ML20077E971
| ML20077E971 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/04/1991 |
| From: | Cottle W ENTERGY OPERATIONS, INC. |
| To: | NRC |
| References | |
| NUDOCS 9106110325 | |
| Download: ML20077E971 (210) | |
Text
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O E= Entcrgy '"'"" " " " * " * ' " * -
Operations "
l W. T. Cottle June 4, 1991 It.S. Nuclea r Regu lat ory Gomm iss ion lh f l St a t. f on PI- 13 /
W a n h i tigt ot t , D.C. ;'0555 Atiention: Doc o tiie ti t' Con t t'o l Desk.
Sohject Grnnd Guli Nonlear Stat ion Unit 1 Dockot No. 50 416 I. i c etm n No, N P P- ;"8 Report of 10FFi&0.59 So f ot.y Evnlunt ions - June 1, 1990 through Decembnt 31, 1991 GNRr).o I /00001 Gont lemen:
.i n a cco r < lance wit h t he requirements of .10C FR 5 0. 5 9 ( li) . Entergy Opei n t ions ,
Inc. is report itig t hose changes, tests, and e<poriments under t he retin i r nt 4 nn t s of 10CFR50. 59 f or the p9r lod of .huie 1, 1990 t hrough Pnennber 31, 1990. A sommary of these chnneos, tesis, and exper imnrit s f r.
cont a llied (ti the a l.t a chmen t .
!L has l'oen t he pr o r.t I ce of l'.h t e r gy 'Ipo ral lotis , lor, to sohmf; the lOGFP90.59 rr pori n sem i-atinnq l ly . I n a ct_.o nin oc o with l o(TR;0. 5'l( h ) ,
Ent e rgy Opi . n t f ons , l u e. . will in t ho future s ubm i t. the i oport s on on innuti l bq s it. . Annual toports covering tho nn f ot y ovninationa for CGh5 l ar anch ro le t h r year wi11 l>e < nhoi t Li ed pr f or to . inly 1st of the fr ilow fne ya ,r.
i Ven r< t!nly.
A? I Mwq WTr/GWP/nms att nehment
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Gr I n s111/ ssi,1CI'f,P - 1 '( p 91061103;;5 910604 i
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June 4, 1991 GNRO-91/00001 Pagn 2 of 3 cc: Mr. D. C. Illntz (w/o)
Mr. J. L. Mathis (w/o)
Mr. R. B. McGeboo (w/o) ,
Mr. N. S. Reynolds (w/o)
Mr.11. L. Thomas (w/o)
Mr. F. W. Titus (w/a)
.Mr. Stewart D. Ebneter (w/a) i Regional Administrator l ti.S. Nucinar Regulatory Commission .
Regfon 11 101 Mariatta St., N.W., Suite 2900 At lonta, Georgia 30323 Mr. L. L. Kintner, Project Manager (w/a)
Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 11D21
- Washington, D.C. 20555 i
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June 4, 1991 GNRO-91/00001 Page 3 of 3 bec: Mr. R. W. Ilyrd (w/o)
Mr. l.. F. Daughtery (w/o)
Mr. M. A. Diet rich (w/o)
Mr. J. O. Fowler (w/o)
Mr. W. K. Ilughey (w/o)
Mr. C. R. Ilutchinson (w/o)
Ms. F. K. Mnngan (w/o)
Mr. M. J. Moisner (w/o1 Mr. G. W. Muench (w/a)
Mr. D. L. Pace (w/o)
Mr. T. C. Reaves, Jr. (w/o)
Mr. J. L. Robert son (w/o)
Mr. G. W. Rogers (w/2)
Mr. M. J. Wright (w/o)
Mr. G. A. Zinke (w/o)
File (LCTS) (w/a)
Filo (llard Copy) (w/a)
File (RPTS) (w/a)
File (NL) (w/a)
File (Central) (w/a) ( 215 )
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G9105311/SNI.1CFl.R - 3 J
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TABII 0F CONTENTS OF 10CFR50.59 SAPETY EVAL.UATIONS FOR Tile PERIOD JUNE 1 1990 Tl! ROUGH DECEMDER 31, 1990 i
SRASN DOCUMNT PAGE NPE-90-022 DCP-88-0005-800-R00 1 NPE-90-023 MNCR-90-0183 2 NPE-90-024 DCP-85-4007-S00-R00 4 NPE-90-025 HCP-89-1042-S00-R00 6 .
NPE-90-026 HCP-90-1079-S00-R00 7 .
NPE-90-027 CN-90-0105 8 NPE-90-028 DCP-84-0149-S00-R00 9 NPE-90-029 DCP-88-0042-S00-R00 10 NPE-90-030 Calculation NPE-E22F004 11 ,
NPE-90-031 EERR No. 90-6162 12 NPE-90-032 CN-90-0125 14 NPE-90-034 Calc. EC-Qll.21-85001, R02 15 NPE-90-035 CN-90-0185 17 .
NPE-90-036 W. O, 19998 18 NPE-90-037 CN-90-0182 19 NPE-90-039 EER-90-6228 20 NPE-90-040 Calc. MC-QlE30-90112 21 NPE-90-041 DCP-82-0056-S00-R00 22 l NPE-90-042 DCP-82-4178-S00-R00 23 NPE-90-043 DCP-84-0250-800-R00 24 NPE-90-045 DC P- 85-4051 -S 00- R01 25 NPE-90-046 DCP-86-0073-800-R00 26 NPE-90-047 DCP-87-0034-S00-R00 27 NPE-90-048 DCP-87-0048-800-R00 28 NPE-90-049 DCP-88-0027-800-R00 29 NPE-90-050 CN-90-0318 31 NPE-90-051 DCP-88-0029-S00-R00 33 NPE-90-052 DCP-88 356-S00-R00 34 NPE-90-053 DCP-88 o057-800-R00 35 NPE-90-054 DCP-89-0343-S00-R00 36 NPE-90-055 DC P- 89-034 3-S01- R00 38 NPE-90-056 NEPFSAR-89-0041 39 NPE-90-057 DCP-90-0005-S00-R00 41 ;
NPE-90-058 DCP-90-0060-500-R00 42 NPE-90-059 QER-323-39 45 NPE-90-061 MNCR-90-0032 46 NPE-90-062 DCP-90-0344-S00-R00 47 NPE-90-063 DCP-90-0547 49 NPE-90-064 FSAR-CR-90-0032 51 NPE-90-066 CN-90-0101 52 NPE-90-067 SERI-JS-08 53 NPE-90-068 MNCR-0124-90-R02 54 A900129. TOC /SNI,1CFER - 1
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TAPI.E OF CONTENTS OF 10CFR50.59 SAFETi EVAEUATIONS FOR Tile PERIOD JUNE 1, 1990 THROUGil DECEM8ER 31, 1990 SRASN DOCpMENT PAG.f.
NPE-90-069 DCP-90-0551-80 & SI-R00 59 NPE-90-070 CN-90-0391 68 NPE-90-071 MCP-89-1098-S00-R00 69 NPE-90-072 MCP-89-1102 71 NPE-90-073 MCP-89-1103-800-R01 72 NPE-90-074 MCP-89-1135-S00-R00 73 NPE-90-075 MCP-89-1126-800-R0 & R1 74 NPE-90-076 MCP-90-1004-800-R00 76 NPE-90-077 HCP-90-1001-S00-R00 78 NPE-90-078 MCP-90-1017-S00-R0 & R1 79 NPE-90-079 HCP-90-1020-S00-R00 81 NPE-90-080 MCP-90-1042-500-R00 82 NPE-90-081 MCP-90-1054-800-R00 83 l NPE-90-082 HCP-90-1064-S00-R00 84 !
NPE-90-083 MCP-90-1055-S00-R00 85 NPE-90-084 MCP-90-1056-S00-R00 86 NPE-90-085 MCP-90-1063-800-R00 87 NPE-90-086 HCP-90-1073-S00-R00 88 NPE-90-087 MCP-90-1097-S00-R00 89 NPE-90-088 MCP-90-1098-800-R00 90 NPE-90-089 MCP-89-1112-S00-R00 91 NPE-90-090 NPEAP-807, 320, 332 92 l NPE-90-091 NPEFSAR-90-0044 93 NPE-90-092 MNCR-89-00293 95 NPE-90-093 NPEFSAR-90-0056 96 NPE-90-094 NPEFSAR-90-0021 98 NPE-90-095 CN-90-0268 99 NPE-90-096 NPEFSAR-90-0042 100 NPE-90-097 EER-90-6388 101 NPE-90-098 Enginonring Report GGNS-40-0028-R00 10? j NPE-90-099 EER-90-6231 105 ;
NPE-90-100 MNCR-90-0176 106 i NPE-90-101 EER-90-6385 108 Nite 102 MNCR-90-0093 110 NPE-90-103 EER-90-6401 111 NPE-90-104 EER-90-6417 112 NPE-90-105 CN-90-0523 113 NPE-90-106 CN-90-0537 116 NPE-90-107 Ops w/o Purge Flow to Rnactor Recirc. Pump 117 NPE-90-108 EER-90-6466 118 I
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TABI.E OF CONTENTS !
l 0F 10CFR50.59 SAFETY EVALUATIONS FOR TIIE PERIOD JUNE 1, 1990 TilROUGil DECEMBER 31, 1990 SRASN []OCUMENT PAGE Pl.S-90-011 UFSAR Appendix 3A 119 Pl.S 012 FSAR C/R 90-0005 120 PLS-90-013 TSTI-1G17-90-003-0-S 121 PLS-90-915 OQAM FSAR 17.2 122 PLS-90-016 04-1-01-N19-1-TCN 25 123 PLS-90-017 W.O. #00014194 124 PLS-90-018 Deinting Operator Actions 125 PLS-90-019 TST1-1E51-90-002-0-S 127 PLS-90-020 FSAR C/R 90-0008 128 Pl.S-90-021 W,0. 6185-Alternate Fire Water Supply 129 PLS-90-022 UFSAR 7.7.1.11.4.2.b 131 PLS-90-023 UFSAR C/R 90-010 132 PLS-90-024 01-S-06-2 135 PLS-90-025 MWP-90-1151 136 PLS-90-026 DCP-88-0051 137 ,
Pl.S-90-027 CR-90-011 138 PLS-90-028 TS 3.0.4, ACTION c 139 PLS-90-029 CR-90-006 141 PLS-90-030 Temp Alt 90-0004 142 PLS-90-031 W.O. 27751 143 PLS-90-032 MWO 26063 144 PLS-90-036 MWO 26064 146 PLS-90-043 TST1-1G17-90-004-0-S 148 PLS-90-044 W.O. #29996 150 Nf.S-90-002 SERI Operat ions Manuni to 151 Operations Management Manuni l N!.S-90-003 GGNS Emergency Plan Section 6.6.S6 152 NLS-90-004 TS 3.7.2, Action b.1 153 NLS-90-005 TS 3.6.4, Actions b & c 155 NLS-90-006 TS 3.6.6.2, Actions b & c 159 I NLS-90-007 TS 3.4.9.2. Act ion a 161 NLS-90-008 TS 3.7.1.1, Action b 163 N1.S-90-009 TS 3.4.9.2, Action b 165 NLS-90-010 TS 3.6.4, Actions b & c 169 NLS-90-011 TS 3.5.6.2, Actions b & c 172 N!.S-90-012 CR-NL-90-009 174 NLS-90-013 TS 3.9.11.2, Action a 175 NLS-90-015 CR-NL-90-006 179 l
NLS-90-016 TS Position Statement 128, R0 180 I NI,S-90-017 TSPS 128. R00 182 NLS-90-018 UFSAR CR-NL-90-014 184 4 NLS-90-019 OpCon 4 Entry While in TS 3.5.3 185 l
A900129. TOC /SNI.lCFI.R - 3 l
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l TABl.E OF CONTENTS DF 10CFR50.59 SAFETY EVALUATIONS FOR Tile PERIOD ;
JUNE 1, 1990 TilROUGil DECEMilER 31, 1990 SRASN DOCUMENT PAGE j NLS-90-020 OpCon 4 Entry While in TS 3.3.6 and Bases 188 NLS-90-021 OpCon 4 Entry While in TS 3.3.7.5 190 NLS-90-022 UFSAR Appendix 13A 191 NSP-90-003 Change of Executive Director, Operations 192 Support to Vice President, Operations Support NSP-90-004 Onsite Storage of New Funi for GGNS Cycle 5 193 NSP-90-005 RF04 Fuel Management 196 NSP 90-006 Refueling Operations with Revised Coro 198 Loading Plan NSP-90-007 UFSAR 15.5.1 200 NSP-90-008 Cycle 5 Ops with Revised Corn 201 Configuration NSP-90-009 Cyclo 5 Opn with 9 x 9.5 Reload 202 1
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Attnehment tn GNRO-91/00001 SHA3N: NPE-40-022 DOC No: pCp-88-0005-800-R00 SYSTEri: G36 DESCRIPTION OF CilANGE- This change replaced the Reactor Wat er Clenn-Up (RWCll) resin metering pumi,with a new larger capacity pump cautpped with n flow monit orieg sight glass. A new piping and backwash system was also installed.
REASON FOR CllANGE: Tae previous peristaltic pump experienced frequent ruptures of the Tygon Tubing, and was not of adequate capacity. The new RWCU resin metering pump provides higher capacity and greater flexibility with precont resin injection rnt es. The new system includes backwash capability on the resin oump suction and discharge lines for clenning purposes.
3AF TY EVAI.UATION: The so f et y evnlunt ton concimled t hat tbe cb' ge did not involve an unr eviewed safety quest lon. The pump, p . , '. n g , nupports, and assor.inted equipment replaced by this change are non-safety reint'd, seismic Category 11/1. This change meets all requiremants of the or iginal RWCU resin metering system and does not compromise any sa fet y related systems or component s or prevent a safe reactor shutdown. No safety relat ed circuits or interfaces are added or affected by this change. Opnrnt ion of the RWCU resin metering pump and associnted equipment are not required to mitigate the consequences of an accident..
The RWCU resin metering pump or associated equipment. nre not addressed by the GGNS Technicn1 Specifications, nor does any of this equipment impact the margin of safety of any syst ems addressed in the Technical SpecifIcat icas.
NPEuo/SNhlCil,R - 1
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Attachment to GNRO-91/00001 SRASN: NPF,-90-023 DOC NO: MNCR-90-0083 SYSTEM: E22 DESCRIPTION OF CilANGE: This evaluation identified a condition in which the accident load profile for the Division 111 batteries exceeded the proflie described in FSAR Tabin 8.3-8.
The current load profile .i s :
276 amperes for the first 60 sec., ,
216 amperos for the next 59 min.,
218 amperes for the last. 60 min.
As a result of ef forts to review t.he design basis of the eInctrical systems, the Division 111 battery load profile was '
revisited and installed as-built loads were calculated. The resulting minimum required test profile, with margin built in, was calculated to be:
c65 ampares for the first 60 sec.,
220 amperes for the next 59 min.,
220 ampores for the last 60 min.
