ML20115A372
| ML20115A372 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/06/1992 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20115A367 | List: |
| References | |
| NUDOCS 9210140265 | |
| Download: ML20115A372 (12) | |
Text
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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM s
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 snown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostati: testing with:
a.
A maximum heatup of 100*F in any one hour period, b.
A maximum cooldown of 100'F in any one hour period, and A maximum temperature change of less than or equal to 10*F in any c.
one 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 teinperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-cf-limit condition on the fracture toughness properties of the Reactor Coolant System; dettrmine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STAF08Y within the next 6 'ours and reduce the RCS T and pressure to less than a
200*F and 500 psig, respectively, within the ygliowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
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 system 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, at the intervals required oy 10 CFR 50. Appendix H,tu ausurdanse with the ;chedui:
---in-Teble 4. 4-5.
The results of thtse examinations shall be used to update Figures 3.4-2 and 3.4-1 l
SUMMER ~- UNIT 1 3/4 4-29 AmenomentNoS(
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r TABLE 4.4-5 REACTOR VESSEL HA!LRIAL SilRVEI11 ANU PROGRAM - WilllDRAWAt SCliE0 tit E 9
CAPSULE VESSEL LEAD WilllDRAWAL IIME, IDENTIFICA) CN LOCATION IACTOR EFPY j
U 343" 37 1st Refueling g
V 107*
3.1 3rd Refueling 287*
- 3. J Sth Refue1ing b
r W
110*
2.7 10th Refueling Y
290*
- 2. 7 17th Refueling I
340*
- 2. 7 siAMDBY J
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REACTOR VESSEL INTER. SHELL A9154-1 N</ I/
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j l/ L llllll i lllllllllll1 l ill/ if l/ li11llllll1 B1000 llIIIIl Iiii i If I l/ t /Ii I i ACCEPTABLE I I l l llIIIl' I _l l i i I / i AII I ll l OPERATION HEATUP RATES: i i l ii i i !I ]!!i I TO 50'F/HR ) l IM l/ I Jll l l l1{' T01c0*r/ Hah g l]lll}LI m g i i g lllTlli,Ill ll! i l i l li 'll< !I111111 1 { f. i l l l i.t l l l > lilli t l llBil ~l 111111 1 I lllllllll ll111ll1 li il l llIllil t h llllllll lilllllllil ill l l l l 11-1 .c l i l i l l l l l ' Ill 11 ll 11 I I l lll! li 100 200 300 400 AvtRAGE At ACTOR COOLANT SYSTEM 7tMM AAfuma t+#1 re,====== [ sumu unti i us 4.n Amenecentno.% f w W1 Y 7ted W f -
1 nE ACT02 COCLANT SYSTDs '''''IlilIIlIillI&IilllI I roo ' iiiiiiie ii I b IIIIIIIIIIIIII I Q
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~ !IIIIII y[ _!$ITI kT,h~=30'T IIIIIIIII IiIIIiiI IIIIiI'! tjV \\_ RT., um 10 trn: mi <0rr l 3/4T = BZ'F 1IlllIIIl lllll1l f ~ CURVES APPLICABLE FOR C00LDOWN Iil1lll!I{lIl1l\\lI RATES UP TO 100*F/HR FOR THE i;;; g;, 9;;;I;gg l -~ RVICE PERICO UP TO 8 EFPY AND C AINS MARGINS Of 10'F AND lil1ifilI lll1 lIl l 60 IG FOR POSSIBLE INSTRUMENT "~};ggg g ggg i g l Y '_' 2000 ERRO UNACCEPTABLE I '! II IIIIII ) lIIII l ( l l 1 I l\\0PERATION Ii ilIl[lIii\\\\llIIl { 3 l l I i l l I i hj i lIiI'tIII l1Il/llIIi ilIi!i I llllIIilA i I I I l l I I l l l l/ IillIlllilII , [I {liIIIIIi NIIIil i l l l [' IIlII\\iliIIi fy i j lIIII! lit l\\l1IIl11 I !