ML20065D645
| ML20065D645 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/30/1994 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20065D639 | List: |
| References | |
| NUDOCS 9404070169 | |
| Download: ML20065D645 (26) | |
Text
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* .BRAIDWOOD STATION UNITS 1 & 2 g REVISED PAGES:
VIII < e H' 3/4 4-32*
- 3/4 4-33 (Figure 3.4-2a) 3/4 4-35 ~-(Figure 3.4-3a) 3/4 4-37 (Table 4.4-5) m,
- 3/4'4-39
-!3/4 4-40 (New page) 3/4 4-40a (Figure 3.44a) (New 3/4 4-40 page) _
3/4 4-40b (Figure 3.4-4b) (eliminated, new 3/4 4-40a page); B 3/4 4-8
- NOTE: THESE PAGES HAVE NO' CHANGES BUT ARE INCLUDED FOR CONTINUITY ' .
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS : SECTION PAGE TABLE 3.4-2 REACTOR COOLANT-SYSTEM CHEMISTRY LIMITS.......'........ 3/4 4-25 TABLE 4.4-3 REACTOR, COOLANT SYSTEM CHEMISTRY SURVEILLANCE REQUIREMENTS.......................................1 3/4 4-26
-3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
>l pCi/ GRAM DOSE EQUIVALENT I-131.................. 3/4'4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................... 3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS ,.
Reactor Coolant System................................... 3/4 4-32 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1)...... 3/4 4 FIGURE 3.4-2b REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... 3/4 4-34 FIGURE 3.4-3a REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS APPLICABLE UP TO 32 EFPY.(UNIT 1)...... 3/4'4-35 FIGURE 3.4-3b REACTOR COOLANT SYSTEM'C00LDOWN. LIMITATIONS APPLICABLE UP TO 16 EFPY (UNIT 2)...... '3/4 4-36 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE....................... 3/4'4-37 , Pressurizer.............................................. 3/4 4-3B Ove rpres s ure Protection Systems. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4~4-39 -l FIGURE 3.4-4a NOMINAL PORV PRESSURE RELIEF SETPOINT VERSUS
- RCS TEMPERATURE FOR THE COLD OVERPRESSURE - i,.
32
/ ,.;
PROTECTION SYSTEM APPLICABLE UP T0 i&-EFPY (Unit 1) 3/4'4-40e(
-FIGURE 3.4-4b NDMINAL PORV PRESSURE RELIEF SETPOINT VERSUS RCS . ' p<
TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION P i-
, weucesco ve to u, creY 4- - S
' SYSTEM *tDNIT 2).................................... 3/4 4-400- er 3/4.4.10 STRUCTURAL INTEGRITY............................... ..... 3/4 4-42 3/4.4.11 REACTOR COO LANT SYSTEM VENTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4'4 ,
BRAIDWOOD - UNITS 1 & 2 VIII AMENDMENTNO.k
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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ' 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on~ Figures 3.4-2a and 3.4-3a for Unit 1 (Figures 3.4-2b and 3.4-3b for Unit 2) during
- heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 100'F in any 1-hour period,
- b. A maximum cooldown of 100'F in any 1-hour period, and
- c. A maximum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
APPLICABILITY: At all times. ACTION:
;' With any of the above limits exceeded, restore the temperature' and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T,yg and pressure to less than 200*F and 500 psig, respectively, within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during systes heatup, cooldown, and inservice leak and hydrostatic testing operations.- 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall-be removed and examined, to determine changes in material properties, as required by 10 CFR Part 50, Appendix H,' in accordance with the schedule in Table 4.4-5. . The results of these examinations shall be used to ~ update Figures 3,4-2a and 3.4-3a for Unit 1 (Figures 3.4-2b and 3.4-3b for Unit 2), and 3.4'4a for Unit.1 (Figure 3.4-4b for Unit 2). 1 BRAIDWOOD - UNITS 1 & 2 3/4 4-32 AMENDMENT NO. 30 L
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. possit(e instrument: errors
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TEMPERAWRE:(283*F)1 1 , N FOR UP 70 32!EFPY*-THE SEWICE PER { 250 ' A 3 0 0 90 1 1H 200 2H- 300 M0 40C - = 440 - ' 300 -- INDICATED TEMPERATURE (T) *
.e n_[
t 9 FIGURE 3.4-2a
/ REACTOR COOLANT SYSTEM HEATUP LIMITATIONS :
APPLICABLE UP TO 32 EFPP(UNIT 1) I
- applicability date has 'been reduced per Regulatory Guide'1.99 Revision'2 tot ,