REASON FOR CilANGE: Calculat lons were performed to ensure t he batteries capacity to deliver the energy. These calculations use the methodology presented in IEEE 485-1978 and shows that the existing bat.tery is sized adequately to deliver the required energy.
The FSAR change is a change to the profile presented in FSAR Table 8.3-8 and section 8.3.2.1.7.2.
SAFETY EVALUATION: This safety evaluation concluded that the change did not involve an unceviewed safet y question. The change being' evaluated is a revision of t he load profile to reflect the actual emergency loads imposed on the Division III batteries.
This change does not reflect a physical hardware chnnge to the facility, but imposes the proper requirements on the existing batt ery system. Calculations have been performed in accordance with IEEE 485-1978 which ensure the capability of the existing hattery banks to meet the newly calculated load profile. This industry standard is the governing document. for determination of bat.tery sizing. Compliance with t his standard ensures that the battery system can perform its intended funct lon.
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Attachment to GNRO-91/00001 NPE-90-023 Page 2 The lond prof fle was developod f rom the postuinted design basis a cc id en t scenario and the capability of the battery bank was determined using IEEE 485-1978 battery capacity methodology. This ensures that the new profile is greater than the ac* uni emergency load and that the installed battery system is sized properly to carry the load. No change is being made to the installed plant hardware and the capabilities of the existing hardware to meet the design requirements have been verified.
With the imposit ion of the lond profile specified, additionni margin to actual accident load profiles has been included. This additional lond value has been verified to be within the capnbilities of the installed hardware and above the load imposed on the batteries by t he postulated accident. scenario. The margit.
provided by the current technical specification surveillance has been increased by the profile provided. Therefore, the margin of safety as defined in the basis for any technical specifiention will not be reduced.
NPE90/SNhlCFhR - 3
Attachment. to GNRO-91/00001 SRASN: NpE-90-024 DOC NO: DCP-85-4007-800-R00 SYSTEM: N71 1
1 DESCRIPTION OF CHANGE- This change provided the design change necessary to install a Condensor Tubo Cloaning System (CTCS) on
, the Circulat ing Water (CW) system which provides an on line cicaning method for the condenser tubes. The operational principle for the CTCS is to continuously inject sponge cleaning balls into the CW flow on the inlet aido of the Lp condensors and to collect the cleaning balls from the CW flow on the dischargo side of the llP condensors. The clnaning balls are desigund to be randomly distributed throughout the CW Flow. A cont.rol panol will provide annunciation in the control room for specified CTCS malfunctions. The CTCS control panel, recirculation pump, and ball collector tank are to be located in an area of the Turbino Building which is accessible during normal plant operation.
Prior to operation of the CTCS, the condenser tubes worn c1 caned to removo excessive tubeside fouling, in orcer to achieve free ,
tube passagn for the CTCS cleaning balls. The tubes were cleaned during thn second refueling outagn by implementing a NALCO chemical cleaning process which requires the circulation of tannin solution, sulfur.ic acid, citric acid, and iron dispersants through ,
the tubes. NALCO representatives provided continuous coverage during the cleaning process. The cleaning process has been laboratory tested bv Ent ers;y to ensure that the process is benign to the materials of construction used in the condensers and CW piping components. The waste water generated by the process was transported to thn Unit 2 cooling tower basin for storage until such time as approval had been granted by the Mississippi Dept. of l Natural Resources for discharge t:) the environment. CGNS l
Operating License Con ('ition No. 2.c.(27) contains a provision which prohibits f!lling the Unit 2 cooling tower basin. The NRC hns been contacted regarding the discharto of waste water into the Unit 2 cooling tower basin. The NRC respese stated that tho operating 11censo condition does not apply since Unit 2 is not. in
- operation.
REASON FOR CilANGE- This changn provided for an automat.ic CTCS on -
t.hn CW system to prevent fouling and thus reduce flow restriction and improve heat transfer.
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Attathment t o GNRO nl/00001 Npn-90 0?!.
pnge ?
S APl"lY IWAl.UATION: 1 hei o is no inciense in t he pr obnbilit y of occuttence or in the consequentes of an nccident or malf unct ion of equipmen' importnnt to soIety ptnyionsly evntunteil in tbe Sniety Annlysis Kepost. CW systetti acc ident s pret iously cyninnt ed in the ISAR nie l i tn i t ed to the potentini flooding of snfety teinted equipment due t o t he f allut e of n CW syst em cornponent . Two ponnible r.ources of CW nyst em f ailuie hnve been previously ident ille d as: 1) failure of the expansion joints, nint 2')
f ailure o f the but t er f ly vnives. CTCS installation will nihl nn adslit ionni sour < o by t he instalIntton of flanged stininct sections on the CW piping diruhntgn lines i n s irle the Tuthine building. The finngs d connec t ions nt e designed t o wit hst nnd 95 PSIG which en ceds the 90 PSIG des ign pi ensuie of t he Condenset s atul f ni exceed' the CW pump mnximum shutoff piesnute o f appt ox iturit e l y 66.1 PSIG. 1he ndded weight of .he CTCS stininer sections does not exceed the nderpincy of t he ex is t ing suppoi t s pe r the applicnble ca lculat lan. The C1CS reiliculation piping is designed p%: ANSI 101.1 Powe r Piping Code sequirements t o wit hst nini 150 PSiti to cofrjiells n t e Ior' t he ni$ fled pr ensul e i equiI ett t o I e liij e0 t t lie clenning balls into the CW flowpnth at the 1.P cotulen ser inlet.
CTCS inst n1 Int f on does not inn n11 new snfety ieinted epilpment, n1t er t hn lornt ton of exlat lug sa f et y ieIat ed equipment , on add volume to the CW sy5 tem fluid invent ory . Irnplement at lon of Ihe design < hnnge wil1 piovide nei enhain einent to tbe CW sy8teins abilit y t o maint niti design bu kpressuies inside t he inn in cotidef tse r s .
The CW system serves no safety function. Syst ems nualysis hns shown that f ailuie of the CW syst em will not compromise nny snfety-telated systems or prevent safe reactor shutdown, inst nlint ion of the C1CS will enhance the nbility of t he main rondenser t o rna lnin in des ign bn< k pi essut o niul will enhnnce the reliability of the t ornienser t ubes by reducing t he poss ibilit y of pit t ing rot ios f or . linplement at ion of t he change will not
'dversely affer.t the f unct innnl < har nct er ist its of the CW system.
Ch ;, st em component f luid bonininn y f ailur e has beeti previously cyni inted in the I SAR and no afblit innal mndes of Inflore are pos t i ' n t ed . "I h o i e f o i e , there is no cient ion of a pocs ibilit y for an n< ident or ic<il f unct inn of n d i f f erent type than any evnlunted pr ovi >ua 'y in the Sa f et y Ana lys is Report .
The do<.ign bnses foi the CW system ns defined in the GGNS Tecbolcal Specifications does not contnin piovisions foi nny specified margins of sn f et y t enniding t he fnllure of n C7 system corn pone n t . Therefoto, ivrplementation of the design change does not reduto the mntnin of safety as defined in the basis foi any Technical .Npectficniton.
hee 90/SNblCil.R - 5
Atinchment to GNRO-91/00001
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SRASN: NI'E 02 5 DOC NO: HCP-89 1042-800-R00 SYSTEM: G17 )
t DESCRIPTION OF CllANGE: This change removed the legend pinten fiom the listed annunciator windown and replaced them with binnk '
i mylnin. The alarm cards woro permanently pulled and no noted on j all of thn nanociated drawings. j Liquid Radwnsto Pf it ern Tabl. Ann./SG17-UA-L602 !
Floor Drain Wanto Evap. Trbl. Ann./SG17-UA-L604 !
Solid Radwnsto Syn. Trouble Ann./SG18-UA-1.600 ;
CNDS Coll. Tk, Level liigh-liigh Ann./SN12 I.Allli L610 CNDS Rin. Stg. Tk. Level Illgh Ann./SN12-LAll-L658 Chorin. System Trouble Alarm Ann./SN72-9Allh-L600 ;
Hokeup Wtr. Tttet. Syn. Tbl. Ann./SP21-UA-L601 i
REASON FOR Ci!ANGE: The annunciatorn listed are connected to f non-nnfety rnlated equipment which in not being utilized with the exception of the IJquid radwnste nintms. Ther n alarms arn f provided in the Radwnnte Control Room nr.d are not required to hn l in t he Hain Cont rol Room, Theno annuncintors are not required por i IEEE 279 and there in no requirement for the annunciatorn for equipment protection. The alarm cards were removed under an l Operations Nulannce Annunciator Ptogram. L i
f l SAFETY EVALUATION: The narcty evaluation concluded that the I
chnnge did not involvo an unroviewed anfoty question. Thenc [
nnnuncit.toin serve no safety function or support equipment important to anfoty. Unilure of thenn annunciators will not 7 compromine any nnfety related system or component and will not ;
prevent nnfo reactor shutdown. Thero in no probable accident t annociated with thin equipment. The annunciators are not '
connect ed t c. equipment which in related to any plant unfety funct lon. Disabling thcan annuncintors createn no new failurn ;
modon not nircady enveloped by prr,nont TSAR analysis.
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Theno nnnunciators are not addressed in any Technical Specification nor are they essentini in monitoring the plan
- for I compilanen with the Technical Specifications. !
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SWASN: NPE 90 026 DOC NO: HCP-90 1079-S00 R00 SYSTEH: E22 DESCRIPTION or CilANGE: Hodifications were made to cortain power circuits to reduce voltage drops within long power and control circuit runs. Spara conductors within some power cables were utilized for a parallel feed on the posit ive lead to reduce tho voltago drop on theso circuits. For contial circuit ICA701. spara conductors of exist ing Division 3 cables were ut ilized to roconfiguro the lipCS Diosol Generator Breaker 152-1701 autoclose circuit in order to climinate an excessivnly long control circuit route. Also. certain conductors within the control circuit cableis woro paralleled to further aid in voltage drop reduct fon. Spare conductors withie existing Division lit cables were utilized to ensurn both divisional separation in accordance with Reg. Guido 1.75 and proper cablo qualification. Thn conductors utilized are of adequato ampacity for their application.
REASON FOR CllANGE: Haterial Nonconformance Report 0083 90 identified certain Division 111 125 VDC circuits whose devices may not receive manufacturer's minimum voltage values during a Design liasis Accident. (DBA). The deficient voltages have been at t ributed to voltage drops within long power and cont rol circuit runs.
SAFETY EVALUATION: The safety evaluation concluded that the change did not involyn an unreviewed safety question. No system function has been altered and no nov equipment was installed.
Only spare conductors within existin, Division 111 cnbles were utilized to ensure both divisional sogaration in accordance with Reg. Guido 1.75 and proper cabin qualification. Since proper separation is maintained, a failure in Division 111 circuits cannot propagate into another safety system thus limiting the failure to Division 111. The conductors utilized are of adequate ampacity for their applicat ion. .
l This design change dons not affect the ilPCS system in consideration to items addressed in GGNS Technical Specifications such as flow, chemistry, antpoint. capacity, levnl. or pressure.
This chango is limited to termination / sparing of existing conductors and delet ion of jumpers to improvo voltago condit f ans of these Division 111 circui u.
i NPE90/SNLICTLk - 7 Y
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Attnrhment to GNFO-91/00001 SRASN: N1'E-90-027 !)DC NO: CN-90-0109 SY S11.M : 1.21 Di~ SCRI pTION Ol' Cil ANGE' 1his change teplaces resistors in 125 VDC grourut detection citruita with lower value resistors to allow full senic roet er de f lect f on.
KEASON FOR CilANGE', The change to the lower resistance value sosistors nilow full scaln meter dnflection. The change to repinre the contact blotk of punhbut ton swit ch pit 2 nilows isoint ton of t he test cit ruit f or t he meter f rom grourel to ensure that nity exist itig groutuis on t he nyatem will not interiete with t est itig of t he tret et .
SAIT.TY EVAhl1ATION: Tbn sa f et y evnlunt ion r onc luileil that the change in til not involve an unteviewed safety questlon. The modification will have no physioni impact to any sa f ety e clat ed t ornpon e n t s , structures, or systems described in the I S.:R. The changes are conf ined t o t he interior of non safety relate.1 panels and do not affect the function of the systems. Existing fuse prot ect lon on the cont s ol circuit ensures thnt mnifunctlons will not piopngat e t o t he DC Dist ribut ion l'ntiel . The conseepient es of f ailure of the dist ribut ion panels, however, are enveloped by accident 8 or oc cut rences already evnlunted.
The inodificat ion will have no impact on systems, c ottiponen t s , or functions that could alter any technical specification safety ma rg itis . Ir.olation and separation of the cittuit per Reg. Guide 1.75 will prevent propngntton of failures t o any ot her e<piipment .
NPE90/53hlCI'!.R - 8
uun At t nchnie nt to GNKO-91/00001 SKASN: N i'): 02 8 1100 Nf): liCI'- 84 014 9- S00- k00 SY STI.vi : K(iO 1)l.SCK il'T 10N Ol' Cil ANGl'. . Sn f egun t ds DC1' Kl'.ASON }TK Cil ANGl:- Safegunid% lirl' S AIT.rY IWAl.tt AT l hN : Safeguntds DC I' The sn f et y 1:vnlunt 100 is nyntinble foi review at Ginnd Gulf Norlent Sintlon.
NI'l:90 / SNI,1 C1'i,K - 9 l
l Attachment to GNKO-91/00001 SRASN: N1E-90 029 DOC NO: DCP 88-0042-S00-k00 SYSTEH:
1)ESCRIPT10N OF CilANGE: This chango provided for the crect ion of o
an enclosed structurn in the Motor Control Center Area (HCC) at !
Elevation 133' of the Turbinn fluilding. j i
REASON M CllANGE: This change provided Nuclear Operat ion 'II' l
!j personne e permanent workbase to use for planning and scheduling net (vitlen.
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SAFETY EVAI.UATION: The safety evaluntton concluded that the changa did not involvo an unreviewed safet y quest ion. This facility is located in the HCC Aren at Elevation 133' of the Turbine Building and is not in closo proximity to any safety I reinted components. Addit ionally, the minimal amount of safnty related components in the Turbine innilding nrn designed to fall
- safn or in a manner that doca not compromiso any required safety .
) function. In accordanen with the original det.ign crit erin for j
, structurns located within the Turbine fluilding, the facility was designed to satisfy Uniform iluilding Code (UllC) requirements, l including seismic. The incility is const ruct ed and finished with non-combustible materials and contains a smoke detector to provido early warning detection. The facility itself, however, is not required ;o havo firo rated boundaries. The londs assocint ed wit h !
this facility, including livo lond, nro wnll within the livn lands
- 5.pecified for this port ion of the Turbinn Building and thoroforo do not adversely impact the Turbine fluilding 133' floor slah or structural ateol, Additionally, this facility does not house and
, is not located in close proximity to any equipment or component ;
used in mitigating the consequences of an accident.