/I l l i t l i l l i I I I 1 llM Iilil I/\\lIIIilllIllll IllIIIlll l ) Il111ll11 !Il Ailll /Illillli 111111 g(/ \\!IIllIl1' ll1IlhIl/ IIlllIl1.ltIl l / k 5 liI)IlIil l l I I I I IV llIlf ACCEPTABLE i l f ( ' OPERATION l T E '000 lIl)ll!llll\\l I I If k lIII l C00LOOWN RATES TO: l J III,llN lllIlli,1I Ifi h f l STEADY l I Yi!lll lll Niil l\\l\\\\\\ G i l i I I i 'h !II III!I [ 5 $ fg) ; @C [ % 4 l l\\ \\ \\M l I IIIlll SO'F/HR, ll ll ll b '100*F/HRr gl {jg,j gg{gggjg g gggggg II!!i l IljI I I I? lillllllll A llliti (r IlllIli lll jlll IlllIll )gjj N iill i III lIIl III IIII! !!!!I Il k\\ I I 0 0 100 200 300 400 ) AvtR AGE RE Acton coot. ANT SYSTEM TEMPERATURE l'M i a..eio, coo.em s, . Teme-== umiu Verwa Coo 4 deem Raise l -- _A J $ M R
- UNIT 1 3/4 4 3g Amenoment No.
i ~ REACTOR COOLANT SYSTEM 3/4.4 9 DRES$URE/TEuPERATURE LIMtTS GEACTOR COOLANT SYSTEM LIMITING CONC.' TION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and i pressure snail be limited in accordance with the limit lines snown on Figr'es 3.4-2 ano 3.4-3 during heatuo, cooldown, criticality, and inservice leak and hydrostatic testing with: a. A maximum heatuo of 100*F in any on hour period. D. A maximum cooldown of 100*F in any one hour period, and A maxiinum temoerature change of less than or eoual to 10 F in any one nour period during inservice nyarostatic ana leak testing operations above the neatuo ano cooldown limit curves. APPLICABILITY: At all times. ACTION: With 3ny of the above limits exceeoed. restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluatior, to determine the ef fects of the out-of-limit condition on the fracture toughness properties cf the Reactor Coolant.5ystem; determine that the Reactor Coolant System remains acceptable for continued cperation or be in at least HOT STANOBY within the next 6 hours ano reduce the RCS T and pressure to less than 200*F and 500 psig, respectively, within the*Y811owing 30 hoe s. SURVEILLANCE REOUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to De 'within the limits at least once per 30 minutes during system heatup, cooldown, ar.1 inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irraciation surveillance specimens shall be removed and examined, to determine changes in material properties. at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. 1 SUMMER - UNIT 1 3/4 4-29 Amenoment No. --m s u -*F
e. s -) l I - I -l i i I' THIS PAGE IllTENTIONALLY LEFT BLANK f I r l s c SUMMER - UNIT 1 3/4 4-30 S y sy y m.,n~g y- +-w- -r ~ 9-e m g= --~r+www. f wrr r e vv-1--t---v-n g ..w -wma-ed w e=we--- -w-+- a-+ew-=.--A - we 5 ie w-.-e- ., =-
. -REACTOR COOLANT SYSTEli MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: LOWER SHELL INITIAL RTNDT: 10*I ART AFTER 14 EFPf: 1/4T, 96*F 3/4T, 83'F 2 5 0 0 rt.:.m v a, i, ii is i ! I i ii,i i, i i-I i i ' 6 r! I i ! !. 'l - 8 i i I. i ,i i s ! ! i i i t i i ,
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- i 6
i,,! i I I i i ! f 'ii! I t i iii t i i ! ! ! t i !i 1 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEo.r) Figure 3.4-2 V. C. Sutimer Unit 1 Reactor Coolant System Heatup Limitations (Heat up rates up to 50 and 100*F/hr) Applicable for the First 14 EFPY (With Margins 10'F und 60 psig For Instrumentation Errors) SU MER - UNIT 'l 3/4 a-31 x _m.. m
REACTOR COOLANT SYSTEM MATERIAL PROPERTY BASIS CONTROLLillG MATERIAL: LOWER SHELL INITIAL RTilDT: 10*F ART AFTER 14 EFPY: 1/4T, 96'F 3/4T, 83'F 2500 ra m aa, i i j iii i, i i i i i i i, i
- ; i i!
