1. FUEL IN REACTOR
- 2. s 16 EFPY 200
/
- 3. 7.1 x 10" N/CMb 1MEV 4 RTNDT = 93*F (114 T)
/
100 5.15 PSIINSTRUMENT LOCATION CORRECTION INCLUDED BOLTUP
- 6. REG. QUIDE 1.99 REV. 2 70*F 0
,,o 0
100 200 300 400 500 600 TEMPERATURE ('F)
BRUNSWICK - UNIT 1 3/44-15 Amendment No. 189
FIGURE 3.4.6.1-2 PRESSURE-TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPERATION WITH CORE CRITICAL 1200 1100-1000 900 i
i j.___
j 800 r[
J____.
700 i
I y
I NORMAL OPERATION a
f CORE CRITICAL - UNIT 1 g
600
/
HEATUP / COOLDOWN b
LIMITING CURVE
'i OPERATE TO RIGHT AND/OR BELOW g
500 LIMITING LINE j
OPERATION IN CROSS-HATCHED AREA
,____11 400 PERV'TTED ONLY WHEN WATER LEVEL
[!_..
OPERATION.
1 is wimlN NORMAL RANGE FOR POWER
[l BASES 300 if; 1 FUEL IN REACTOR y y.
L//:
Ifr
- 2. s 16 EFPY 200
[,'[!
- 3. 7.1 x to N/CM > 1MEV 8
y//r f// //!
L'ff/:
- 4. RTNDT = 93*F (1/4 T)
N///'
5.15 PSIINSTRUMENT LOCATION i
100 m
/ //?
CORRECTION INCLUDED fifff//:
/e//,7/
L/,
- 6. RcG. GUIDE 1.99 REV,2 ff//////!
/ y,r/ /ff/
e' 0
ro 4
sa zo-O 100 200 300 400 500 600 TEMPERATURE ('F)
BRUNSWICK - UNIT 1 3/4 4-16 Amendment No.189
FIGURE 3.4.6.1-3a PRESSURE TEMPERATURE LIMITS REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200
}
4 g..._
f..
1100
__ ___,/
_f,-___
1000-
/
o j._
900 r
j l_,_
~~
~
800 j/
f 700
/
.g r
s
/
g
~~~
~' 60 P psig"~/
~~
E 600 g
k j'
~
500 HYDROSTATIC
_a PRESSURE TEST-UNIT 1
~~
400 LIMmNG CURVE
~"~~~
~
~~~
~
'l OPERATE TO RIGHT AND/OR BELOW
_313 psig LIMITING LINE 300 BASES;
- 1. FUEL IN REACTOR
~
- 2. s 14 EFPY 200
- 3. 2.39 X 10" N/CM > 1MEV (N16 Nozzle) 2
__[
- 5. BOTTOM HEAD REGION RTNDT = 10*F (1/4 T) 100
- 6. BELTLINE REGION RTNDT = 85.4*F (1/4 T)
-~
-BOLTUP COR E N
D
((.:
~
~
~~~
~~"
- 8. REACTOR NOT CRmCAL 0
v..
+..,
.m
.v 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 TEMPERATURE ('F)
BRUNSWICK - UNIT 1 3/4 4-17 Amendment No.189
FIGURE 3.4.6.1-3b PRESSURE-TEMPERATURE LIMITS REACTOR VESSEL HYDROSTATIC AND LEAK TESTS 1200
___f__
g g
~
~
~
(
~
~
~
1100 j
1
/
1000 L[-
900 i
e
/
f 800
/
v[
df
__j
~
700 if
____.{
/
/
~
S u.i v
g 1___
600 s a psis g
500
~~
~
HYDROSTATIC PRESSURE TEST-UNIT 1 400 LIMITING CURVE 4
OPERATE TO RIGHT AND/OR BELOW
_L 313 psg LIMITING LINE 300 L
pASES:
- 1. FUEL IN REACTOR
~~~
- 2. s 16 EFPY 200
- 3. 2.73 x 10" N/CM > 1MEV(N16 Nozzle) 2
- 5. BOTTOM HEAD REGION RTNOT = 10*F (1/4 T) 100
~~~
- 6. BELTLINE REGION RTNOT = 85.4*F (1/4 T)
~"
7,15 PSI INSTRUMENT LOCATION
_ ~_~ ~ _
"'BOLTUP CORRECTION INCLUDED 70*F
- 8. REACTOR NOT CRITICAL 0
"I'"
o-1
..+..