1 ,
The new nt ructute does not degrade the ability of any Fire Protection System to perform its intended function, does not int roduco new or dif ferent failuro criterin, nnd does not. l ndversely a f fect or invr ~ idat e exist ing analyses for postuinted design basis fires. ;
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l NPE90/SNLICFLR - 10 e a,,m,. , - . - - - - - - - , _.,-,--.n,r,mm--,,:,.,...,..,--,-,.,m-, , ,.,--,,vw-eeww-mer,,.mm,-, - v~~-
l Attachment to GNRO 91/00001 :
l SRASN: NPE-90-030 DOC NO: CAhCUhAT10N NPE-E220004 SYSTLH:
f i
I 1)LSCRIPT10N OF CilANGE: 1his cniculation revised thn maximum stem i
thrust that can be applied to valvo E22P004 whilo mnintaining nli components within code allowable limits. The mnximum st em t hrust !
providen a mnximum upper bound limit for ihn Mechanical [
Specification for torque switch antting on motor operated valves. l The torqun switch is used to stop the motor f rom providing a st em i
- thrust higher than the valvo design will allow. I
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- REASON FOR CilANGE
- The original nelsmic stress calculations determined a required thrust value based on thn expected change in -
pressure in the valve. This value was only baned on empirical formulan. This supplemental calculat ion determined the maximum thrust based on the actual valvo design. Thn cniculation rennalyzes only thone components which are affected by stem l
, thrust. :
SAFETY EVAhUATION: The nnfety evaluation concluded thnt the change did not involve nn unreviewed safety question. Increnning ,
thn stem thrust value for valve 1022r004 in the noismic stress I nnnlysin does not physically change the valve or modify the use of I the valve. This cniculation only shows the mnximum stem thrust 1 which can be cbtained while maintaining all valvo components, both pressure retnining and non-pressure ratnining, within allownble codn limits. This calculation shows all stressen nrn within the !
- codo allownble limita and that pressure integrity and st r.icturni ;
integrity is maintained for operational loads, internni pressure [
londs and meismic loads. This thrust value will only he used an a mnximum t otal thrust- limit in Hechanical Specification SERl-MS-25.0 for the testing of motor operntcr valves and the sett ing of the torque switches. The torqun switch in used to stop the valve motor operator at a thrust lower thnn the maximum thrust det ermined in the r.upplement al cniculat.Jon in order t o maint ain .
l the integrity of the valve. The supplemental stiess eniculation in performed to show the valve can maintain ASME code allownbles i for pressurn ret aining components with n larger stem thrust than !
was previously cynlunted in the nriginal cniculation, ,
t Providing supplemental seismic stress cniculations for a valvo !
wi11 not affect the hasis for any CGNS Technten) Specificatfon.
The eniculation in performed to show thn valve will still per form t
- its intended function during normal operation or any accident condition and the valve component stresses are still within the ,
originnl design basis. Therefore, thn margin of nofety an defined in the GGNS Technical Specificat ton has not been chnnged.
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NPE90/SNI,1CFhR - 11 l
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i Attachment to GNRO-91/00001
]
't SRASN: NpE-90 031 DOC NO: EERR NO. 90-6162 SYSTEM:
14 i
DESCRIPTION OF CilANGE: Engineering Evaluatton Request Responso I (EERR) No. 90/6162 was issued fer the installatton of a temporary l snubbor t est ing f acilit y on Elevat ion 166'-0" in the Southeast :
i Quadrant of the Auxiliary liollding ( Area 7). The testing facility j consists of an 8' x 17' room, which housen a computer and printer. ;
cont rol console, two desks or tables with chairs and filo '
cnbinct; and an adjacent 12' x 32' test room which honnes thn i snubber t est bench, a work table, and storage cnbinnts.
i REASON FOR CilANGr: This EERR speelf f ed the requirements for the temporary snubber testing facility that. was installed to support j snubber test ing during RF04 "
SAFETY EVALUATION: The safety evaluntion concluded that tho changn did not involve an unroviewed an fety quest ion. There are l
. no design basis events (anticipated operational occurrences and I accident s) described in the UFSAR that are applicable to the !
installation of the temporary snubber test facility or its ;
supporting equipment. Appendix 9C of the UFSAR requires that. "a single exposure firn cannot affect rodundant sa fe shutdown-related f components". The temporary testing facility was installed in Firo Zonn 1 A403 of Fire Area 19. Per the Fire llazards Analysis (Fila) for GGNS Unit 1. this firo zone contains only Division 1 safo shutdown components. Sufficient physical separation in provided from adjacent Division 2 mafo shutdown components to ensurn that a postulated fire in Fire zonn 1 A403 does not af fect nor propagato to affect more than one safn shutdown t rain / division. The analysis of safe shutdown in the event of a fire, as described in
- Appendix 90 of t he UPSAR. is not adversely affncted. ,
. The temporary testing facil.ity does not adversely af fect tha !
existing operation of plant systems, structures, or components -
required for the mitigntion of a postuinted event. The potential !
radiological dose rates postulated for accident crind i t ions described in the UFSAR and as limit ed b) '0CFP20 and 10CFR100 are not increased.
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Attachment to GNKO-91/00001 NPE-90-031 I pngn 2 The temporary snubber test facility dons not int roduce intervening i
combustibles which would compromise thn separation of Division 1 and 11 safo shutdown components as described iti the fire Hazards
- Analysis (IHA). The t emporary olect rien t power to the facility is supplied f rom non-safet y related fl0P power receptacles. To ensure that equipment import ant to safety is not affected, the power feed is installnd to provide physical separntion f rom safety related l
equipment per thn requirements of Reg. Guide 1.75. 1he temporary test facility is a non-anismic sttucture and a seismic 11/1 i walkdown hns been performed to ensure that no sa f ety rnlat ed systems, st ruct.ures , of Components nin MIfected. The integrity of the Auxiliary Ilutiding structure for the addit ionnl londs crented a by the temporary test facility and its reinted equipment was verified. A partial blockngo of an exist ing emergency light is erented by the temporary consttuctlon. This partint blockagn han
! been cynlunted t o ensure t hat sufficient lighting will be mnintained along the a f fected ingrens and egre'sn rout ns in accordance with 10CTR50, Appendix R, Section lit.J.
1 The installation of t hese t empornty power supplies and t he selection of cable sizes performed fu accordance with Reg. Guide 1.75 and Articln 310.15 of the National Electric Code to ensure that possible accidents remain within tho bounds of existlug annlyses evnlunted in the UFSAR. proper sizing of cabin for the temporary power feeds t o the snubber test ing facility ensures that.
there are no adverse effects to existing plant equipment. The power feeds are elect rically isolated and physically separat ed from existing safety rninted components. Clamping devices and a support rostraint are installed to components of thn snubber test machinn, in conjunction with special requirement s for operation.
3 to prevent pot entini miullen which could compromiso exist ing plant safet y rnlated equipment . Therefore, thern is no crention of a possibility for nn accident or malfunction of a dif ferent typo than any nvalunted previously in thn Safety Analysis Report.
The snubber test ing dons not mcxtify, deletn, or add any new or unanalyzed londs to existing plant electrical or mechanical systems or components that could change the operational or funct ional characterist ics of the plant thnt could result in a change to the safety limits of conditions of operation na defined in the bases for the Technical Specifications. The Auxillary i linilding St ructure has been qun1ified for the added loads irom the test facility and its nquipment. Thernfore, t he const ruction and use of the temporary snubber testing incility does not reduco any of the margins of safety defined in the bases for any Technical Spectficatlon.
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NPE90/SNhlCFI,R - 13 l
Attachsent to GNRO 91/00001 l SRASN: NPE-90-032 DOC NO: CN-90-0125 SYSTEH: E12 DESCRIPTION OF CilANGE: This change adds two new mnsmal vent valves to the ADilR system and deletes vent valves E12-r427 and E12-T418. These valves are the new safety to non-safety boundary of the vent system and will perform the function of venting and I i isolation of the Alternative Decay llent Removal ( ADilR) system.
REASON FOR CilANGE: The previous two valves, E12-r418 and f E12-F427, were difficult to access, i
SAFETY EVALUATION: Tha safety cynluation concluded that the 4
change did not involve an unreviewed safety question. The snfety related piping and pipe supports designs meet ASME Section 111-requirements and are qualified as seismic category 1.
The non-safety reinted piping and pipe support meet ANSI 1131.1 requirements and are qualified as seismic cntegory 11/l. The addition of the piping and pipe supports does not affect the integrity of the interfacing piping systems or any safety system.
The piping and pipe supports will function in their intended manner. This design change will allow easier venting of the system. Tlic operation or function of the E12 system, as analyr.ed in the FSAR, is not. af fected by the modifications of this change.
The installation of the piping and pipe supports to the system '
will not chnnge the system function or operat ton as defined by any bases for the Technical Specifications.
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i NPE90/SNI.ICFLR - 14
. . _ _ _ , _ . _ . ~ - _ . _ _ . _ , _ , _ . _ . _ _ . . . _ _ . _ -__. _ _ _ __ _ ._ _ ..- _ _ _ ____ __ _
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Attachment. to GNRO-91/00001 j i
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SRASN: NPE-90-034 DOC NO: CALC. LC-Qll.21-85001,R02 SYS1EH : I J
- DESCRIPTION OF CllANGE
- Revision 2 to this calculatinn is revising j t he Division i nnd 11 bat tery load profiles. I REASON FOR CllANCE: Calculation EC-Q11,21-85001, Kev. I was issued !
to voriiy thn adequacy of the Divislon i ninl 11 125V DC batterleu ;
during a worst caso scenario (1,ons of Offsite Power and associated diesel generators in conjunct ion with a LOCA). Calculatton ;
EC-Q1h21-85001, Rev.1 ident ified a diosol gener ator field i' finahing lond of 70 nmps during the first nint third 1 minuto -
periods of the bat teries duty cycle. Per 10E0 485-1978, if a discrete sequenco of momentary londs can be established, the load 3 for the 1 minute period shn11 be assumed to be mnximum current at any lustant. Since the flehl finshing circuit for the diesel l genesator is opened prfor to the generator and fIrat sequencing load group brenker's spring chnrging motors energized, and the lond for the brenkers spring chniging motor envelops the ,
l gener.itor's f ic ht flashing lond, the generator's field flashing !
lond will not be list ed for t he first ami third cycle of the bnttely londing tables. Also, since t he duty cycle of the 4.16KV
- spring charging motors is 2 p,oconds nml n 5 secosul delay exist for i one of the two first sequencing lond group londs, two concurrent swit chgear operat ions will be cons blered (diesnl generato anni one first sequencing lond group bronker). Also tho lond identified in cniculntion EC-Q1L21-85001, Rev. I for t hn Unint errupt (ble Power l Supplies (UPS) will be increased to allow for nn added margin i betwenn t he existing UpS load and fut ure lond athlitions. UFSAW i Tables 8.3-6 and 8.3 7 will be revised to reflect the results of 1 this calculation for the exist ing loads on the 125 VDC EHf batteries A and 11 This calculation is based on the methmlology 1
! described in IEEE 485-1978 ' Recommended Pract .len for Sizing 1.ntne Lend Cell lint t eries For Genernt ing St at.f ons niul substations' .
l SAFETY EVAll'ATION: The battery lond as determined by this ;
j calculat ton is lower than or equal to the Technical Specification '
lond ut ilized for the opernt ional surveillance for EST Bat teries A nnd B. These bat t eries have demonst rated the capacity to mnint ain the minimum allowed t erminni volt age of 105V using the t est ing e lond thus, demonstrating the capacity for t.be load det ermined by this calculat ion. The calculat ion also shows that the battery l chargers are adequately sized to recharge the bat teries in less ,
t hnn 12 Ilts , as presently inquired per UFSAW H.3.2.2.1. The UFSAR I table changes per formed for t hin calculation r equire no modi ficat ion t o the 125 VDC ESP Division 1 or 11 bat t eries to accommodato the revised lond calculat Jon. The modificatlon to r UFSAk tables 8.3-6 ntnl 8.3-7 is only a sof tware change r equired to ;
updat e t he UTSAR to reflect the resulta of the revised lond ,
calculation.
NPE90/SNLICFLW - 15 e
-+n ----> ~ - - - - , -- - _ n.--m - - n-,- ~,--n-,,-,,-.,,,,-.m. ,,--vrnn,,-n.--r.-v.------,,,--- we--...,-, v-,w-mv,,-r-,--w
i Attachment to GNRO-91/00001 1 i l NPE-90-034 )
Page 2 5 1 1 The actuni worst caso lond an determined by this calculntion is [
within the capacity of the 111 vision I and 11 EST bntterien ,
according to the methodology specified in IEEE 485-1978, ;
'Rocornmended Practico For Sir.ing 1.ntge 1.ond Stot age tint t orien for
- I Generating Stations and Substation'. Also, the lond in less t han [
or equnt to the Technical Specification lond during all timo '
periods. The batacries have demonstrated thn ability to necommodato the Technient Specificntion lond and thun, hnvo demontitrated the ability to accommodato the worst caso lond an determined by this calculation.
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NPE90/SNIlCFl.R - 16 i
l Attachment to CNRO 91/00001 l
SRASN: NPE-90 035 D0d NO: CN 90-01M SYSTEM: E12 DESCRIPTION DF CilANGE: This. change notice will add an annunciator on cont rol room panel till3-P601 17A for the Alt ernat ive Decny llcat ;
Removal System (ADilRS). The annnnciator will alarm on ll1 ADilR ,
llent Exchanger inlet Temperature or 1,0 ADilk System flow. Should i this alarm occur, the ADilR heat exchanger inlet t emperature j 4
Indicator or the ADilR system flow indicator, both mounted on 5 control room panel 11113-P601-1711, will provide indication to nilow ,
! operations to determine which parnmeter caused the alarm. !
REASON TOR CilANGE: This dealgn change is an enhnnecment to the existing ADilR r.ystem, providing an nudible alarm in the contre.1 room and thus this change will not have any impnet on the existing .
design functions or operntion of the ADilk system. i SAFETY EVALVATION: The safety evalunt ion concluded that the !
l change did not involve nn unroviewed safety question. The !
- recommended alarm setpoints are consistent with existing technical specificat ion requirements for t he applienble reactor opnrnt ional
condition. The control toom annunciator system is a non-safety l related system. The design will ut flize existing t ransmitt ers to provide input to alarm cards which will provide input to the annunciator logic. The alntm cards will be installed in an 1
existing card rack. Proper separation will be maintained within ,
thn panels for the alarm corda, transmitters, and annunciator I logic. Since there are no ESP devices within panels 11113-P84 & l t
lill3-P63, this design will not create any seismic 11/1 concerns,
- One cable will be routed in non-divisional floor cable ducts
, within the control room, maintaining proper separatfon. Failure ,
of any component added or modified by this change will not :
initiate any transient or accident previously evalunted in the -
UFSAR. The char.ges made by this design will not prevent any equipment relied upon to mitigate the consequences of any cynlunted accident from performing its safety funct ion. No equipment important to safety is affected by this change. All
- ' necessary requirements and commitments are met by the new design ;
and no new accident precursors are created. l The addition of this annunciator does not change the originnt design intent of any equipment and all applicable design and installation requirements are met.