4.5 EFPY' . The calculation to determine applicability utilized actual copper.
- ' BRAIDN' ent of' 0.05
. UNITS.1 & 2;wt%. 3/4.4v33 AMENDMENTtN0;f 30 ; '
1
d RTERIAL PROPERTY BASIR CONTROLLING MATERIAL: CIRCUFERENTIAL WELD INITIAL RTxter: 4 0*F ART AFTER 32 EFPY: 1/4T, 159'F 3/4T, 13 5*F These curves are applicable for heatup rates up to 100*F/hr for the service period up to 32 EFPY and contain margins of 10 F and 60 psig for possible instrument errors. 1500 a a i r. , .- i i! '> i ,<! i i =t - ' i i i
;, 'l l Leak Test Limit '[
, j j "i ii ti I 6 11 i i f 1250 l.. ll'
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UP T0 100*F/HR- ' S c 750 lll
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i 500 l , e Criticality Limit,Sased on i / Inservice Rydrostatic Test
~
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. Temp.
Feriod Up (304'F) To.3 3.E7FT ISO , , i . , , f i lI l 9 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEWPERATURE (DEG.F) 1 FIGURE 3.4-2a REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 32 EFPY (UNIT 1) BRAIDWOOD - UNITS 1 & 2. 3/4 4-33 AMENDMENT NO..
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., Per 'up to 32 IFFT*and contains margins'of 10
. ,o- . so.tro t .r,,,,
and 60~ for? W 33pq ' x ; > 2250 ' a.
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i > x ( g~1000 ' ACCEPTABLE S- , a
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x OPERATION - ai e- a, j 75cggs q RATES S ' x
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200 360 aos 460 :400- 430.. , 300.'- seicalto 1tematuat]sts.r) . y FIGURE 3.4-3a R REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS APPLICABLE-UP.TO.32 EFPY*(UNIT 1) -
- applicability has been reduced per Regulatory Evide 1'.99 Revision 2 to!
12 EFPY. The calculation to determine'. applicability utilized actual. ' copper content of 0.05.wt%. ~ v LBRAIDWOOD --UNITS 1 4 2
.~ 3/4 !.4-35 AMINDMENT NO.: 30, y a.,2 = , -v -,-,e-t,- = . - m-- e4 ,w ,t-
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--------- - = 7 - - - - fj
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, -j h - - ~. . . . . y.; e y .O k 3 g_ s- MATERIAL PROPERTY BASIS I
CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD INITIAL RT,er: 4 0'F ART AFTER'32 EFPY:. 1/4T, < 159 8F 13/4T, 135'F-These curves are applicable - for cooldown rates' up tor 100'F/hr for - A, the service ~ perlod up to 32 EFPY and contain margins of 10 0F'and 60 psig for possible Anstrument errors. H' '2500, w ,,w _ , , ,
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,8 :: . Operation ~ . i ' ~ '
O i # ,ii r 750 + LAcceptable,. g g L0peration; _ 4 g 'o'olema' ' .y
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0 50 100 150 200--. 2503 3300: :350 1400 4503 -5005 s INDICA TED. TEMPERATURC (OCC.r) - '1 L . h FIGURE:3.4-3a? h.; , < REACTOR COOLANT SYSTEM COOLDOWN/-LIMITATIONS: $ --APPLICABLE'UP TO;32 EFPY.'(UNIT'1) l' ' Y
!BRAIDWOOD- UNITS l':& 2 3/4 4-35 AMENDMENk'No.;
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REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM'- WITHDRAWAL: SCHEDULE - y
. o. UNIT-1
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- Wi thdrawali timej may.;be i modi fied : to) coincide ;wi th'.:. t'hosei re Se l ing ; o'utages; or :reac torfshutdowns -
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':most closely approaching"the wi_thdrawal.:>scheduleh ,
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REACTOR COOLANT SYSTEM L OVERPRESSURE PROTECTION SYSTEMS f LIMITING CONDITION FOR OPERATION C 3.4.9.3 At least two overpressure protection devices shall be OPERABLE, and ' I each device shall be either:
-(
- a. A residual heat removal (RHR) suction relief valve with a lift M setting of less than or equal to 450 psig, or k
(
- b. A power operated relief valve (PORV) with a lift setpoint that >
varies with RCS temperature which does not exceed the limit established in Figure ( {g 3g4 g g g APPLICABILITY: MCDES 4, 5, and 6 with the reactor vessel head on. ( ACTION: (
)
- a. With one of the two required overpressure protection devices ()
inoperable in MODE 4, restore two overpressure protection devices to L. OPERABLE status within 7 days or depressurize and vent the RCS through at least a 2 square inch vent within the next 8 hours. ( (
- b. With one of the two required overpressure protection devices 7 J
inoperable in MODES 5 or 6, restore two overpressure protection devices to OPERABLE status within 24 hours or vent the RCS through .) - at least a 2 square inch vent within the next 8 hours.