i i i i r i -- i i i i i i / i, i F i i , i 2250 I i i i
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, 1500 ! i I! o /1 i ! i i ! i i !,i,; ,i,, i i ri i i, i i. ,,,, ii;,i;, i; m L i' i i t / e i i . i 4, i, i i, i i i ,i l! !Y !.!!i 1250 !.4 i, '/, cr ! ii! ! t i i i i i 4 4 i
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d i ' ! '/' ' ' E 4 0 0 0 -i ' W' ' ! ' ' i ' i i sii,! i.,,' e i i, ' i ! ! i i I i !., i i i i i i i ,, i, i, o ' ' ' i I i i j i3 ; i, i 4 i i u i ' ' i yr i i 4 i i ; a i i , i, i i,,, y ? i i ' ! i i ' ' u 750 --COOLDOWN RATES i l ll, ll ,! ',',',i ',' !,' iJ E z !,'F/HR iei ii O ii ,i i i 4,,, 4,,,,,,, !J. 25 i i i. i i 6,., i i i i,, i ! i 500 50 W-i 'l i i ' i i ee i ! i 4..i l 100 -r ' t i i i i I I ' i ' .'.' i e 'i,' ! i i 4 ! i e i iiiie i 6 i ! i!,,, 4 i,;,3 , i i ,iie i i ( j , I i, j,, ; , ; ;, j,, j 250 ,' i,! !, i ' ' ' ' i 4 6 i i i i ! ! i ri i i i, i i i i i , i4 ,, i ,i, ,,i 3 : i i i. >.., ,ii,,,, 4 i,44 .,,,i, i
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!'f ? 8 ' ' 4 i i 1 i '* i e i i is i e i, 9, j 1 ! ! ! * * > ! !, s i ! ; i i, ,.,ei;,i e 9 0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEC.F) Figure 3.4-3 V. C. Summer Unit 1 Reactor Coolant System Cooldown (Cooldown rates up to 100'F/hr) Limitations Applicable for the First 14 EFPY (With Margins 10'F and 60 p;ig For Instr _.1entation Errors) SUMMER - UNIT 1 3/4 4-32
. _... = _.. _ _ . to Document Control Desk Letter ISP910005 Page.1 PROPOSED TECHNICAL SPEClf! CATION CHANGE - TSP 910005 - VIRGIL C. SUMMER NUCLEAR-STATION LIST Cr AFFECTED PAGES Pragg Spgcification - Description of Chances 3/4 4-29 4.4.9.1.2 Deleted reference to the Reactor Vessel Material Surveillance Program i Withdrawal Schedule 3/4 4-30 Table 4.4.3-Deleted Table 4.4 Page intentionally left blank 3/4 4-31 Figure 3.4-2 Replaced with new curve 3/4 4-32 figure 3.4-3 Replaced with new curve J-t 1 - i { =., - - -, = -e- ,. +, - -ey e w- ..-p..y, y,-y -r,- ,1-g y 2,
. to Document Control' Desk letta 'iSP910005 Page 1 of, PROPOSED TECHN; CAL SPECIFICATION CHANGE - TSP 910305 VIRGIL C. SUMMER NUCLEAR STATION DESCRIPTION AND SAFETY EVALUATION DESCRIPTION OF AMEWuMEh! REQUEST SCE8.G proposes to modify ihe VCSNS TS, Section 3/4.4.9, Figures 3.4-2 and t 3.4-3, to provide new PT cerves consistent with analysis results of v imination of specimen X of VCSNS D diation Surveillance Program. Tabl-4.4-5 and the reference it in Sur.eillance Requirement 4.4.9.1.2 are deleted in agrement wi- .e guidance provi u e Generic Letter 91-01. SAFETY EVALUATION Yhe pr-hange to the PT curves reflect 'ne resu'its of the analysis o perfcee 1 wecimen X, and calculaticns prepared using the guidance of Regulate, uuide 1.99, Rev. 2. Radiction Embrittlement of Reatt_or Vessel Materials, and Appeadix G to 10CFR50, " Fracture Toughness Requirements." The new PT curves continue to provide conservative t.dmiaistrative restrictions or,the RCS pressure to minimize stresses on the RCS due to ncemal operating tr6nsients, thus minimizing the likelihood of brittle fracture due to pressure transients at low temperatures. Deletion of the schedule for removal of the reactor pressure vessel material surveillance capsules from the VCSNS Technical Specifications does not impact 4 the safety of the plant. The schedule is controlled.by the requirements of 10CFR50, Appendix H, cnd the schedule will be included in a future revision of the FSAR.