60 70 30 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 TEMPERATURE (*F)
BRUNSWICK - L' NIT 1 3/4 4-18 Amendment No. 189
i j
i 4
THIS PAGE INTENTIONALLY DELETED.
g BRUNSWICK - UNIT 1 3/4 4 Amendment No.189 l
. ~ _ _ _ _
. _ _ _ _ _ _ _... _.y 1
L,.
REACTORCOOLANTSYST1tl j
BASES PRESSURE / TEMPERATURE LIMITS (Continued) start-up art.1 shutdown the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
During heatu). the thermal gradients in the reactor vessel wall produce thermal stresses w11ch vary from compressive at the inner wall to tensile at the outer wall. Thermal-induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
During cooldown, thermal gradients to be accounted for are tensila at the inner wall and compressive at the outer wall.
The reactor vessel materials have been tested to determine their initial RT. The results of these tests are shom in GE NEDO 24161. Revision 1.
I ReEctor operation and resultant fast neutron. E>l Mev, fluence will cause an increase in the RT Therefore, an adjusted reference temperature.. based upon the fluence,,an be predicted using the proper revision of Regulatory c
Guide 1.99. The pressure-temperature limit curve Figures 3.4.6.1-1. 3.4.6.1-
- 2. 3.4.6.1-3a, and 3.4.6.1-3b include predicted adjustments for this shift in -
RT, tion of the pressure-sensing instruments.at the end of indicated EFPY, as loca 4
The actual. shift in RT of the vessel material will be checked periodically during operatio,n by removing and evaluating, in accordance with ASTM E185-82, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Since the neutron spectra at the irradiation samples and vessel inside radius vary little, the measured transition shift for a sample can be adjusted with confidence to the adjacent section of the reactor vessel.
The 3.4.6.1-2, pressure-temperature limit lines shown in Figures 3.4.6.1-1.
3.4.6.1-3a and 3.4.6.1-3b have been provided to assure compliance j with the minimum temperature requirements of the 1983 revision to Appendix G of 10CFR50.
The conservative method of the Standard Review Plan has been used for heatup and cooldown.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing-these specimens are provided in Table 4.4.6.1.3-1 to assure compliance with the requirements of ASTM E185-82.
BRUNSWICK - UNIT 1 B 3/4 4-4 Amendment No, 189 l
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UNITED STATES y
g j
NUCLEAR REGULATORY COMMISSION t
WASHINGTON D.C. 30eeHoot
'+4.....,d CAROLINA POWER & LIGHT COMPANY, et al.
DOCKET NO. 50-324 l
BRUNSWICK STEAM ELECTRIC PLANT. UNIT 2 eM_ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 220 License No. DPR-62 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by Carolina Power & Light Com any (the licensee). dated January 7. 1997, as supplemented on Jul 25, 1997. August 27, 1997, and September 15, 1997, complies wit the standards and requirements of the Atomic Energy Act of i
1954. as amended (the Act). and the Comission's rules and regulations set forth in 10 CFR Chapter 1:
I B.
{
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the
)
Commission:
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public..nd (ii) that such activities will be conducted in compliance with the Commission's regulations:
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment:
and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:
1 s.
i a
(2)1 ' Technical Snecifications
~
The, Technical Specifications contained in A>pendices A and B; as j
revised through Amendment No. 220. are here)y incorporated in the l
l license.
Carolina Power & Light Company shall operate the facility in accordance with the Technical Specifications.
3 3
This license amendment is effective as of the.date of its issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION L[ M James E. Lyon. 01 ector-r i
roje:t Directo e 11-1 i
D'v1 ion of Reactor Projects - 1/11-Of e of Nuclear Reactor Regulation
~
4
Attachment:
-Changes to the Technical Specifications Date of Issuance: L 0ctober 7.1997 i
4 f:
3 4
4 i
}
1
eIIA 8 MENT TO LICENSE AMENDMENT NO. ?pn A1 EAflLITY OPERATING LICENSE NO. DPR 62 DOCKET NO. 50-324 Replace the following Sages of the Appendix A Technical Specifications with the enclosed pages.
Tae revised areas are indicated by marginal lines, Remove Paael Insert Paaes 3/4 4 13 3/4 4-13 3/4 4-14 3/4 4-14 3/4 4-15 3/4 4-15 3/4 4 16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-18 3/4 4-10 3/4 4-19 3/4 4-19 B3/4 4-4 B3/4 4-4
REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION l
3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on (1) Figure 3.4.6.1-1 for heatup by non nuclear means, cooldown following a nuclear shutdown, and low power PHYSICS TESTS: (2) Figure 3.4.6.12 for operations with a critical core other than low power PHYSICS TESTS or when the reactor vessel is vented: and (3)
Figures 3.4.6.1-3a or 3.4.6.1-3b. as applicable for inservice hydrostatic or I
leak testing. with:
a.