- c. With both of the required overpressure protection oevices inoperable, 7 depressurize and vent the RCS through at least a 2 square inch vent <
within 8 hours. X
.( ,
- d. With the RCS ventw per ACTIONS a, b, or c, verify the vent pathway >
at least once per 31 days when the pathway is provided by a valve (s) ( that is locked, sealed, or otherwise secured in the open position; }- otherwise, verify the vent pathway every 12 hours. (
) '
- e. In the event either the PORVs, RHR. suction relief: valves, .or the RCS vents are used to mitigate an RCS pressure transient ~, a Special >
Report shall be prepared and submitted to the Commission pursuant to ( Specification 6.9.2 within 30 days. The report shall describe the. L circumstances initiating the transient, the effect of the PORVs, RHR suction relief. valves, or RCS vents on the transient, and any k corrective action necessary to prevent recurrence. ( i (
- f. The provisions of Specification 3.0.4 are not applicable.
BRAIDWOOD - UNITS 1 & 2 3/4 4-39 AMENDMENTNO.ffI[
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FIGUE 23.4-4 a L ~ N . NOMINAL:PORY PRESSURE. E LIEF SffPO!arr VERsus(. .y - RCS TEMPERATURE FOR-THE CtH.D OVERPRE55URE PROTECTIONisYSTEM;. " APPLICABLE- UPcTO:10 EFPY*(UNITL1).- . . . NJ
-L* applicability has been. reduced per Regulatory Guide 1.99 Revision 2, to 4;51EFPY.c . i The calculation to determine applicability-utilized actual copper contentfof'0;05 wth
- i f.
BRAIDWOOD. UNITS 1 1 21 3/4 4 40a-
- AMENDMENT;NOP30 r
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}i-~-- .- / g- 177.0 '605.0 5==i' F."-
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227.0 745.0 ji_ . g =____; .-- 277.0 725.0 _ . -3 2 - 500
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50 100 150 200 250 300 350 400 T(RTD), Lowest COMS RTD Temperature (*F) FIGURE 3.4-4a NOMINAL PORV PRES 8URE RELIEF SETPOINT VERSUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 32 EFPY (UNIT 1) BRAIDWOOD - UNITS 1 & 2 3/4 4-40 AMENDMENT NO.
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50 100 200 300 400 TRTyLOWIST C0 pts RTD TD(PEPATURE (DEG F) FIGURE 3.4-4b NOMINAL PORY PRESSURE RELIEF SETPOINT VERBUS RCS TEMPERATURE FOR THE COLD OVERPRESSURE PROTECTION SYSTEM APPLICABLE UP TO 16 EFPY (UNIT 2) BRAIDWOOD - UNITS 1 &2 3/4 4-40a AMENDMENT NO.
egn : x ,vwn w n , e + , : imp i y . -
+
y .. . C REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) Heatup and cooldown limit curves are calculated using the most limiting v&TDyofthenil-ductilityreferencetemperature,RTNDT, at the end of 32\ effective full power years f~o7 nit 1 (16 effective full power years for N nif 2) of service life. The 32h E PY for Unit 1 (16 EFPY for Unit 2) service J life period is chosen such th t-t limiting RT NDT at the 1/4T location in the core region is greater than the RT of the limiting unirradiated material. NDT The selection of such a limiting RT NDT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements. The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-la for Unit 1 h (Table B 3/4.4-lb for Unit 2). Reactor operation and resultant fast neutron ? (E greater than 1 MeV) irradiation can cause an increase in the RT There-NDT. fore, an adjusted reference temperature, based upon the fluence, copper content and nickel content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART NDT computed by either Regulatory ( Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials" or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2. s The heatup and cooldown limit curves of Figures 3.4-2a (3.4-2b) and 3. 03a((3.4-3b) include predicted adjustments for this shift in RT NDT at the end o 32\ FPY for Unit 1 f 1 4 (16 EFPY for Unit 2) as well as adjustments for possible errors in the pressure and temperature sensing instruments. -Rev4+ e d-hea tup-a nd-c ooldown-c u rves-ha ve- 4(
-been-generated-for-Unit-2-4eeeeerdance-with-Regtriatory-Guide 1.09 Revision 2. p foe 4 nit-1-the-eurves-rematn-the-samer-However -the-app r 14 cab 444ty-date-bas- 4 bee n-reduc ed-pe M egul a to ry4ui de4r99-R e vis i on-2-to-4r5-EFFY-for-hea tup-and- 4 0-EFPFfor co&Mowrr.