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i i i Attachtnent to GNRO 91/00001 ;
SRASN: NPE-40-036 DOC NO: W.O. 19998 SYSTEH:
4 DESCK!pTION OF CilANGE: Work Order 19998 provides d ' rect lonn l necessary for the npplicntion of Induction llenting St ress I improvement (11181) on 34 Reactor Prennuro Vessel Nozzle Woldments.
The work order providen direct ion for the location of major components of thn 11181 process and will establish temporary I
sources for elect rical power and cooling water necessary to thn process. 11151 wns implemented witl* the reactor in Mode 5.
REASON FOR CllANGE: Thn 11181 process is intended to be applied to welded joints of austenitic stainless steel which are the prirnary materials for which the stress improvement process was developed.
The 11151 treatment in being performed to mitigate thn 4
~
susceptibility of the inconel weld materinin to intergranular stress corrosion cracking (IGSCC).
J SAFETY EVAhUAT10Nt The an fet y evaluntion concluded that tho procons does not involve nu unreviewed safety que at 'on. Sinco 11151 changen only t he residual st ress st ato at the insido surface of thn piping weldment from tensiin to compressin and the exist ing design is unchanged, no moden of failure are int roduced.
With the climinat ion of a major st ress f actor, t ho incident of IGSCC is significantly reduced and therefore the probability of an accident in reduced.
The implementation of 11181 on the Reactor Vessel Nozzle Weldment s does not change the existing design, physically or operationally, therefore existing safety evaluations re.nain unchanged. With thn climinat Ion of a major strens factor, the incident of IGSCC is nignificantly roduced, therefore, reducing the probability of a failurn of the Nozzinn.
The application of lilSi will ensurn that.thn structurn) integrity of the Rnactor Pressurn Vessel is maintained by climinating n l majo" st ress factor as a contributor to IGSCC.
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l NPE90/SNhlCFl.R - 1A l
l Attnehment to GNRO-91/00001 1 )
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, SRASN: NpC-90-037 DOC NO: CN-90-0182 SYSTEM: G41 '
1 i DESCRIPTION Or CllANGE: This change disablen or removen the stop check vnives currently utilized on the return linen to the spent fuel pool and provides redundant. pannive anti-niphon vents.
- REASON FOR CllANGE
- The stop check vnives were a frequent maintaince item, and this change provides hiphon protection for the subject linen through passive anti-siphon vents.
i SAFETY EVAhUATION: The safety evaluation concluded that the change does not involve nn unreviewed safety question. Siphon
, protection in provided on the supply lines which terminate below the minimum pool level an required by TSAR nection 9.1.
The active siphon protectinn system previously provided in being removed f rom the system and replaced by a possive system. Ilcing passive, it does not rely on active components and thus incrennen the reliability of the system. No other equipment is affected by i this change. This change doen not affect the compliance of the overall design to 10CFR50 Appendix A criteria 61 and 62 an dincussed in FSAR paragraphs 3,1.2.6.2 and 3.1.2.6.3. Ai1 of the limits for stored f uel shielding, cooling, and reactivity control an described in FSAR paragraph 15A.6.2.3.14 arn unaffected by this change. The cask drop in the spent. fuel pool nccident described in TSAR nubsection 15.7.5 and the fuel handling nccidents i donctibed in FSAR subsections 15.7.4 and 15.7.6 nrn also unaffected by this change.
i The denign provided by thin han been evaluated against the applicable design criterin, installation, and. operational requirements, it was determined that all necessary requirements and cerimit ment s a re met by the new design and that no new accident.
procursors are created. The overall capabilities of the spent fuel pool as described in FSAR sections 7.1, 7.4, 7.6, and 9.1 nro not reduced by this changa. Therefore. there is no creation of a possibility for an accident or malfunction of a dif ferent type I
than any evaluated previously in the Saf ety Analysis Report.
The refetenced technical specifications and bases have been reviewed to determine if the margin of safety will be reduced by the implementation of the change. Technical Specifications require a minimum pool level to be maintained. The siphon protection method or function is not specifically addressed in the bases. The sipon protection is provided in the design to prevent inadvertent draining of t he pool below elevation 202'5-1/4 '. No reduction in the margin of safety results beenune of the the alternate method of siphon protectton provided by this chnnge.
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NpE90/SNhlCFhR - 19
At t n r httietit to GNKO-91/00001 SRASN: N 1'l:- 9 0 - 0 39 1100 NO: EI'.R - 9 0 - 6 2 ? 8 S Y S10.M :
1)ESCRIPTION OF CilANGE: This evnluntton nilows tempointy lenil shiel(ling t o be at t ache <l to cert ain Konctor l'ressur e Vei.sel nozzles. The lend shieliling will be inst nlleul ilur ing Opeint Ing Mo<les 4 an<1 % only, nint reust be temoveil pr ior t o t est ni t .
REASON I OR CilANGl;. This chnnge was maile in niiler to reduce rniliation exposure to personnel performing knik in this nren.
S Al'i.TY LV Al.11 AT I ON : 1hc < hange does not liivolve nn unteviewe<l sn fet y quest ion. Al l appl i c able AS'tr. Ct=le a l lownble 'tiesses nre mnt. Thn probability of occuitence of nn nrcident r esult inn f rom a seismicnlly init inte<l pipe bienk is not i nc i en s e<l . There will be no chnnge to exist ing designs af ter the l enti sbloiding is t etnov ed . These t empor nry changes do not affect the structurnt integrity of the nozzles or associnted piping during told shutdesn or f or Operat ing Modes 4 niul 5. liist a llnt ion of lend shielding temporarily does not change the limit ing rotulit ions for opei nt ion, applicability, or suiveillance r e<pi t i ement s ns de f ined in t he basis f or t he Technic.nl Spec i fic nt insis.
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NPL90/ SNI,1Crl,R - 20
, Attachment to GNk0 91/00001 ,
I SRASN: NPE-90-040 DOC NO: CALC: MC-Q1E30-90112 SYSTEM: l l !
x I DESCRIPTION OF CllANGE: NPE Calculation MC-Q1030-90112 was ;
performed to determined the effect an Upper Containment Pool (UCP) '
available water volumn reduction would havn on the containment i i analynim and to "an-built" UFSAR Tablo 6.2-50, "Supprennion Pool !
Ocometry - 251 Plant". The valuen resulting f rom the calculation
, were then applied to Table A-10. "Drywell and Suppression Pool l Geometry" of Appendix 6A to the UPSAR. Tablo A-10 contains the numerical values for parametern utilized for the GGNS Containment i Analyses. The engineering evaluation analyzes any resulting i differences to verify that the current "as-built" condit ionn are I bounded by the existing analysen. {
REASON TON CllANGE: Design Change Packago 86/0083 added an 18 inch extennion to the Uppor Containnent Pool (UCP) Dryer Separator .
Wall. The extension han two gaten supplied which are renoved ,
i duringplant operatfon. The extension required the addition of a 2 7/8 nill to the top of the exinting wall. This mill reduced i the UCP volume available for suppression pool make-up, t
SAFETY EVALUATION: This safety evaluation concluded that the l chango did not involvo an unreviewed safety quent ion.
l The calculation was performed to provide a basis for the parameters 7 utilized in tho various calculational modola for the Mark 111 '
containment. The dif ferencen identified between the paramet ers derived in the calculation and those utilized in the containment i analyses worn evaluated in the enginnering ovaluation. The i conclusion of the evaluation is that all identified differences ;
are bounded by the current analysen. No adverno effects on !
{ syntems, structures, or components previounty evnlunted will -
i result due to the conclusionn reached in the calculation. The l calculational results were evaluated as to tho impact of each l
) paramotor change on the various containment analyses. This ovaluation demonstrated that the parameters are still within the i design capabilities of the affected safety reinted structures and '
equipment or the "as-built pinnt configuration. Thorn is no adverso impact on syntoms, ntructures, and components necessary to '
mitigato a postulated accident af fecting the drywell or
! containment or to safely shutdown thn plant. ,
The referenced technical specifications and bases have been I reviewed to detervino if the margin of safety was reduced. A
[
limit for submergence of the top row of vents was identified as a design variable to verify dur.ing the calculation. Even with the reduend UCp make-up volume, the required 2 foot submergence is }
maintained. Based on thin fact and the results of the engineering evaluation, the plant design is bounded by the current accident analysis for t he "as-built" configurnt ion of the Suppression pool !
and Drywell. No reduction in the margin of safety results from the values determined by calculation no. MC-QlE30-90112, Rev. O.
l NPE90/SNLICFLR - 21 l
i
Attachment to GNRO-91/00001 l SRASN: NI'E 90 041 DOC NO: DCp 82 0056-800-R00 SYSTEM: P75 i
DESCRIPTION OF CllANGEt DCP 92/0056 changen the orientation of the ntnndby Dienol Generntors Start ing Air Storngo tank relief valven J Q1P75P025 A, B, C, and D f rom tho horizontal t a the ver tical l position. An albow and an additional picco of pipe were used to l reorient the valves to the vert ien t ponit ion. The valves were j previously attached in the horizontal ponition with a ningle picco j of pipo.
l REASON FOR CilANGE: The current vertical position of the relinf valves is lenn nunceptihic to inadvertent actuation and improven !
] valva rescating. ,
SAFETY EVAh0AT10N: The chango does not involvo an unroviewed '
anfety quest ion. The operation and function of the af fected i system will not be altered. The valve orientation and piping ;
nupplied by the DCp rneotn all applienble design requirements and ;
will function in their intended manner. The mounting of the f valven in the vertical position will enhance the reliability of i Division I le 11 Dicani Generators and will not impact tho !
capability of the Dienci Gennratoin to mitigate the consequences of an accident. The valves wore reoriented to the vert ical !
position un recornmended by the manufacturer and will not af fect 6 the structural integrity of the st arting nir storage tank. j Because thin DCp does not change the limiting condition for !
operation,' applicability of survalliance requirementa an defined I 1
in the basis for technical specifications, therc in no redut. tion in the margin of r.nfet y. ,
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. 1 NpE90/SNhlCpl.R - 22 I i
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At t at httent to GNRO 91/0000)
SRASN: N IT.- 9 0 - 0 4 2 DDC h0: D'.:P A2 417 8- S00- R00 S Y S i l'.M : I;31 Dl'.SCR il'T ION 01' Cll ANGl;' 1)CI' A2-4178 s oplac es t he lenk detectton turbine iteters used fot the uppe r c ont a i nitent pool 11 tie t , spent fuel pool liner, nnel r ef ueling bellows wit h s ight ginsnes. *l h e sight gins *.en will be per toelically chec keit by operat ions.
ki:ASON l'OR CilANGE: The tutbine met er s t enil to clog when lenknge f low occ urs. The use of sight ginsses is lot etuloil t o a llevint o this problern.
SAIT.TY LY A1.UATION : The change does not involve nn unrevieworl snfety question. There is no accident evnlunted in the I'S AR t ha t pos t u i n t e a, n failure of the exist ing lenk detertton instiuments.
1hete is no aut omnt to sa f et y f unct ton n5. soc int ed wit h t his equipment. All pipliig niul suppoi t s u s eil in this DCP vocet nll des ign s egui rernent s niul will f uin t ton in theit intetuted mnoner, l'a l l u r e of this flow monit oring syst em could result in nn init ially uselet ect ed lenk in the refueling bellows or pool liners.
Any lenknge in excess of the 25 gpm bientified lenkage limit will be det act ed by sump f ill annuncint oi s or by pump out timers.
No mit gin of saf et y as de f itu d for any t echnical 5.peci fic at ion is nfferted by this f low mon it oring subsyst em.
NPL90/SNLICILR - 23
Attachment to GNNO-91/00001 SRASN: NPl;90-043 DOC NO: DCp 84 0250-500-R00 SY STI.H t N31 Dl;SCRil' TION Or CllANGI;t DCP 84-0?$0 changen the operating rangen of the tuthino bearing pedestal and shaft vibration measuring innt rument at ion. Thn DCp will also modify all annociated comput er, 'antiuncint or , and a ccorder scales rangen and actpoints an approprinte.
4 RI;A90N FOR CllANGl;: The input circuit bontdh of the MenRor amplifier cards will be modified to narrow the operating tange as j per vendor recornmendt.tlonn. This will providn greater resolution and r endability of vibrat ion values in the lower rangen nasociated with normal operat ions.
l SAIT.TY FNA!.UATION: The change does not involve an untoviewed i safety question. There in no sofoty related function associnted 1
with the turbino generator control system. No accident provfously analyzed in the ISAR relies on tho turbinn generator bonring and shaft vibration monitoring system to mitignta the contiequenceu of 1 an accident . No malfunction of equipment important to safety previounty evaluated in the FSAR in predicated on a f allute of the turbine generator bearing and shaf t vibration monitoring ll equipment.
1 Thin DCP does not af fect the operat ion or function of the turbine generator bearing and shaft vibration monitoring system Compotief t t n .
- Itecauno thero in no af fect on the opnrat ion or function of the turbine generator bearing and shaf t vibrat ion monitoring nyntem components, there in no reduction in the margin of safoty as defined in the basis for any Technten) Specification.
1 NPl;90/SNhlCFbR - 24
At t achment to GNRO 91/00001 SRASNt N I'E- 9 0- 0 4 5 !)DC Nfh DCP-AS-4051-S00-R01 S Y STI.?i : p64 DESCRll' TION OP CllANGE- 11CP- 85 / 40 51 installs n pressuie switch in the pneumat ic actunt ion piping in order to provide auxiliary t rip f unct ions on manual act unt 10n. This DCP provides a means for shut t ing down t he llVAC inlet f ati inot ors atul f or clos ing t he int.ocinted fire dampets after a manuni init int lon of the N1P64D006-N lin ton suppression syst em. The innnun i act unt ion control panel (N1P64D006D) is bo ttig relocnt ed to out side the
,1nr.ntd nren.
REASON POR CllANGI.- To enh'ince fire protectloti pe r f o r rnanc e niter a mnnuni init int ion of the Comput er anu Cont r ol Panel Room lin t on 1901 fire suppression system.
S AIT.TY EV Al.11AT10N : The modificnttons performed by this DCP meet all applienble requirements of system specifications and fire prot ect inn st atninids. Mod i f icat lons nic consist ent wit h t he originni syst em design and vetulor r ecommendnt ions / t o< pili ement s.
The CMll wall from which the Automan 11-C panel and pressure sw it ch nie support ed hns been analyred t o assure it s st ructutal int egr it y. 1his DCP does not change t he sequence of event s or t he conkequerir es of a f ailure of the linton 1301 fire suppression sy s t erns for t he comput er and cont t ol panel s oom. Porther, this design change will not reduce the capabilit y of t his equiptnent to performing its intended function. The operation of Fnfety related equ i prne n t will not be affected by the litplernent nt ion of t his DCP.
No new interfaces with other equipment will be crented.
Normal nutomatic system actuation is not affected. System power, inst rument at ion niid cont rol are such t hat t olinbilit y is not reduced.