. to Document Control Desk Letter TSP 910005 Page 1 of 2 PROPOSED TECHNICAL SPECIFICATION CHANGE - TSP 910005 VIRGIL C. SUMMER NUCLEAR STATION DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION DESCRIPTION OF AMENDMENT REQUEST SCE&G proposes to modify the VCSNS TS, Section 3/4.4.9, figures 3.4-2 and 3.4-3, to provide new PT curves consistent with analysis re.91ts of examination of tpecimen X of VCSNS's Radiation Surveillance Program. Table 4.4-5 and the reference to it in Surveillence Requirement 4.4.9.1.2 are deleted in agreement with the guidance orovided by Generic Letter 91-01. BASIS FOR DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION SCE&G has evaluated the proposed TS change and has determinao that it does not represent a significant hazards consideration, based on the criteria established in 10CFR50.92(c). Operation of VCSNS in accordance with the p oposed action will not: (1) Involve a significant increase in the probability or the consequences of an accident previously evaluated. The proposed change provides up-to-date pressure and temm reture limits for operation of the reactor coolant system during heat up, cool dowr., criticality, and hydrotesting, thus protecting the reactor v ssel fron, brittle fracture by clearly separating the region of acceptable operation from the region where potential brittle frt:ture of the raattor vessel may occur. Failure of the reactor vessel is not a VCSNS design basis accident, and, in general, reactor vessel failure has a low probability of occurrence and is not consida*ed in the safety analysis. The new P1 curves will provid' edditional csaservatism, making the reactor vessel failure an even less credible event. Deleti^n of the schedule for removal of the reactor vesse) material y surveiilance capsules from the Technical Specifications and inserting it in the FSAR is administrative in nature, since the schedule is a duplicate of the 10CFR50, Appendix H, requirements. g (2) Create the possibility of a new or different kind of accident from any previously analyzed The proposed change does not introduce a plant design change or a new operating. procedure. It simply adjusts the PT curves to reflect the shift in nil-ductility reference temperature of the reactor vessel due to neutron ir radiation.
v_ ,4 7 - l -Enclosure 3 toLDocument Control Desk Letter- -TSP 010005 Page~2-Deletion of the schedule'for.1coval of t. reactor vessel niaterial surveillance capsules from the Technical Specifications and inserting it in the FSAR is administrative in nature,_since the schedule is a duplicate of the 10CFR50, Appendix H. requirements. (3) Involve a significant reduction in a margin of safety. The new PT curves ensure that the 10CFF.50, Appendix G, requirements _are not exceeded during normal operation including reactor coolant system transients during heat up, cool down, criticality, and hydrotesting.- The new PT curves were prepared for a projected reactor vessel exposure of 14 EFPY. The new curves shift to more conservative limitations,Lthus. H providing increased margin against non-ductile fractures.' Since L administrative limits remain in place to ensure that 10CFR50, Appendix l G limits are not challenged, the margin of safety specified_in the TS Cases is not s1gnificantly reduced by the proposed change. Deletion of the schedule for removal of t'*c reactor vessel material surveillance-capsules from the Technical 7pecificatic~- snd inserting it in u.: FSAR is administretive in nature, since-the r,chedule is a duplicate of.th! 10CFR50, Appendix H, requirements. r n}}