Values of ART NDT oatermined in this manner may be used until the results from the material surveillance program, evaluated according to ASTM E185, are available. Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H. The surveillance specimen with-drawal schedule is shown in Table 4.4-5. The lead factor represents the rela-tionship between the f ast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule exceeds the calculated ART NDT NDT for the equivalent capsule radiation exposure. aFoMn i t-1-appl i cab i l i ty-da te rha v e-been-re v i s ed-i n-accorda nce-wi t bRegul atory- 5 494de-b9 %Revi si on-2r-to E F Pbfo r-he a tup-a nd-124-E E PY--f or-cooldow%. 2 BRAIDWOOD - UNITS 1 & 2 B 3/4 4-8 AMENDMENT NO.'3q
ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-72 AND NPF-77 Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed ~ amendment would not:
- 1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or
- 2. Create the possibility of a new or different kind of accident from any-accident previously evaluated; or
- 3. Involve a significant reduction in a margin of safety.
The proposed changes will incorporate new pressure-temperature curves and new low temperature overpressure protection curves for Braidwood Unit 1 covering plant operations through 32 effective full power years (EFPY). The changes will also revise the specimen removal table. A. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated. - The use of new pressure-temperature limit curves and low tem'perature overpressure protection curves will not change any postulated accident scenarios. The revised curves were developed using industry standards and regulations which are recognized as being inherently conservative.1 The pressure-temperature low temperature ' overpressure curves provide reactor coolant system.(RCS) limits to protect the reactor pressure vessel from brittle fracture by clearly separating the region of normal operations from the region where the vessel is subject to brittle fracture. The heatup and ~ cooldown limits are designed _to ensure that the 10 CFR 50 Appendix G Pressure Temperature limits for the RCS are not exceeded during any condition of normal operation including anticipated operational occurrences, kinla:brdwd:ltop:19
fy. %.. - 3 4> , General Design Criterion 32 of 10 CFR'50 Appendix A requires .that the' L reactor coolant boundary shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident condition, (1) the boundary behaves in a nonbrittle manner and (2) L the probability of rapidly propagating fracture is minimized. L 10 CFR 50 Appendix G, " Fracture Toughness Requirements,". requires that . the effects of changes in the fracture toughness of reactor vessel materials - caused by neutron radiation throughout the service life of nuclear reactor be considered in the pressure-temperature limits. The change is used in- [. conjunction with the material initial reference temperature (RTmn-) to establish the limiting pressure-temperature curves. Regulatory Guide 1.99, Rev. 2, contains procedures for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used.for light-water-cooled - reactor vessels. Using the Regulatory Guide 1.99, Revision 2, Braidwood Unit 1 Surveillanc, L Capsule U results, and Appendix G to 10 CFR 50, new Pressure-Temperature curves prepared for the projected reactor vessel exposurc at 32' EFPY of operation. These new curves, in conjunction with the heatup and. cooldown ranges and the revised Low-Temperature Overpressure Protection System setpoints, provide the required assurance that the reactor pressure vessel is protected from brittle fracture up 32 EFPY of operation. No changes to the. design of the facility have been made and no new equipment j has been added or removed. The revised an'alysis and resultant adjustment R of the operating limitations provide assurance that.the Reactor Coolant System is protected from brittle fracture. Revising the Reactor Vessel Material Surveillance Program Withdrawal Schedule does not result in the addition or removal of any equipment, or any design changes to the facility. Capsule lead times are revised and, for.
.Braidwood Unit 2, Capsule X will be removed next vice Capsule W. The <
proposed removal schedules remain consistent with ASTM 185-82. Therefore, the proposed amendment to the pressure temperature limitations. does not involve a significant increase in the probability or consequences of an accident previously evaluated. kanla:brdwd:ltop:20 mb
q , -- %- ,,. , w m , 1 1
- !.w ' % <
N ' 5 B. The proposed change does not create the possibility of a newLor;
, .different kind lof accident from any accident previously evaluated.3
'( , . The use~of the.new pressure-temperature operating limits and the naw low,
' temperature overpressure protection curve does notl change any postblated; accident scenarios. The new curves d6 not represent any appreciable!
change in the current methodologies; they merely; provide assurance that'
'4 the Reactor. Coolant System is protected from brittle fracture.jNo newg accident or malfunction mechanism is introduced byLthe amehdment andino; f' physical plant changes will result from this amendment.