This DCP does not alt er the abilit y of thin equipment to meet fine prot ect ion st ntulat ds and sy s t em speci ficat ions. Putther, it does not affect the ability of t he eq u i prnen t to perform in accordance wit h t he original design and vendor retornmendat ions. No system or components will be expected to operate out side of design or Technical Speci ficnt ion limit s.
NPE90/SNh1 cpl.R - 25
Attachment to GNRO-91/00001 1 j l i
t SKASN: NPE-90-046 110C NO: DCP 86-0073-800-R00 SYSTEM: P71 i-DESCRIPTION OF CHANGE: 11CP 86/0073 replacon the lubricating of1 f pump aNMemblien on Plant Chillern N1P7111001 A-N, ll-N, and C-N.
REASON FOR CHANGE: Part s of the instalici lubricating oil putnpn ase no lonAer available f rom the manufactutor.
SAFETY EVA1.11AT10N: The chango does not involvo an unreviewed safety question. The Plant Chilled water system in a non-nafety
) related system whose failure will in no way compromino any safety l related system'. or components or prevent a nato reactor chutdown. .
Further, the Plant Chilled water nystem doen not. function to l mit igate the consequences of an accident. The new and old lube ;
i oil pump nanemblien are very nimilar in design and construction '
with the new lube oil pump annemblies being vendor supplied i equivalents. The centrifugal comprennors are the only potential ;
missile s.ource on the plant chillors ovaluated in thn FSAk, ;
9 Ilecauan of the similarity of the old and new designs, any analysen !
of the old lube of1 pump assemblics with respect to missilo I
har.ards would be valid for the new lubo oil pump assemblies while j some piping modif(cations are required to facilitate the i installation of the new lubo oil pumps, none are safety related or ;
acismic. No new failure moden are being int rodur, d. j i
The plant chilled water system in not addressed by the GGNS Unit 1 :
Technical Specificntion. :
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t NPE90/SNI.]CFl,R - 26 m- w ww- .,, c m ~, ve.m<m,-,,,-, - c w- w - w -w ,m.,,v-- e m., ,m -,m., rr,.-w w -m w m mm me --w m ~ w
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Attachment to GNRO-91/00001 i
SRASN: NPE-90 047 DOC NO: DC P- 8 7- 00 34 - 500- R00 SYSTEM: 1,21 I
4 DESCRIPTION OP CilANGE: DCp 87/0034 installs funes on the load
- side of the four safety related GE model AK bronkers.
1 i REASON POR CllANGE: The fusen will ;rovido the necessary circuit !
protect ion for t he 12% VDC llus feeders. The existing breakers l will serve an disconnects only. This was dono on renponse to SER
- 28-83, which portains to f ailure of breakers of this typn. ,
1 SAFETY EVALVATION: The change dock not involve an unreviewed E safety question. The addition of the fuses added by the DCP j provides additional short circuit protection without changing the [
a circuit function. The use of theno fuses ensures a high j
- reliability in the prevention of spurious trips of the 125 VDC l
. s.y s t e m .
l However, the GGNS Elect rical Dist ribution Syst em functional ;
Inspection (EDSrl) documented in GNRl-91/033 contained one notico i
! of violation (NOV) which addresned a concern the NRC had with !
DCp 87/0034, NOV-50-416/90-24-01. The violation cited the lack of ,
an adequato engineering evaluntion of the 125 VDC Distribution l l Syst em Fune/llrenker contdinntion. l l
In the response to Notice of Violation, GNRO 91/00054 Entergy '
Operat ions has taken the following steps to correct t he problems *
- 1. A design review of the breaker coordination associated with I the DCP was performed. We nrn in the process of determining
. an approach to renolve the design deficiency. ;
- 2. A memorandum was issued to Design Engineering personnel ,.
involved in the application nnd coordination of protective devices. The purpose of this memorandum was to mako !
approprint o pernonnel aware of this violat ion and the potential consequences of failure of fully coordinato alt breakers associated with a modifiention. l 1
Af ter review of thn modification, it was determined that this deficiency ,
(NOV) is not connidered safety significant in that the fuses provide full ,
protection of the feeder cables and thn DC Dfntribut.lon System is not '
designnd for operation with a fault.
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NPE90/SNLICFhR - 27 r i
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At tac hment to GNko 'il/00001 SRASN: NI'l;-90 04 R 1)DC NO: DC P- 8 7 - 004 8- S00- K 00 S Y S11.kl : p75 1)f SCRil'T10N Of CilANGl.: 1his 11Cp r eplaced t he rarbon St eel st ort ing air manifolds with stainless st eel manif olds for the Division 1 St amiby Diesel Generat or ,
ki:ASON l oR CllANGr.: This c hange will eliminnt e corrosion caused by moisture in the manifold.
SAIT.TY IWAbt'ATION : 'the change does not involve n USQ. Replacitig the carbon r, teel piping will not change the opetability, function, at surveillance requirements of the Ulvel Generator utart ing air subsystem. The piping supplied by thi DCP meets all npplienble design recibitements and will function in its intended manner.
This (hange in no say tirpacts the Diesel Generator capabilit y f or mit ignt tog the consequences of an accident.
There is no change in the limit ing condit ion for operation, applicability of suivoillante requirements. operation or function of t he Diesel Generat or and consequent ly, there is no reductton in the margin of safety as defined in the basis for any Technical Spociff(ntlon.
N1'fM0/ SNI.lCI'l.R - 28
l Attachment to GNRO-91/00001 )
i i 1 SRASN: NpE-90-049 DOC NO: DCP-88-0027-800-R00 SYSTEM Ril l DESCRIPTION OF CilANGE DCp 88-0027 adds direct indication that selected t ransf ormers are energized and available for a power feed. This indication will be added directly to tbc panel mimics
] for Ilussen 111tD, 1211E, 13AD, 14AE, ISAA, 16AB & 17AC to ensure the operator has indication within close proximity of the breaker f handswitches. The status lamps for safety related bussen 15AA, 16AB & 17AC are being installed in parallel with the existing breaker synchronization handswitches, fed by a potential transformer on the incoming feed. For non-safety related busses 1
111tD,12HE,13AD & 14AE, the status lamps will utilize spare potential transformers and will have no effect on safety related equipment.
REASON TOR CilANGE: During plant operation, the electrical busses mimicked on the P807, pB64 and P601 panels do not have readily availabic indication to determine whether they arc energized. The operator must locate the proper meter on the vertical section of the panel to determine if the bus is energized. Due to the spatial relationship between the control and its associated indication, switching to a dead bus for power feed could occur. A i dead bus transfer could cause undesirable plant ef fects a nd ;
possibly a plant scram, t SAFETY EVAhUATION: This change does not involve an unreviewed ,
safety question. i t
This design change installs neon status lamps to indicate when voltage is present at feeder breakers for Busnes 1111D,1211E, 13AD, !
14 AE,15 AA,16All & 17AC. The voltage present status lamps are being added to aid the operator during breaker alignment changes.
The lamps are solid st.at.e passive components, which use lamp '
holders that have been seismically toted and qualift, Divisional separation requirements for each lamp being installed is maintained.
For non-safety related busses 1111D , 12HE, 13AD & 14AE, these status lamps utilize spare potential transformers and will have no offeet on safety related equipment.
1 for t he stat us lamps monitoring safet y related busses 15AA,16AB &
17AC, their failure could cause loss of synchronization capability for the reeder Breaker associated with the faulted bus status circuit. Ilowever, this failure will not cause Bus De-energization nor prohibit bus sync. or transfer with other available power sources. The preferred system lineup uses an offsite power feed for Busses 15AA, 16AB and 17AC. If synchronization capability is lost and the preferred power source is maintained during an evnluated event, these busses continue to be fed from the preferred source. In this situation, loss of synchronization will
- not prevent any equipment form completing their intended safet.y
] functfon. ,
i NPE90/SNh!CFhR - 29 4
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At tachment t o GNRO-91/00001 N I'E- 9 0 - 0 4 9 Page 2 If t he preferred power source for It u n s .e s 15AA 6 16All degi ndes or is lost, t he I,ond Fliedding nin! Se<piencitig syst em ( l.S S )
nutomnt ically t rips thn incoming bronkers ntnl t ies the diesel generat or t o the approptinte bon.
For hun 17AC, if the pr e f er reil power f eed degrades or is lost, the uintet volt age prot ect ton syst em notomat icnIly t rips t lie itu.oming breakers ntui tles t he llPCS diesel generntor t o the bus, l.oss of syra.hroniznt ion enpabilit y is not considered as an initinting or sequence event in any (TSAR nccident nnnlysis.
Failure of the st ntus lamps will thus not affect any safety teinted functlon.
N Pf.90/ SNI.1 Cl I,R - 30
l, Attachment to GNRO-91/00001 l 1
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SRASN: NPE-90-050 DOC NO: CN-90-0318 SYSTEM: E12 l F
- P"ECRIPTION OF CilANGE: This change to the manual overrido logic ;
for ECCS injection valves 1E21T005, 1E12r042B, lE12r042A and ;
, IE12F0420 provides a timo delayed contact in the closing circuit i auch that when the injection overrido circuit is scaled in while i the valvo is stroking open two seconds must pass af ter the valvo limits open hofore it will cycle clonnd. !
REASON FOR CilANGE: The breaker of valve 1E12T042A had tripped while testing the manual overrido logic. The logic allowed the l operator to seal in the overrido logic while valve was in !
- midstroke. The breaker tripped when the actuator tried to reverse !
, itself while still coasting in the open direction. !
SAFETY EVAhUATION: This chango does not involve an unreviewed safety question. The change in the low pressure ECr8 automatic '
injection logic bypass circuit meets all applicable denign requirements. The chango will not causo any system or component to operate beyond its design limits nor will it af fect overall system performance in a manner which could lead to an accident.
4 The hpCS and t"C1 manual overtido control requirements to prevent i opening of the injection valves without the RPV high/ low pressurn .
Interlock permissivo for prote tion against int ersyst em h0CAs are ,
unaffected by this change. Tb shutdown cooling event is likewise t unaffected since thn design reqiiremonts for the valvo control ,
handswitches are maintained. l 3
The postulated loss of a division of ECCS considered in tho
- determination of the most limiting failure for the various !
applicabin UFSAR Chapter 15 accidents is completely unaffected by this design change. No accident procursors ovaluated in the UFSAR are af fect ed by this change. The effect of a component failurn or single error in the operation of the manual override as modified by this 1)CP and CN remains bound,d by the most limiting divisional failure. The accident mitigation functions associated wit.h tho l use of the manual ovnrrido are addressed in the EPs. Thus, since this design chango does not. alter any of the assumptions or degrado any of the required actions and barriers relied upon for !
mitigating an accident, the consequences of previously evaluated I
&ccidents are not affected. Therefore, thero is no creat ton of a po aibility for an accident or malfunction of a diffnrent type i j than any ivaluated previously in the Safety Analysis Report.
The changea in the automatic injection overridn circuits for LPCI-A, B. C and hPCS do not affect any existing bases for the Technical Specification requirements and do not introducn any new [
requirements. Althcugh tnis change increases the capabilitics for manually disabling automa, c low pressurn ECCS injection i functions, the existing colacident system init iation signal logic and override annunciation features and requirements are not reduced. The apnn and close valve stroke times associated with Npn90/SNLICfhR - 31 t
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' I Attachment to GNI(0 91/0000' j a ;
', Ni'E 50 j l' age 2 l t
t he automat ic arni manun t net tvo sa f et y relat oil f unct ione (e.g. , ;
, injection and isolation) nro unnifectert by the new design.
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I Ni'E90/SNI,1CFl.R - 32
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Attochment to GNRO-91/00001 SRASN: NPE-90-051 DOC NO: DC P 00 29 - S 00 - R 00 SYSTE!!: N23 DESCR1PT10N or GilANGl; Inst n11 r edutulant I.ow-1.ow 1. eve 1 Switch on exist ing llent er lirain Tank lustrument Stiongback.
RI'.ASON FOR Cll ANUE : To induce scram frequency by incronsing r e l i ab i l i t y o f t h e 1.ow- 1.ow 1. eve l T r i p C i r c u i t .
SAIT.TY EVAL.UATION : This chnngn does not in.olve an uni ev iewe<l safety question. The installation of the reduininnt low level switch as proposed by this DCP will incronse thn reliability of t ho llent er Drn in Pump t rip signa l . This will reduce the probability for a loss of reedwater flov: which is nddressed in the UFSAR sectfan 15.2.7.
The present ly inst nIled swit ch (IN23-1.SI.l.-N081) does not hnvc any offoot on t he consequences of an accident evalunted in the USAR.
Section 10.4. 7. 3 st a t es "The cotulensnt e niul f eedwat er system serve no safety function. System analysis has shown that tallure of this system will not compromise any sn fet y-reinted syst r is o:
revent sn fo shutdown." Also in this sect ion the UFSAR stat es p'The cerulensat e rond feedwnt er syst em is not. required to ef feet or support t he sa fe shutdown of t he reactor or per f orm in the opernt. ion of reactor sn fet y features." The new reduinlant switch will have the same function and design bases as the original swit ch nnd will serve o.ily to hoprove the rallnbility of the llenter Dra in Pump t r ip a igna1.
The switch added by this DCP and all associnted piping and supports are in the Turb!ne llullding and have no seismic qualifications, therefore no 11/1 hazards will be crented by mod i f i en t ions made to the Instrument Strongback. The power for the now switch will be supplied from the samt bus which supplies power to the presently inst alled swit ch. This bus contains no safnty related equipment.
Neither the llonter Drnin Tank hovel nor llenter Drnin Pump Cont rol is sddressed in the Technical Specificatlon. The DCP does not change the original function or design bases niul therefore ions not affect the margin of Safety dnfined in the Technical Specifirntlon.
NPE90/SN1.ICFl.R - 33
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Attachment to GNRO-91/00001 SRASN: NPE-90-052 DOC NO: DCP-88-0056-800-R00 SYSTEH: H31 DhSCRIPTION OF CilANGE: This channo repinced the 12 ton capncity drywnll valve handling crane with a 5 ton hoist.
REASON FOR CllANGE: MSIV/SRV valve maintenanco at the crane inner and out er t ramrnils was ortromely dif ficult dun to the sir.o of the 3 hoist. <
i SAFETY EVALUATION: This chango does not-involve an nnroviewed safety question. DCP-0056 changes the drywell valvo handling crano from a 12 ton hoist to a 5 ton holst. The drywell valve handling crann is the only equipment affected. The crano is only used for maintennnco activities in the shutdown or refueling modes. Thir crane is Seismic Category I for structural integrity.
Iloth the 12 ton hoist and the 5 ton hoist are non-safety related Seismic Category il/1, Thorn are no structural changes required to the crano to accommodate tho 5 ton hoist. Tho 5 con hoist will not adversely impact the crano snismic qualification na the new hoist capacit y is loss than half of tbn existing capacity. Tlio new holst. la ccmpatibin for its antJcipa:ad servico environment and thorn are no changes to the crano's function. Plant procedures are in placn to assure the crane and hoist are inspected and maintained as appropriate and are only operated by qualified personnel. The new hoist will be fed f rom the existing pownr supply.