A maximum heatup of 100*F in any one-hour period, except for inservice hydrostatic or leak testing a*. which time the maximum heatup shall not exceed 30'F in any one hour period, b.
A maximum cooldown of 100*F in any one-hour period except for inservice hydrostatic or leak testing at which time maximum cooldown shall not exceed 30*F in any one-hour period, c.
A maximum temperature change limited to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.
The reactor vessel flange and head flange temperatures greater than or equal to 70*F when reactor vessel head bolting studs are under tension.
APPLICABILITY: At all times.
ACTION:
With any of the above limits exceeded. restore the temperature and/or pressure to within the limits within 30 minutes: perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the reactor coolant system; determine that the system remains acceptable for continued operations, or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4
4.4.6.1.1 The reactor coolant system temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inse vice leak and hydrostatic testing operations.
BRUNSWICK - UNIT 2 3/4 4-13 Amendment No. 220 l
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-2 within 15 minutes prior to the withdrawal of control rods to bring the reactor to criticality.
4.4.6.1.3 The reactor material irradiation surveillance specimens shall be removed and examined to determine changes in material properties at the intervals shown in Table 4.4.6.1.3-1.
The results of these examinations shall be used to update Figures 3.4.6.1-1. 3.4.6.1-2, 3.4.6.1-3a. and 3.4.6.1-3b as applicable.
The cumulative effective full power years shall be determined at least once per 18 months.
r BRUNSWICK - UNIT 2 3/4 4-14 Amendment No. 220 l
FIGURE 3.4.6.1-1 1
PRESSURE-TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPERATION WITH CORE NOT CRITICAL 1200
,ip___.
II w,
1100-g r
i i
1000 I
I
~
900
/
i I
f 800 l
r J
700-
/
i
_en J
l NORMAL OPERATION CORE NOT CRITICAL - UNIT 2 m
4 a:
600 6
- P8,7----
mmpf wawg by LIMITING CURVE o.
500 1
I f
I OPERATE TO RIGHT AND/OR BELOW 400
[
UMITING LINE I
1 I
BASES:
300'
[
f
- 1. FUEL IN REACTOR
/
f 2,5 16 EFPY 200
[
- s. 7.1 x so" tvCu2 > iMEV i
/
- 4. RTNDT = 81.4*F (1/4 T) s 7
100 5.15 PSIINSTRUMENT I;X:ATION CORRECTION INCLUL,dD BOLTUP ~
- 6. REG. GUIDE 1.99 REV. 2 7o*F 0
m 0
100 200 300 400 500 600 TEMPERATURE ('F)
BRUNSWICK - UNIT 2 3/4 4-15 Amendment No. 220
1 i
FIGURE 3.4.6.1-2 PRESSURE-TEMPERATURE LIMITS REACTOR VESSEL NORMAL OPERATION WITH CORE CRITICAL 1200 i 61 li II lI 1100 lI II' i
1000 l
1 T
?
900 I
f 800 J
J f
f I
700 i
i i
g 660 psig __ f a
NORMAL OPERATION 3
CORE CRITICAL - UNIT 2 600 HEATUP / COOLDOWN LIMITING CURVE 11.
500 OPERATE TO RIGHT AND/OR BELOW UMITING LINE OPERATION IN CROSS-HATCHED AREA 2
400
,1 PERMITTED ONLY WHEN WATER LEVEL IS WITHIN NORMAL RANGE FOR POWER
[
OPERATION.
EA.EEE 7
300
- 1. FUEL IN REACTOR f.,
IJ
- 2. s 16 EFPY pts 200
/2
- 3. 7.1 x to" N/CM > 1MEV 2
F/
\\
L/ /
- 4. RTNOT = 81.4*F (1/4 T) f/ /
/l
/ / /
/
100 5.15 PSIINSTRUMENT LOCATION
/ / /
/
CORRECTION INCLUDED
/// // l g/ /
/ />
- 6. REG. GUIDE 1,99 REV. 2 J' / / /./
/'
//
// /
//
0 7o.
,at 2io O
100 200 300 400 500 600 TEMPERATURE (*F)
BRUNSWICK - UNIT 2 3/4 4-16 Amendment No. 220
FIGURE 3.4.6.1 3a PRESSURE TEMPERATURE LIMITS REACTOR VESSEL HYOROSTATIC AND LEAK TESTS 1200
~
"4 M [y_
M
'1>.
l-4 1100 r'.
miq
. p/
-. 4r[... _
.j 1000
~
~
~.I
- ~
~
~
~
~
"d h.