e, Revision of the Reactor Vessel Material Surveillance Program Withdrawal! lf ,- Schedule does not introduce a new accident.or malfunction mechanism.' Capsule lead times are revised, and, other than' changing the order of: i ~ specimen removal, consistent with ASTM.185-82, no' physical plant changesO - will result from this revised schedule. n .. Therefore, the proposed changes do not create the possibility'of a new or different" kind of accident from any accident previously evaluated. < e V C. The propose'd change does not involve a significant reduction in'ai < R + margin of safety.. , The new pressure-temperature operating limits low temperature ? overpressure protection curves were generated with the currently' acceptsd . L conservativo methodology using capsule surveillance data.1 The new;
~
pressure-temperature' curves were developed usin'g industryistaridards andl . y regulations (ASME Code Section'III, and NRC Regulatory Guide.1'.99,' . ?
~
Revision 2) which are recognized as beihg inherently; conservative.yThe uss' *, e of the new pressure-temperature operating limits;and low temperature . overpressure protection lim _its would'not change postulated ' accident: , scenarios. . 4 y The proposed revision to the Reactor Vessel Material Surveillahce_ Program , Withdrawal Schedule.would not change. postulated _ accident:synarios.
' Capsule lead times are revised, and, other than changing th". Mrofe94
' specimen removal, consistent..with" ASTM 185-82, no physical' plant changes : .
will result from this revised schedule. Thereforelthe proposed chariges doc
~
not involve'a significant reduction in a margin of safety. , g ,
SUMMARY
( y; Based upon'the above. evaluation, CECO has concluded that these chahges involvo' no 'significant hazards considerations. , , t kanlatbrdwd:ltopr21
}
s _
. ?.U c
/c
' V
. a
n w, : .. m m y - . - , : a Ti (n3 _
..e id.wr , -
k& '
- ATTACHMENT D- '* +
ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A-t TECHNICAL SPECIFICATIONS OF; _ l , FACILITY OPERATING LICENSES: l' NPF-72 AND NPF-77 ..
,s ,
3 n y
,a ?:. ?
Commonwealth Edison has evaluated the proposed amendment and determined-that it meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9).
~
This determination is based upon the followingf The proposed amendment . L . changes requirements regarding the installation _ and use of facility components : , located within the restricted area (as' defined in 10 CFR 20) arid' surveillance < requirements; and the proposed amendment involves no 'significant hazards 2' ' considerations, no change in the amount or type of.any ofIluent' thrit 'may be- _. released offsite, and no increase in individual or cumulative occupational radiation > cxposure. Pursuant to 10 CFR 5L22(b), neither an environmental impact-
- statement nor an environmental assessment is necessary for'.th~e proposed
, amendment. W
-M 3.
? >
h n',: ?:1 v , v lv 4 a
/
{ h i .kinlaibrdwdsitop:22. Si Y s w/ -d 1 > i,
1,. _ a: , e - i s
+
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APPENDIX IL
? FIGURES: _.
CONTENTS: ia - l I
'c
$'3 : Figure B 3/4.4-1 Fast Neutron Fluence (E>1MeV) aa a Function of Fulli, ~ N Power Service Life - r -!. . . .
- Figure B 3/4.4-2 Effect of Fluence and Copper on' Shift of RTum. for L E Reactor Vessel Steels Exposed to Irradiation at 550_.*F .
' l.5 rc oi
?
5 4 5
.p- .
f f:. 9 l s e ,
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ty
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+-
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i ..i .. q- .i , a . s . f . . , e. .
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.r.u. =,:.=. ~ . -
.ur.=.: -+ - , - . ..w_..-... . ,
g
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. . _ = 4
? ? f"r I. t 5 ' w= . ira!J5 . tf F =;r '--W
) I.v Cr i !"r } / - lT !IJ $ ~~i--if3N
~~:~
2"' E T i 3 / a '. . E *-Ei 4 T i I? $ F.:'?:si
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.: .;.:.: a : o
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_.; : 4 G - _- ,._ : . f.L: ,. u. =_ .: : :._ _ .:. _:.. . __ . ____ _ .--- -- .. ,- .. :.-~ --- 3 z --
.:t===- r :*
w -
- -r. -___._-.-
u - _ . - .., i z - 1
-3 w . . . _ .
.3/4. T . 2.7 x 10 .8,N/cm .
1 2 3 2- . . . a
- u. . . . . .
-. . . . .6...., ,
g ... . .. % , . . - m .