The drywell valve bandling crano is not used in any technical specification to dof fno the margin of safety. Nor do the changes hnrein require that it be used as such a base.
l NPE90/SNLICFl.R - 34 l
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Attachment to GNRO-91/00001 i
SRASN: NPE-90-053 DOC NO: DCP-88-0057-S00-R00 SYSTEM: R61 DESCRIPTION OF CilANGE: DCP 88-0057 provides three additional !
public address stations and four sound powered telephone stations insido the drywell.
REASON FOR CilANGE: To provido additional personnel safety and to improvn the efficiency of operations conducted insido the drywnll. l SAFETY EVAI.UATION: This chango does not involvo an unroviewed safety question. This chango will not af'foct the operation of any safety related equipment nor modify the operation of existing safety related systems. Scismic supports are provided for raceway and equipment to ensuro no II/I scismic hazards are created. The added BOP raceway, equiprent and cable will be installed to meet.
thn Regulatory Guide 1.75 separatton requirement.s. 6 No accident parametnra or exist.ing safety functions are being modified.
L The added sound powered phones and PA will be added to the existing sound powered phonn and to the nxisting PA system respectively. Each will meet a)) applicablo design, soismic, and
, separation requirements.
The PA system is not. addressed in the Technical Specifications, nor does the added capability adversely affect any system addressed in Technical Specificat.lons.
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1 NPE90/SNLICFI.R - 35
Att at.hment t o GNRO-91/0000)
SRASN: NPE-90-054 DOC NO: DCP-89-0343-500-R00 SYSTEM: R93 DESCRIPTION OF CilANGE: DCP 89-0343 installs a lightning dissipation system. The purpose of the system is to dissipate the charge in the aren of protection before a lightning strike occurs.
The system consists of a variety of dissipat ion systems including Paragon Arrays, pyrnmid arrays, parapet arrays, rim arrays, spline ball ionizers, hemisphere nrrays and DDS lines. This patented lightning protection system was installed on the following plant structures:
PLA.NT STRUCTURE ARRAY TYPE Cooling Tower RIM ARRAY Turbine Building PARAGON ARRAY Enclosure Building PYRAMID ARRAY Auxiliary Building PARAPET ARRAY Radwnste Building PARAPET ARRAY Water Treatment Building PARAPET ARRAY Radial Well Switchgent llouse llEllISPilERE ARRAY Radial Wells 1, 3, 455 llEMISPilERE ARRAY Meteorological Towers SPLINE BALL 10NIZER liigh tinst linio Lights llEMISPilERE ARRAY liigh Mast Balificio Lights SPhlNE BALL 10NIZER REASON FOR CilANGE: To help prevent various transients to plant equipment due to lightning strikes. Induced current caused by light.ning can cLuse plant monitoring systems to give erroneous information resulting in spurious trips or plant scrams.
SAFETY EVALUATION: This change does not involve an unreviewed safety question. Tc prevent equipment damnge from occurring after a lightning strike, protective relay schers automat.ically dis.onnect electrical sources and londs to mitigate damnge and regnin electrical grid stability. Reducing the potentini for lightning strikes will reduce the risk of creating plant monitoring syntem instability and thus reduce the risk of damnging plant equipment. Failure of the system (lightning strike) will not have a detrimental effect since the dissipation structares will fall as air terminnis.
Lighting could cause physical damage to protect ion ct>mponents .
Reducing thn poten t itil for lightning strikes will reduce the risk of dnmnging or degrading t he performance of protection system components.
The dissipntion system will also reduce the risk of transients in plant monitoring systems during high storm activity. This in turn will provide the operators with more rolinble monitoring instruments.
NPE90/SNLICFI.R - 36
Attachment to GNRO-91/00001 NPE-90-054 Page ?
The dissipation structures have been designed to withstatul a wital lond of 110 Hl'il. The n<hlit innnl londs irnposed on exist ing plant structures is small and has no offeet on t heir struct ural int egr i t y. Failure of t he dissipat ion st ruct ures utuler t ornado loads has been reviewed.
Since tornndo generat ed missiles were considered as the 1Imiting natural phenomenn in the der.ign of all st ructures that nie sa fety r einted, failure or collapse of the lightn.ing dissipation structures will not affect the ability of the Seismic Category I structures, systems, or components to parform tholt intended f unct ion. This system is a passive syst em and is tied into the existlug pinnt grounding system, it will neither interact with nor have any effect on other pinut systems.
[
E F
E F
NPE90/SNhlCFl.R - 17 l
Attachment to GNRO-91/00001 SRASN: NPE-90-055 000 NO: DCP-89-0343-S01-R00 SYSTEM: P47 DESCRIPTION OF CllANGE: DCP 89-0345-1 installs surge protect ion on the liard wired instrument power and data lines that run between the radial walls and the switchgear house. Surgo protection was also added between the Meteorological Monitoring station and the plant. The base packago of this DCP added a plant wMa lightnjng dissipation array system which included the radial wells, radial well switchgear house, and the meteorological menitoring station.
REASON FOR CilANGE: Surge protection reduces the detrimental effect on plant equipme?t caused by niectrical transients induend on the plant grounding system. '
SAFETY EVALUATION: This change does not involve an unreviewed safety question. The radial well system has no safety related function. It provides makeup to the standby service water system cooling tower basins through the PSW system, but this makeup capability is not required to safely shutdown the reactor following a LOCA. Failure of the surge protectora at the radini walls will not have a detrimental offect on plant safety. The surgo protectors will also reduce the risk of damagn to meteorological monitoring equipment due to transients during high storm activity. This in turn will provided the operators with more reliable monitoring instruments. The meteorological monitoring system serves no safety related function. Failuro of this system (i.e. failure of the surgo protectors at the meteorological monitoring station) will have no adverse of fect on plant safety.
Addition of this system will not creato any new failure modes of plant systems due to nicctrical fallute, it is a passive system which will be tied to the existing grounding system only, it will not interact or affect the operation of other plant systems. It is designed to reduce the probability of lightning induced transients causing damage to and/or inadvertent act.uation of plant equipment.
Surge protection is not addressed in the GGNS-1 Tnchnical Specifications nor does the Installation af fect any safety related system addressed in the Technical Specifications, l
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NPE90/SNLICFLR - 38
Attachment to GNRO-91/00001 SRASN: NPE-90-056 DOC NO: NPE-FSAR-89-0041 SYSTEM:
DESCRIPTION OF CilANGE: This chnnge adds section 3.2.6 to the UF..AR which will allow the replacement and modification of ASME Section III, Clast 1, 2 and 3 components which are not code stamped but meet code requirements. This is allowed by Generic Let ter (GL) 89-09.
This change also provided foi the addition of TABLE 3.2-5, ASME Section III Component or Component Parts obtained to the Guidance of Generic Letter 89-09.
REASON FOR CllANGE: Numerous ASME accredited suppliern/
manufacturers have allowed their Certification of Authorization to expire. This has created difficulty for licensees to obtain replacements in full compliance with the licensing commitments.
The Nuclear Regulatory Commission has respanded to the issun w.ith the issuance of Generic 1.etter 89-09 (GL). The GL provides generic relief allowing the use of non-code stamped replacements for items that were originally code stamped. The GL contains numerous staff positions that require consideration and compliance by the licensee for its implementation. One of the staf f positions requires thn licensee's FSAR be revised to address the GL and identify the jccms obtained using its guidance.
SAFETY EVALUATION- This change does not change involved an unreviewed safety question. As indicated in Generic Letter 89-09, the replacement of Section III items' using the guidance of the GL provides an acceptable-level of quality and safety. The requirement imposed by ASME Section XI and III to use stamped items with accompanying documentation imposes undue hardships on the licensees without a compensating increase in the level of quality and safety over that provided by the alternatives contained in the gtidance of the GL. The technical requirements of thn code are still maintained and assured by the use of a 10CFR50, Appendix B Quality Assurance Prog sm in lieu of an NCA-4000 Quality Assurance Program. Use of the GL does not alter any evaluations that depend on the function of a Section III component.
Items affected by the FSAR change request are in accordance ASME Section III, Classen 1, 2 or 3 and are also classified as Quality Group A, B or C, respectively. Because of compliance with the rules associated with these classifications, margins of safety can be established and maintained. As recognized by the NRC, when replacements are not available in full compliance with the code stamping and documentation requircaents of ASME Section 111, the result is an undue hardship on the Licensee without a compensating increase in the level of quality and safety over that provided by the alternatives contained in the guidance of GL 89-09. The generic letter guidance provides an acceptable level of quality and safety to ensure continued function of the components that is
/
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At t achinont to GNRO-91/00001 NPE-90-056 Page 2 -
commensurate with those obtained having an ASME Section Ill code symbol stamp with accompanying documentation. The existing margins of safety are not reduced by the use of GI. 89-09.
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NPE90/SNLICFLR - 40
At t achment t o GNRO-91/00001 SRASN: NPE-90-057 DOC NO: DCP-90-0005-800-R00 SYSTEM: B21 DESCRIPTION OF CllANGE- DCP 90-005 replaces the carbon stent air accumulators and recnivers used to supply the Mnin Stenm Isolation Valves (MSIVs) and Main Steam Sniety/Rolinf Vnives (HSRs) with stainless steel accumulators.
This DCP niso char es valvo P53F012 from a stop check valve to a manuni globe valve.
REASON FOR CHANGE- The nir accumulators used to supply thn MSIVs and the MSRs worn fab. ' r el out of carbon steel with a protectivo conting on thn interf ac . The coatfng foiled and was replaced. The replat 'in necumulators with stainless
= teel vnrsion will ni v s problem.
Thn stop check valve wn; rer ed due to problems in performing the automatic depressuri>.stton system ( ADS) drop test. The test requiius the valve to be i the open positlon. The only way to accomplish this w<is to manue.'ly prop open the disc. The mnnual valve will allevinte this nrob'em.
SAFETY EVALUATION: This chango does not involve an unreviewed safety quest (on. The raplacement of the accumulators wIth staininss stent will not affect the operation nr the function of the HSIV's since the accumulator size and locatton has not changed. Thn replacement will in fact incronse the reliability of the MSIV's by reducing the possible int roduct ion o f part. icles to the valve accumulators. The replacement of the P53 F012 with a globe valvn will not a f fect the funct ion of the ADS nir supply since as a stop-check vnIvo its only aniety function was in the open pasitinn.
All piping and pipn supports changes meet ASME Section ill requirements and nrn qualified as seinmic category I. The safety related piping and pipe supports changes will function in thnir intended manner. The design provided by this DCP has been evaluated ngninst all applienble design critnrla as well as applicabin inst allat. ion and opernt (onn I requirements. It was dnt ermined t hat all necessn ry requ i rement s and commi t ment s are met by the now design and no new equipment failure modes arn introduced. The design change will not result in an operationni or functional change to the systems involved or to any other sa fet y related system.
The design change <ioes not affect the operation or functton of the MSIV's. The chnnging of t he P53 F0l? valve will not impact the operat ion of the ADS ns defined in the hoses for any Technical SpecifIcatlon.
NPE90/SNhlCFhR - 41
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l 1
-SRASN: NPE-90-058 DOC NO: DCP-90-0060-500-R00 SYSTEti: E12 DESCRIPTION OF CllANGE: DCP 90-0060 changes the manual overrido logic for valves IE12F042A, 11, C and IE21F005. These valvos arn the Low Pressure coolant injection (hPCI)-A, B, C and how Pressurn Coro Spray (LPCS) inject.lon valves.
This logic chango also af fects the manner in which the overrido can bo disabled and the automatic functions reenabled. With the existing design, thn automatic valvo control can be reset by nither: 1) placing the valve handswitch to "0 PEN"; 2) resetting '
the system initiation loalc following thn clearing of the initiation signals; or 3) the automatic resett.ing of the RPV high/ low pressure intntlock by a subsequent increnso in reactor pressure above the permissive for.valvo opening. The overrido scal-in installed by this DCP will of rectively prevent item (3) from disabling the override and will allow enabling of the overrido featuro prior to thn RPV pressure interlock being clenrod.
REASON FOR CllANGE: The mitigation of cortnin accident events involving operator act.lons governed by the Emntgency Procedures (Ers) requirns the disabling of the automatic low pressure ECCS injection functions. By the existing GGNS design, the automatic injection functions for the low pressure ECC systems (e.g.,
LPCI-A, B, C and LPCS) can be manually overridden by turning the applicable handswitch (E12-IIS-t1609A, B, C for LPCI injection valves E12-F042A, B, C and E21-ils-ti601 for LPCS injnct ion valve E21-F005) to the "CLOSE" position following the receipt of n '
system initint.fon signal. Ilowever, since the RPV pressuro
-interlock must first be clonrod before the override signal will seal in, the overrida attempt will not interrupt the initial automatic opening of the valvo. When the valvo 12mits open, the overt ide enn then be snaled in and the valvn will close. The override will remain sonled in and will inhibit all further automatic initint.f on sigunis until the logic is manually reset. or I
nutomatically reset from the unusual case of an increase in reactor pressure above the permissivo (which would then reenable the injection valyn intorlock, reclosn the valvn, and reset thn logic).
As a result of the EP requirements, this DCP will modify the ovnrrido logic for thn I,C P I - A , 11, C and LPCS nutomatic injection ,
circuits to permit the overrido function for the initial as well as all subsequent injections. This logic change is being accomplished by relocating thn override rniny contact from l downstream of the RPV pressure interlock relay to upstream of that
! relay such that the valvn automatic open signal can be inhibited following receipt and sent in of the ECCS initiation signnl (with
- bus power nynilabin) but prior to thn RPV pressure interlock clearing. The valve handswitches are also being replaced with new handswitches having additional contacts which will be used in associated with thn momentary contacts in the mnnuni overridn NPE90/SNhlCFLR - 42 L, . . - . . _ _ _ _ _ _ _ _ _--_ - - -- .- - --
Attachment to GNKO-91/00001 u ni,. 9 n. 0 g P1ge h logic circuit. These ndditionni contacts are used to seal in the override ci:roit when the handswitch is placed to "Ch0SE" t he valve and to allow the operator to break the override circuit ses.1- in and open t he injection valve when the applicablo handswitch is taken to "0 PEN" position. In conjunction with the handswitch contacts, new relay contacts are heir.g added in each vnivo closure circuit to sen1 in t he "Ch0SE" signal to the vnive mo t.o r . If the valve is stroking open at the time the override function is snaled in, these relay contacts will reclose the valve after it limits open.
SAFETY EVAhUATION: This safety evaluatton concluded that the change did not involve an uareviewed safety question. The change in the low pressura ECCS automatic injection logic bypass circuit, meets all applicable design requirements. The change will not cause any system or component to operate beyond its design limit s nor will it affect overall system performanco in a manner which L could lead to an accident. The hPCS and hPCI manual over ride control requirements to prevent opening of t.he injection va'.ves without the RPV high/ low pressure interlock termissive for protection against intersystem h0CAs are unaffected by this DCP.