'ruk-.p.
1>
l.
,.g_
q 900
.o/,
y
~
~
7.1:
~
~
800
_J..
/
.j 700
_/
. L
-q, y
a j
600-
/
s
-__--pj pg g
~
HYD90STA410
~
~
PRESSJRE TEST-UNIT 2 400*
uumN000RVE gggRIGHT AND/OR BELOW 300 BASES.
_~
_~_
- 1. FUEL IN REACTOR
- 3. ? 39 X 10" N/CM > iMEV (N16 Noule) 3 1:
~
~~
- 4. N10 WOZZLE HTNDT = 78 6*F (1/4 T)
- 8. BOTTO:i HEAD REGION RTNDT = 40*F (1/4 T) 100
~
~~~
~
~5
- 6. BELTLINE i'EGION RTNOT = 81.8'F (114 T)
- BOLTUP
- 7. g6 P Ig EggTION 70 F
- 8. REACTOR NF T CRITICAL 0
- i.
60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 TEMPERATURE ('F)
BRUNSWICK - UNIT 2 3/4 4 17 Amendment No. 220
FIGURE 3.4.6.1 3b PRESSURE. TEMPERATURE LIMITS REACTOR VESSEL HYOROSTATIC AND LEAK TESTS 1200 -
_..)
u-.
.,q M
N H
H
-f PI P'=
,a p'-
_ p 1000 h
r/
900
/
.p
_ y 800
__. j/..j 700 -
f.-
_v.
t
-~
w 600
/.-
. W P$$.
n
,.#p-o.
~~
~
PRESSURE TEST-UNIT 2
~
+_
400 LIMITING CURVE
~
~ [
1 OPERATE TO RIGHT AND/OR BELOW 33 3 pg, LIMITING LINE 300 -
bases.
- 1. FUEL IN REACTOR
~
~
~~~
~
~
~
--~
2.s16EFPY 200
- 3. 3.73 x,o geu:,,ugy cy,, g,,i,3 n
['
~
~
- 6. BOTTOM HEAD REGION RTNDT = 4o*F (114 T) 100
~
- 6. BELTLINE REGION RTNDT = 61.8'F (1/4 T)
J dhp:
~~
~~
- 8. REACTOR NOT CRITICAL
'i'0RhMIOWNh"LIMSAD N ro r 0
m e...... r 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 TEMPERATURE ('F) 4 BRVNSWICK - UNIT 2 3/4 4-18 Amendment No. 220
THIS PAGE INTENTIONALLY DELETED.
l i
l 4
BRUNSWICK - UNIT 2 3/4 4-19 Amendment No. 220 I
REACTORCOOLANTS,ySIQj i
LEES PRESSURE / TEMPERATURE LlHITS (Contin,ugd),
~
start up and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent i
with the design assunptions and satisfy the stress limits for cyclic operation.
l During heatu), the thermal gradients in the reactor vessel wall produce L
thermal stresses w11ch va from compressive at the inner wall to tensile at the outer wall.
Thermall induced compressive stresses tend to alleviate the j
tensile stresses induced y the internal pressure.
During cooldown, thermal gradients to be accounted for are tensile at the inner wall and compressive at the outer wall.
The reactor vessel materials have been tested to determine their initial
[
RTev.
The results of these tests are shown in GE NED0 24157 Revision 2.
l Reactor-operation and resultant fast neutron. E>l Nev, fluence will cause an j
increase in the RT Therefore, an ad sted reference temperature, based upon the fluence, er.can be predicted usi the proper revision of Regulatory j
Guide 1.99. The pressure / temperature 1 mit curves Figures 3.4.6.1 1, 3,4.6.1-
- 2. 3.4.6.13a, and 3,4.6.1-3b include predicted adjustments for this shift in 4
g RTer at the end of indicated EFPY, as well as adjustments to account for the location of the pressure sensing instruments.
i The actual shift in RTer of the vessel material will be checked periodically during operation by removing and evaluating. in accordance with ASTM E185 82, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius vary little, the measured transition shift for a sample can be adjusted with confidence to the adjacent section of the reactor vessel.
The pressure / temperature limit lines shown in Figures 3.4.6.1 1, 3.4.6.1 2, 3.4.6.1 3a. and 3.4.6,1 3b have been provided to assure compliance with the j
minimum tem)erature requirements of the 1983 revision to-Appendix G of 10CFR50. Tle conservative method of the Standard Review Plan has been used for heatup and cooldown.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4.6.1.3 1 to assure compliance with the requirements of ASTM E185 82.
BRUNSWICK. UNIT 2 8 3/4 4 4 Amendment No. 220 l