- ~ f-3 ./ ~ '
( w - 18 z 10 I H=<1 - r<+s- hun.s.= t s . .a i .m
. g . 6. : J . . ar, a v . sa;;y~r u.us-i.ai c.> .rtr.4 4 =m ,
e re : .c == -
. , +3
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- z r .H em - : = .___ : -: . ._ .. _
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. .. _ .. -". 4.. # ". ". : :.:. .
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g " *"! /: . _ s 1 4 r .. i i .1 i 4 s 4.s-i s.s v.. .
- y. p . v.s.r pm.i e is: + 5: 2. - 1.: ; J 1 pY t h' r e?.L+ r.1::i i . p=. r #- %._s l, f b /- : 5-Li. ii9_P ?~r:C L55~.lls M - H-Z Z ~5s :_=:
- .. :.::M. :'l~=3 ~: :i':._r: .. . .--r- r.=t ... ..
. ._:=. . :::. r_.::: _ :. :_ _.~.d . _. r
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_._: ;: --t-2 _ f
,,,.. g . - I
, . i !
' 10 0 *
.. 43; 50l 0 10 20' '30 EFFECTIVE FULL POWER (Yerrs)
FIGURE B 3/4.4e1 FAST NEUTRON FLUENCE (E y 1 MeV) A5 A FUNCTION OF I FULL POWER SERVICE LIFE , b BRAIDWOOD - UNITS 1 & 2 8 3/4 4-13 : o
+ _
~
2- . y.s - f.( # u c.'.,- } y ' ll '
- m
.~. :-
m-g E '. -i I j HI j i !!) ll [ i 10 "iluPPG MMIT -~
" ,@hf]Q O ~ ~~-~~
l0.30% I t iCOPPER llilillitilliitlisil!!hi'14t illilill!!iliild
}{l I % [1g ,,,,
n BASE.0.25% WELD . .i - ... . .. c- .0.25% COPPER SASE,0.20% WELD y,
, g y d.M
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p i a g 180 ";'s,;;i.a # :ei
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- am p , M !
4 ;1l i ' N.- l ! 19 ,:
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,i4=
A. e . T"P'D i:aM7 , ji'
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!ll i !' h5.&,,7:
.MSII !lii lil: ..
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'" E!;
. , y y g.. 4d.s+$. .Q ai: ::!l S. ir N i i s.1s% COPPER BASE.8.10% WELD y;I.
j 8' }! ;ql,i, d i!' & - [ ff f S.10% COPPER SASE.'O.85% WELD
,;;, M #
l
'l k -
j I I: 4 1 h [! i l l! . !P! !! !s
' i J S
'ill li!J .
, i !h I5l ll!{ $ @
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-2
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g w.- I
. FAST NEUTRON FLUENCE INICM . E >1 Movl .
FIGURE B 3/4.4 ' EFFECT;0F FLUENCE'ANDLCOPPER ON SHIFT.0F;RTHDT x.: i REACTOR VESSEL STEEL 5' EXPOSED 10.IRRADI ATION ~ ATL 550'E~ ~
~
l
- . - .- - . :=.
m -. y , e u a; p ,- n Me @ , g, ki, C F-lt; y.t:, bj !; j.c + N ,- ' O M jfgk , t . 4 BRAIDWOOD STATIONi' y *
'oec,7,'19901 .
In reply' refer to.
- n. , -
' CHRON ;,No ?.j180070 7 q h To:.K.L.:Kofron , _
Braidwood Station Manager subject: Heatup and Cooldown' Curves =for'Braidwood Unit 21
Reference:
!NED' letter dated Nov. 27,1990. from E.D,.Swartz;to:-
~ K.L. Kofron M
The reference letter transmitted-revised ~:heatup/cooldown- A-curves for Braidwood Unit'1 bas'ed~on:fluenceLdata.. gather from the analysis of Surveillance Capsule.U.'It7was:noted:inLthe-letter that there'were two (2} typographical errorsion page' 10 of the WestinghouseLtransmittal., Westinghouse has supplied-a corrected page which has been' reviewed:by NED withinoJ comment,.and isLattached to this~1etter. Should-you have any question-or comments, please call-
~
~
[ Bob Waninski.at x-7387, Downers Grove.: YYe f}
. R.E..Waninsx1:
_,1 ngpeer PWR System R. [ Des gm g RDy-% y Pb : esign;;SupervisorJ g a At t a ch::ent F.EK /
.cc: .M.E. Lohnmann (1/0)
[ i G.S. Ger:en (1/0) A.R. Checca .(1/1) ~
- j M.f. Sears (1/1)-
.J.