The probability of a loss of shutdown cooling event is likewise unaffected since the design requirements for the valve control handswitches are maintained. The postulated loss of a division of ECCS considered in the determination of t he most limiting failure for the sequence of events of the various applicable UFSAR Chnpter 15 accidents is completely unaffected by this design change. No accident precursors evaluated in the UFSAR are affected by this change. Implementntion of this DCP "fil not affect the low pressure ECCS initiating circuits, logic. sequencing, bypassed, or interlocks, other t han permitting t he override function (and override seal in) prior to the clearing of the RPV high/ low prnssure interlocks and thus prior to the initial injection. The logic requirement for an ECCS initiation signal prior to an automatic injection override is not altered by this design change.
1: ndvert ent or improper operat.lon of the override function is thus 2 minimized by this coincident logic. Annunciation of the override
( funcLion is also not changed by this DCP.
].:
In evaluating the low pressure ECCS inject ion vnive logic circuit
'F design as modified by this DCP, the following single failures and operntor errors were postulated:
'b 1) any operator errors in the use of a single injection i
valve handswitch;
- 2) failure of the overrido senl-in contacts to make in the handswitch "Ch0SE" posit. ion;
- 3) failure of the override sent-in break contacts to make in the handswitch "0 PEN" position; NPE90/SNhICFhR - 43
Attachment to GNRO-91/00001 NPE-90-058 Page 3
- 4) failure of any of the new logic circuit relay contacts to make or break when required.
It should bn noted that since the new handswltches are seismically qualified, failures which would causo any of the new handswitch contacts to inadvertent.ly makn or otherwise fail as a result of a seismic event nned not be postulated. Inadvertent operation of an ;
injection valvo handswitch (e.g., to "CLOSE") would only enable '
the override function if that. action occurred with a system initiation signal present. The worst caso failuro rnsulting from this action would be a disabling of the associated automatic ECCS injection function. Each of the other malfunctions could result in a failure of the associated ECCS injection function or a failure to override the automatic injection depending on the exact failurn modo. Whether by a single component malfunction or by a single operator error, thn postulated failure of a single automatic ECCS injection function is still bounded by the limiting divisional failuro. In the event of a failure of the injection valvo override, injection can still bn stopped by closing other valves in thn process stream or by shutting down the applicablo ,
ECCS pump. The of fect of a component failure or single error in the operatin of the manual overrido as modified by this DCP remains bounded by the most limiting divisional failure. The accident mit.lgation functions associated with the use of the manual overrido arn addressed in the Emergency Procedur%.
This DCP only changes thn logic and associated handswitches for overriding the automatic injection function for hPCI-A, B, C and LPCS as described above. The replacement switches are seismically qualified and meet all of thn design and installation requirements of the original switches. The design provided by this DCP has been evaluated against the applicable design critoria, installation, and operational requirements. it was detnrmined that all necessary requirements and commitments are met by the now t- design and that no new equipment failure modes are introduced.
The potential for disabling the automatic injnction fut.ction for more than one low pressure ECC system by the usn of the additional overrido capabilities provided by this DCP when not operating by the Emergency Procedures in beyond single failure and singin operator orror criteria.
The new logic circuits and switchen meet all applicable design and installation requirements. The existing system and component design functions are not a f fected. Although this DCP increases thn capabilities for manually disabling automatic low pressurn ECCS injection functions, the existing coincident system initiation signal logic and overrido annunciation features and requirements are not reduced. All margins of safety as dof ined for any Technical Specification thus remain unchanged.
l NPE90/SNhlCFhR - 44
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Attacheent to GNRO-91/00001 1
SRASN: NPE-90-059 DOC NO: QDR-323-89 SYSTEM:
DESCRlpTION OF CilANGE: The auxiliary building isolation damper limit switch trips the noren1 auxiliary and funt building area t ilVAC systems upon initiation of the Standby Gas Treatment. System (SGTS). The purposo of the subject. trips, however, is not to mitigate the consequences of an accident. Instead, the primary purpose is that of providing basic equipment protection for the normal auxiliary and fuel building ilVAC system fans.
REASON FOR CliANGE: This change clarified the FSAR.
SAFETY EVAhUATION: The review that was conducted af ter the issuanco of this QDR confirmed the integrity of provinus design bases and found that existing prohnbilities of occurrenen remain valid. The subject UFSAR revision in in.ilcativn of the fact that while clar.ification was needed, no changes to existing analysis were necessary. '
This UFSAR revision does not change system operation, nor does it imply a reduction in the safety-rnlated capability or classification of existing systems / system components. The reason for this revision is to prov,ldo clarification of existing operation characteristics and not to describo any change from what has alrondy boon used in nvaluating the consequences of an accident. The safet.y functten of the SGTS and the capability of the secondary containment isolation valvos to perform thnir safety function have not changed, This evaluation reconfirmed that the exist.ing system design does not require any modifications. It also confirms the integrity of the evaluation of various accident.s addressed in the UFSAR. ,
The text of the subject UFSAR change emphasizes that. the existing design, utilizing the non-safety limit switches, does not constituto a deviation from required design considerations. The swltches provide the intended function of equipmnnt protection for non-sa fety related fans. As confirmed by the previously described review, the margin of safety provided in the original design has not changed.
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l NPE90/SNhlCTLR - 45 l_ . _ _ . _ . _ , - _ _ . _ _
Attachment to GNRO-91/00001 SRASN: NPE-90-061 DOC NO: MNCR-90-0032 SYSTEM: Z51 DESCRIPTION OF CilANGE: This change revises UFSAR Tabic 18.1-2 to remove the four second closure requirement for the post-accident fresh air makeup valves (Z51F007, F016).
REASON FOR CHANGE: The post-accident makeup air is separated from normal fresh air makeup. It is located such that intake air is filtered prior to distributing in the Control Room. The boundary valvo is normally closed motor operated valvo. It is opened only post-accident to admit fresh air to replenish the oxygen for the Control Room' operators and has no 4 second closure requirement.
SAFETY EVALUATION: This safety evaluation concluded that the change did not involve an unroviewed safety question. The Control Room emergency filtration system functions to mitigate the consequences of an accident, not to prevent an accident. These valves are normally closed and are interlocked closed for 10 minutes post-accident to admit fresh air and replenish oxygen for the Control Room operators. The IST program will be revised to require stroking of the valves during cold shutdowns or per ASME Sections XI rather than quarterly. Surveillances on the 251F007 and Z51F016 standby fresh air valves will be performed only in Operation Conditions 4 or 5 when core alterations are suspended, i.e., handling of irradiated fuel in the primary or secondary containment and operations with a potential for draining t'ao reactor vessel are not in progress. Under these circumstances, the possibility of design basis accidents and abnormal operations transients that can affect the Control Room environment are not deemed credible, and the risks associated with an inoperable filtration system are negligible.
f NPE90/SNhlCFhR - 46
Attachment to GNRO-91/00001 SRASN: NPE-90-062 DOC NO: DCP-90-0344-S00-R00 SYSTEM: E30 DESCRIPTION OF CilANGE: DCP 90-0344R00 adds Reg. Guido 1.97, type C, Category 2, wide range containment water level monitoring instrumentation to support the Emnrgency Procedures (EPs). Two separate channels are provided (Div 1 and Div II) with each consisting of two probes. The probes arn Fluid Controls Inc.
Model CL 86 Invol t ransmit ters. Two ranges are required to be monitored. The first or lower range will be from 20 to 35 ft.
lovel (113 to 128 ft. elevation). This providos an overlap with the upper end of existing instrumentation to a level above the upper limit of the Safety Relief Valvo Tailpipo Level Limit (SRVTLL) as addressed in the EPs. The secowl or upper range spans from a point below the elevation of the top of active funi to a point above the clovation of the concainment p. essure instrumentation tap. This ranga is from 60 to 75 ft. Icvol (153 to 168 ft. o l eva t. ion ) . The tw . cation for the probes will be in the control room.
REASON FOR CilANGE: Emergency Ope.ating Procedures EP-2, EP-2A and EP-3 require the operator to take action at containment water levnis beyond the rango of existing instrumentation. The presently installed suppression pool level indication avn11able in the control room only provides level indication over the rangn of 10.5 ft. (103.5 ft, olov.) to 25.5 ft. (118.5 ft. elev). The Emergency Procedures Figure 1, Maximum CTMT Water Level Limit (MCWLL) and Figure 3, SRV Ta11pipo Level Limit (SRVTLL) require the operator to make docisions based on containment levels outside
.this range. Without this instrumentation, a potential exists for prematurely terminating injection systems from sources external to containment. Because'of this SERI has committed to the NRC to install containment wide range water levnl monitoring.
SAFETY EVALUATION: This safety evaluntion concluded that the change did not involvo an unreviewed safety question. This DCP provides the Control Room Operators with Indication of Containment water level during the performance of the sito specific Emergency Operating Procedures EP-2, EP-2A and EP-3 thus enabling an oporator to identify when an EP level satpoint or decision point is reached. Failurn of the instrumentation installed per this DCP will not compromise any existing safety related system or component nor will it prevent safe reactor shutdown. No new interface is created which would affect components, equipment or systems which perform sefety functions.
The changes modo by this DCP do not prevent any equipment relied upon to mitigate the consequences of a malfunction of equipment important to safety from performing its safety function. These changes do not af fect any Seismic Category I system, structure or component. The circuits and raceways installed per this DCP are associated Div. I and II and are routed as divisional cables in l
NPE90/SNLICFLR - 47 l
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Attnthment to GNRO-91/00001 l
NPE-90-062 Page 2 seismically supported divisional raceways. The instrumentation is to be installed Seismic 11/1 in the containment for the probes and in thn Control Building for the ind ient ion components. The probes are considered to be structurally adequate to preclude electrical failure and all Seismic 11/1 concerns based on calculations performed and are cons idered funct ionally ndequat e to withstand n I There ute no ASMC Section 111, 1,0CA event and still be operable .
- Class 1, 2 or 3 piping or cor..ponents added or modified by this change. Divisionni separation, per Reg. Grido 1.75, of electrient components added or modified by this DCP is not compiomised by thn implementation of this change.
NPE90/SNLICFI.R - 48
l Attachment to GNRO-91/00001 l
SRASN: NPE-90-063 DOC NO: DCP-90-0547 SYSTEM: 112 1 DESCRIPTION OF CHANGE: DCP 90-0547 modifies t he i ns t.rument loops used for thn Safety Paramnter Dispiny Systen (SPDS).
REASON FOR CilANGE: The input inst ruments af f ect ed by this DCP arn being changed to be consistent with Reg Guido 1.97. The only exception will bn tho input for suppression pool level. SPDS is presently supplied a suppression pool level input from E30 1.T ,
N003A[B] whiln the Reg Guido 1.97 inst ruments are E30 I.T N003C[DJ.
Although the SPDS and Reg Guido instruments arn not tho samo they ,
monitor the samn level on the suppression pool, they have the samn rango and aro QF1 inatruments. Therefore nothing would bo gained by having the Reg Guidn 1.97 instruments provide thn SPDS inputs.
The following are the SPDS points to be added by this DCP: ,
RPV Level B21N027A B21N0278 Drywell Temperature M71N013A M71N013B M71N013C M71N013D Containment Tempnraturo M71N007A M71N007B M71N007C M71N007D Suppression Pool Tempornturo M71N012A M71N012B M71N022A M71N022B M71N023A M71N023B M71N024A M71N024B M71N025A M71N025B M71N026A M71N026B SAFETY EVALUATION: This safety evaluation concluded that the change did not involvo an unroviewed snfoty question. The instrument loops which nrn to prov.ido the new inputs to thn SpDS nro indication f r strument loops only and do not provido any control or trip function. Because thesn Instrument. loops do not l provido any control or trip function they could not bn the direct cause for the occurronen of an accident..
1 j NPE90/SNLICFLR - 49
Attachment to GNRO-91/00001 NPE-90-063 Page 2 Addit fonally, the instruments are not associated with any safety reinted equipment and do not provide any mitigating action. They will reduce the possibility of the control room operator receiving conf 1ict ing information during an event.
The new input.s to the SPDS are being obtained by using spare points on existing R Mux units therefore no new seismic considerations are being created. In addition the design of the new inputs maintains t.he proper 1E to non IE isolation.
Because the a f fected instrument loops do not affect any trip or control function there will be no reduction in the margin of safety as defined in t he hasis for any Technical Spect ficat f on.
NPE90/SNLICFLR - 50
Attachment to GNRO-91/00001 SRASN: NPE-90-064 !)DC NO: FSAR-CR-90-0032 SYSTEM:
1)ESCRIPT10N OF CilANGE: This changn provides for the deletlon from FSAR Table 3.10-1 ceitnin Category 11/1 and QP components that are seismically qualified mechanical,16G and elect rical devices.
REASON FOR CilANGE- The equipment is removed from Table 3.10-1 only because special seismic operability considerations beyond that normnlly performed to ensure structurnl and preserve integrity ato not required.
SAFETY EVALUATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evalunted in the Safety Analysis Report. The design requirements for the subject equipment are not mcxti fied. The equipment is not modified. The equipment is removed f rom Table 3.10-1 only because special seismic operability considerations beyond that normally performed to ensure structural and pressure integrity are not required.
UFSAR Section 3.10.2.3.1 allows the delet. ion of Category 11/1 and Qp components from the table. JS-08 designates all of these componentr. as passive. This change does not degrade the ability of the subject components to perform their required functions since design standards are not relaxed.
NPE90/SNLICfLR - 51
Attnchment to GNRO-91/00001 SRASN: NPE-90-066 DOC NO: CN-90-0101 SY STf,M : P44 DESCRll' TION OF CilANGE: This change relocntes thn existing Temperature Control Valve (TV) valves Nil'44F531B/ 532fl w it h t he exist itig but terfly valves N1P44F466B/467B to the condenaer inlet piping for Drywell Chillers N1P72B001B-B/B002il-B.
KEASON FOR CllANCE: The valves are switched so that the butterfly vnive will now isointo the TV valve to provide for propnr clenning and maintennnce.
sal'ETY EVAL.UATION: This safety evaluntion concluded that the changn did not involve nn unreviewed safety questfon. This modification provides for the removal of valve N1P44F925 to eliminate the flow obstruction problem nnd repincing it with a finnged branch line for hydrolyzing. The relocating of the exist ing TV valves N1P44P531B/532B with the exist ing butter fly valves N1P44r466B/467B will nilow for ons ter maintennnce of the TV valves. Thn existing flow point FP-N413A was relocated 15" to allow thn annubar to be installed. 8" JBD 378 pipe south of fourway valve was replaced becausn of the high velocity nrosion resulting from the bypass flow from the fourway valve. The finnged branch 1ine on the 8" JBD-43 was mo"ed nort h 23" to accommodat e pipe inst allat lon. Thn piping and pipe supports design meets ANSI B31.1 Code requirements. The piping is supported to dond weight londs only, since it is installed in a portion of the Auxiliary Building wnern no 11/1 hazards exists.
The Plant Service Water System serves no sa fety function. Systems analysis has shown that failure of thn Plant Servico Water System will not compromise any sninty reinted systems or prevent reactor shutdown. The operat ion or function of t he Plant Service Water System, as analyzed in the FSAR, is not affected by these mod .i f I cn t lons .