Feinster (Westinghouse-Braidwood) (1/0); B* R
.N-6.
g,#
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ft
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y , , , .. 4 , +. 2f, , e ass
. Westinghouse Energy Systems - PinsNtr, Pevsyfvata 15?30 C35$ ' -
- Electric Corporation-CCEL90-324 November 30,91990 f
Ref. SM&RT-190(90), Mr, E. D. Swartz. dated 11/6/90-Commonwealth Edison Company - -(CCE-90-317)
~~
,1400 -Opus Place, .Su_ i te 400 ~
Downers Grove, Illinois 60515 , Commonwealth' Edison Company ~ y Braidwood Unit 1 . . i . Heatuo/Cooldown Final Reoort (Caosule U)' L p.
Dear Mr. Swartz:
Braidwood Unit 1 Heatup/Cooldown Final Report (Capsule U) was ~ transmitt'ed'. you via: Reference 1. ~ In response to' Robert Waninski's request,:the fol_ lowing. changes were..ma the' attached page: ~ 0 o The heatup curve is now identified as 100 F/Hr.. It was-inadvertentlyf 0 identified as.-60 F/Hr. - L o
- Acceptable operatfon" tag has been moved to a different location.
This is done as per the customer's request. - .i If you have any questions, please.do not hesitate to call'. Very.truly yours, ;; 4 E WEST INGHOUSE - ELECTRIC - CORPORAT I r N- '. h P.&Toth,: Y J Yh DbS;, Manager: M V- , G'. . ~ Commonwealth Edison Projects
' Domestic; Customer Projects- .
- - /cer. m ,
!9 cc: G. Gerzen R. Waninski y .1
'K Kofron:
M. Lohmann W. Feimster d U e .
. .1. _.
. c. ... >-
rr .: . g.p
...,- LMATERIAL PROPERTT aASIS
.s .
CONTROLLING'HATERIAL: CIRCUMFERENTIAL. HELD 40*F
. INITIAL RTNDT: ~
ART AFTER 32 EFPY: , 1/4T, 159'F 3/4T, 135'F .
...s CT '* . s : r. j . i
- y. . . < ,
Laak Test Limit J
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. ' '.re t,
Je . 1 e-i . I I I t t 's t r.
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. i e i . . .g. . j, , i. .
. i i i . . . . J. .f. , , .. .
. . .,. ri i .e . . . , ; , Acceptable --
Operation' . "I"",
*.%+.. , . . .. i e e i . i , eit i t i i . . . . i * * # t t , i eis a D p
e i i i i f. , f; . . . . . i i . . . ..
. , . . , . . .- . . 6 e i ., i j 0 .+ 4 6 e i 4 6 . i . .,
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, , . . 6 . . + . . . v t . e . i Fe ! 6 r i l e 6 6 i . . ! I i
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. . i . . . e ij j , . , , , . . . t i i
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ie 6 t e i *
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e ./ 6 .f i . . , . . . . . . i
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e . / f . . i , , . . . . . , . is O i il e 6 i 6 6 ei I . I ' I / 4 . . # e e a i * . .-
, , . A f . . . . i , i.. . . . .
Heatup Rates /. f . . . . . . .
.gp To ' ' ' ' ' ' ' '
. 1 1 -' ' .' . '. . .
a .. 1 F/Hr i t. r- . . . . . .
- :. i , i l. .00 i i ri n/ . . . . . . .
0 . 6 , e i 6 i! / i i I . 6 . . 6 / . .
. / . .
- / .
j , Criticality- Limit Based on
! . / .
-Inservics Prdrost'atic test
!sup. C504'F) for the' 5ervice
' ' ' Period Dp To.32.Errt.:
- .. . . i.ij i-
". 5; 100 ;50 2:; 25; 3cc -350- 400s ,450- 50t.
INOlCATCD TCwACRATURE.(DEC.F). 1 , Figure 1. .Braidwood Unit 1-' Reactor Coolant System Heatup Limitations
~
(Heat up rate up to 100*F/hr) App 11 Cable-for the First 32 EFPY-(Hith Margins 10*F and 60 psig for Instrumentation--Errors)- ! 08310:10/110590- 10 -, 1
^,- - -
L - m - . ,,J; :-
* ~
e" v Wh, ..... s'j
,, , a -t hheQ ? 'A . ,
r
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- *?
'4 ,
ny "' ~~ '
@ .g .Nov.27,1990' 'A
-In reply refe'rvto j CHRON No'. m To:-K.L i Kofron . .
a Braidwood Station Manager
'O
Subject:
7Heatup and Cooldown Curves forlBraidwood Unit 11[ LWestinghouse 11etter'CCE-90-317?. dated?Nov.19,3990. _ 4 5
Reference:
1 tWestinghousel has recen tly completed thelanalysishofLUnitD
' 1 Vesselesurveillance capsule,U. As,partiofLtheir;.: contract;s ,
o M they havefprovided.-heatup/cooldownicurves2to reflect theiflue s. e h issuin'glthe curves;uNED-PWR.cSystensiDesign-Group; discussed ( ' A , with the: Station Operating Engineer and' Tech. Staff thel._ ^ 4$y d' agreed that the heatup/cooldown rates = +
-proposed: curves an -should-remain the'same as:noted"on current Tech Spec 7curvest The reference heatup/cooldbwn curves %ound those! ~
currently in the ' Tech Spec, and therefore theistation' has: the option ofisubmitting<these curves for Tschespec: revision'or continuing to use the curves current 1p?in'the Tech Spec.; However,'the new curves also comply wlth NRC~ Regulatory: Guide:.. l
'1.99 rev42Lfor determining RTNn.c values at?l/4ToandL3/4T;.
r locations.'It must be'noted;thEt thecapplicabilityLdate of thepresentTech-Sgeecurves,to-Reg.cGuide1;99rev.2Eis:4.5 y time,rrevised; Unit;1 curves'aggtbeCincorporated11ntothe{ n
-6 Tech Spec..ThecUnit 2~ curves, which-have;ani.
been revised:to applicability.. comply 4with(. w s, date ofo2.2 EFPY, have alreadyf ' rev.2.of: Reg. Guide-1.99 andEare/ currently.being-processedtby.- : the Nuclear. Licensing Department for! Tech Speciincorporation.7 ;e
~
.NED recommends'theLattachedicurvesibU submitted 3as aC @C Tech-Spec change.intLorder to bring the Unit 111curvesointoL (compliance-with; revision 2 ? of Reg. Guide 1.99.- The schedulek _
- forcsubmittal should;be, determined by! Station /LicensingWItl ,
should be~noted.thatJan: evaluation'ofblowctemperature= overpressure protectioni(LTOP's)Jmust-beiperformed? prior"toj - implementing revised heatup/cooldown? curves.!ThisJisiaz separate; cost,Jno t, assoc iatedivith thefgenerationTof ,
-heatup/cooldown.:curvesEas partiof.the capsule analysisc ~.
Westinghouse has quoted alcostLofiS?27,650.009to; perform an] t P,'
<LTOPcevaluation.-The'avaluation'would beicomplt a e' 1 . , _
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~approximately one' month'from' issuance oft'a. work release.y
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'2 1 4 ItLahould be noted that there are two (2). typographical errors on'page 101of.the Westinghouse report.
0 F
;1'. 600 F/Hr'shown on the graph shall be. changed to 100 F/Hr, and
- 2. The words." Acceptable operation" shall,be relocated.across the criticality Limit line consistant with the revised Braidwood. Unit 2 heatup/cooldown curves.
Westinghouse has been advised.of'these comments and will submitas revised:page.accordingly.' Should you have any. question'or comments; please call; ' Bob Waninski at x-7387, Downers Grove.- ' /r Yse R.E. WaninsKi-PWR System emign in r: w E 'D . f] Swarta; ' PWR System Design Supervisor. Attachment-REW/ cc: M.E. Lohnmann (1/0) 4 G.S. Gerzen (1/0) A.R. Checca (1/1) M.F. Sears (1/1) J. Feimster (Westinghouse-Braidwood) -(1/0) juib c c. Crf.) p
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p7 . x' . 3 Westinghouse Energy Systems . Q34,,ng,gux3 ..*
- Electric Corporation CCE-90-317-November 19, 1990-Mr. E.-D. Swartz
" ' . Commonwealth Edison Company -
1400 Opus Place, Suite 400 ' ,' Downers Grove, Illinois 60515' Commonwealth Edison Company > t Braidwood Unit'1 . . Heatuo/Cooldown Final Report -(Caosule U)
Dear Mr,
Swartz:
^
?- Attached is the final report on "Heatup and Cooldown Curves for.the Braidwood ? Unit 1 Reactor Vessel (Capsule U)". Comments from Comonweaith Edison Company on the draft report-have been incorporated:in.the- final report.: , If'you have any questions, please do not hesitate to call. Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION.. k.
. P. Toth, DttS&
ager Commonwealth Edison Projects Customer Projects' Department-SAP /cem cc: R. Waninski G. Gerzen , K. Kofron , M. Lohmann W. Feimster
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