This modification to the Plnut Servien Water System does not chnngn thn function of operallon as defined in any Bases for the Technical Specifications, therefore, the margin of snfety is not reduced.
NPE90/SNIlCFl.R - 52
Attachment to GNRO-91/00001 SKASN: NPE-90-067 DOC NO: SERJ-JS-08 SYSTEM:
DESCRIPTION OF CllANGE- This revision to JS-08 incorporates the information developed in Phnse 11 atul Phase lit of the instrument Q-bist.
REASON FOR CllANGE* This revision provides design funct ions ,
safety applicat ion, Q-crit erin, and references any QEVAI, associated with an instrument. Alro t he as-huilt luformation identifled by Configuration Management as of 9/89 and the identified discrepancies between JS-08 Rev. I and Rev. 97 of 14echt el Inst rument index are inco rpora t ed in this revision of JS-08 using the new formnt. As a result of chnnging the Q-cInssificntton of some instruments a rrviston to FSAR Tnble 3.10-1 is required.
SAFETY EVAL.UATION: This safety evnluation concluded that the change did not involve an unreviewed safety question. The changes mnde by this revision of JS-08 do not constitute a change to any intended design function for any lustrument, system, or structure.
The change in clnssification is consistent with the As-llu t i t design documents referenced in the FSAR and with the description of syst em operat ions , when given, fu the FSAR. The inst rument clnssificationr. heing changed are connist ent with their f unction in accident mitigntion. The devices deleted from UFSAR Table 3.10-1 are eit her non-essential to nuclear saf ety or do not require specialized dynnmic qualif.! cation to maintain the currently postulated level of functionality or are previously exempted from specin11zod dynamic qua1iftentton in UFSAR Section 3.10.1.4.1,5. No physical modificnLion to any equipment. is being made. No changes to operational process, equipment function, safety level, or system parnmetets or characteristics are mnda with this chnnge. The safety function of the instruments are as documented in the As-llutit pinnt design. This is the function performed in the appropriate safety analysis.
The categor izat ion of these instruments is consistent with the current bases of Technical Speellications and plant accident nnnlysis. The chnnge to Table 10.1 is consistent wit h the current A s- Ilo i l t design documents.
NPE90/SNI.ICFl.R - 5's
Attachment to GNRO-91/00001 ;
SRASN: NPE-90-068 DOC NO: MNCR-0124-90-R02 SYSTEM: E22 DESCRIPTION OF CilA' IGE: MNCR 0124-90 was writ. ten to document Nuclear Plant Engineering's analysis of a torque switch adjustment. :
required for liigh Prcasare Coro Spray (llpCS) valve QlE22-F004.
REASON FOR CilANGE: This adjustment was necessary as a result of recalculation of the maximum allowable stem thrust (MAST) valuo based on the ASME Codo material stross allowables as part of tha :
continuing motor operated valve program in response to NRC Generic '
Letter 89-10. The existing torque switch trip point in the closing direction had been previously set such that the new (recalculated) MAST would bn exceeded. An Operability Review was conducted to verify that the valve had not been subjected to damaging conditions and to state that the valve was fully capabln of performing all required design functions.
The torque switch setting was required to be adjusted to provide -
valve actuator output thrust between the minimum required stem thrust (MRST) and the MAST. The torque switch adjustment was implemented as required to lower the thrust delivered to the velve, llowever, subsequent to that adjustment, the required local loak rate test (LLRT) was performed. This test failed to meet the leakage criteria for that valve (most likely due to compression setting of the valve gate and seat surfaces from previous operation at the higher thrust values). The torque switch setting for this valvo was then reset to the higher pre-MNCR value to provide the closing force necessary to pass thn LLRT. The LLRT was then successfully passed.
NPE reevaluated this situation for acceptability. This ovaluation has determined that allowable valve stresses for pressurn retaining valvo components have not been exceeded and the valve is fully capable of performing all required design functions. In addition, this evaluation concluded that the stress condition for the non-pressurn retaining valve part which would only occur during accident condit Jons is minimal and will not. impact the ability of the valve to perform its intended design funct.fon.
Thnroforn, the closing torque switch setting may remain at the higher as-lef t valun unt il startup f rom the fifth refueling outage.
SAFETY EVALUATION: This safet.y evaluation concluded that thn change did not involve an unroviewed safety question. The FSAR considers various events (accidents) which arn postulated to occur in order to determino the plant's capability to control or accommodate potentially damaging process disturbances and component failures. The accidents whose probability may be increased involve only those evnnts which are related to the ability of the subject valve to provide its passivn function of reactor coolant pressurn boundary integrity or its activn functions of emergnncy corn cool'ng, reactor isolation, and containment isolation.
NPE90/SNLICFLR - 54
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Attachment to GNRO-91/00001 NPE-90-068 page 2 Concerning the activo functions, the ability of the valvo to open or closo is not a precursor to any Design Ilasis Accident or transient. These events include the analyzed loss of coolant accidents, unexpected process or system perturbations, and reactivit.y events (FSAR 15A.6). The only event discussed in the FSAR directly caused by a malfunction associated with IIPCS is an inadvertent.IIPCS injection. This event is assumed to be the ,
result of an unintended manual pump start v.in an operator error.
Therefore, the stress condition of the valve stem does not change the probability of occurrence for this event (FSAR 15A.6.3.3).
Subparagraph NB-2121(b) of the Code stipulates that the Code requirements do not apply to items not associated with the pressure retaining function of a component such as valve stems.
Ilowever, subparagraph NB-3546.3(a) establishes that valvo stems, 1 stem retaining structures, and other significantly stressed valve parts whose failure can lead to gross violation of the pressure boundary shall be designed so that their primary stresses do not exceed Code allowablo values. Thus, the code allowable limits must be applied to that portion of the valve stem penetrating the valve body but need not be applied to the portion outsido the body.
It has been determined by calculation that the worst cr.sn stresses in the portion of the stem penetrating the valve bodt are within the Code allowable limits. Only the threaded port'.on of the stem located outside the body in and just below the ar.cuat.or were calculated to be stressed above the Code allownple values during worst. caso conditions, llowever, these strose. levels were determined to be well below the actual material yield strengths such that damage is not- predicted. The passive function of the '
valvn in thus maintained during all cor.dit.Jons and the applicable design margins required by the Codo for limiting the probability of a passivo pressure boundary fallure are assured. Thnro arn no credible failures of the active functions of this valve which can affect the probability of an ar.cident (e.g., LOCA).
The body-bonnet bolts worn determined to be overstressed when using the nominal yield r. tresses provided in t.he ASME Codes by Calculation NPE-QlE22F004, Rev. 1. A roview of tho valvo code data package showed that the originctly supplied bolts had a minimum yield stress of greater than 120 ksi, which is higher than
, the nominal value of 105 ksi. A visual inspection of the valve verified the original bolts were still installed in the valvn.
Using this information, Revision 2 to this calculation demonstrated that the maximum expected stress in the body-bonnet.
bolts will bn less than the allowable stress provided in the ASME Codes when considering normal operating conditions or abnormal
- accident conditions. The use of the higher yield stress of thn or f ginal body-bonnet bolts adds no additional accident. precursors l
which could af fnct the probability of an accident.
l NPE90/SNhlCFhR - 55
Attachment to GNRO-91/00001 NPE-90-068 i Pnge 3 The QlE22-F004 valve rnust act ively funct ion to mit ignte events which require inject ion of IIPCS to prnvent fuel or containment damage. These events include loss of coolant accidents as well as loss of normal feedwater supply or reactor cooling systems requir ing snakeup w i t h IIPCS ( FS AR 15 A.6. 3), in such events, the llPCS system must provide design flow at up to design pressure to ensure t hat FSAR nccident nun tyses nasumpt ions are inet (FSAR 6.3).
This action requires that the lujection valvo's active automatic opening function remain availablo and that, if closed, the valve will reopen when called upon to do so. The valve must. also be capable of being closed mnnually from the control room to isolatn the llPCS injection line, if necessary, to minimize radionctive release paths (FSAR Table 6.2-44). QlE22-F004 also has nu nctive funct ion of automatic closure on a high reactor water level following injection; however, this function is not assumed in any event annlysis. Addit ionnily, QlE22-F004 must maintain it s passive integrity so that the HPCS flow path is intact and surrounding equipment necessary t o mit igat o accident consequences is not damnged.
The increased stress which occurred with the higher torque switch setting was analyzed to have had no detrimental effect on thn ability of the valve to perform the above functions. This will continue to be true with the torque switch sett ing remaining at the higher value. A visual inspection of the valve has confirmed t h a t. no damage is present.
To date, no stress limits hnve act ually been exceeded.
Body-Bonnet bolt stresses have been determined to be below their allownble stress when actual bolt material yield stresses were evalunted. The stem stresses, while predicted to slightly excned
( 2'. ) the allowable codn stress under the most severe accident environment, will remain well below their yield point. Further, use of the ASME code allowable stress is considered only a guideline in this situntion. The ASME code (NB-3546.3) states that for components whose failure will result in gross vlointion of the pressure retnining houndary, the allowable primniy membrane stress should be used as the limit. This is not the ense in this situntion and standard engineering pract ice allows a higher allowable stress to be npplied (AISC, 1.5.1.3.1). Thus, no entnstrophic failure, hockling or other deformation is predicted to occur as a result of a the potentini overstress. The calculated force requirnd to open the valve remains within the capability of the valve act untor , and the maximum st resses expoeted to occur during opening remain well below the materint allownble. The abilit y of QlE22-F004 to perform its act ive funct.lon will not be affected.
1 NPE90/ SNI,1CFl.R - 56
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Attachment to GNRO-91/00001 NPE-90-068 ,
pago 4 Loss of valve ategrity from the standpoint of the consequences of ,
a resulting ilPCS line break LOCA was ovaluated. The conditions identified are not subject to endangering pressurn boundary integrity, thoa nfu a tb* analysis is not af fected.
Thn FSAR asst..nes the availab2iity of the llPCS system (and associated injec.; ion i ".vo) to mitigato the consequences of failures of equ2f.i'nc important to safoty under various' postulated scenarios. Those ai; ust f one. Include design basis events such as small or largn break LOCAs, and events of higher frequences like main steam linn isolation or loss of feedwater in which QlE22-F004 must open to provide spray coolant flow (FSAR 6.3 and 15A 6).
These may be coupled with failures of other ECCS or makeup systems. They also include situations under which the valvo is required to cloan manually to mitigate thn consequences of equipment malfunctions which could result in the release of radioactivo material outside of the containment. In any event, the consequences of such failures will bn no more severn under the higher torque switch sotting applied to t.his valve por the MNCR.
Thnrefore, thorn is no croation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report, The GGNS Technical Specification Bases discuss the function of the llPCS system and the QlE22-F004 valvo in particular. The valve must open and closo as required to provido its activo function for emergency coolant injection control (TS B3/4.5.1; TS B3/4.5.2) and remote manual isolation of the lipCS injection lino (TS B3/4.6.4).
It must also satisfy its passive function of pressure rntention (TS B3/4.4.3.2).
The margin of safety associated with the valvo's primary activo function involves the ability to open to provido cooling to prnvent exceeding fuel cladding integrity limits. Also implied in the Bases, and assumed in the FSAR (Tablo 6.3-2; Section 6.3.2.2.1), is that the valve will open in a timo period consistent with meeting the required overall system response t.imo of 27 sec., or faster. Because the valve will still perform its intended funct.lons adequately, margins of safety are not af fected.
The "alve will open when required to do so, and surveillance '
proceduros which verify the llPCS system's ability to respond in the necessary timeframo are unchangnd. The lipCS system will thnrefore remain capable of providing its design flow ratn within the bounds assumed in thn TS Bases and FSAR analyses and the margin of sa fety to exccoding funi cladding integrity limits is not reduced.
l NPE90/SNhlCFhR - 57
Attachment to GNRO-91/00001 NPE-90-068 Page 5 The margin of safety associnted with thn passivo pressure retaining functions are also not impacted. The lloses discuss t he requirement for the valve to remain intact to reducn the possibility of an intersystem LOCA (TS 11 3 / 4 . 4 . 3 . 9 ) . Evaluation of the resulting valve stresses, both operat ing and seismic, under accident conditions shows that the valve is no more likely to fall with the torque switch set as recommended. There are no changes to t he surveillance procedures used to verify valve integrity.
Thus, all intended functions of Q1E22-F004 will cont inun to be performed without degradntion, and thorn is no reduction in the margin of safet.y discussed or . implied in thn linses for the GGNS Technica1 Specificatfons.
1 NPE90/SNhlCF hR - 58
_m -- _ . . . - _ . . - -_ -_. - . . _ - . _ . _ . _ .. ..._. ...-.... - . _ _ _ _ _ ~ .. _ _ _
l Attachment to GNRO-91/00001 l
SRASN: NPE-93-069 DOC NO: DCP 90-0551-S0 6 SI-R00 SYSTEM: P41 l l
i DESCRIPTION OF CllANGE: This DCP corrects a potential common mode !
failure of SSW loop A and loop C return header to the SSW cooling tower. Additionally the DCP corrects a lack of design loop C SSW cooling flow to the liigh Pressure Core Spray (llPCS) room coolers.
Loop A and C will be separated by disconnecting the 10" Loop C return pipe from the 24" Loop A return pipe in the Loop A valve room. The Loop C return line will be re-routed to the SSW cooling tower previously designed for Unit 2 Loop A service. The modified Loop C return line will essentially be routed along the existing path for the Unit 2, Loop A return line from the Loop A valve room to the existing Unit 2, Loop A distribution header.
All submerged, Unit 2, Loop A, 24" return piping will be repinced with 10" piping for Loop C service.
The SSW cooling tower previously identified for Unit 2. Loop A l service will be re-defined for Unit 1, Loop C service. The cooling tower for Unit 1, Loop C service will consist of two cells operating as a natural dra f t cooling tower. The existing cooling tower fans, gear reducers, drive shaf t, and all associated drive shaf t components will be removed f rom the Loop C cooling towers.
Modifications to the distribution headers will include replacing the existing spray nozzles with small nozzles, and the repair of all construction welds not previously Code sthmped.
The existing leak detection will be modified to create independent loop detection systems. The flow restricting orifice plates in the SSW Loop C piping will _ be modified to provide design flow to the llPCS Diesel Generator rpoiing water jacketa and the llPCS room cooler. The scope of modifications to the restricting orifice plates is provided in Suppleinent 1 of DCP 90/0551.
The existing Loop C restricting orifice plates are located on thn main supply and return lines. These orifice plates restrict the i
total loop flow to all components served by Loop C. The flow resistance caused by these plates will be reduced by enlarging the i born of these plates. The core of QlP41-D014 will be enlarged to line size, which will eliminate all resistance caused by the plate. The plate will remain in the loop to prnvent necessitating removal of the flanges. The bore or Q1P41-D013 will be enlarged to causa less flow resistance, but will not be enlarged to line size since some resistance is still needed to limit the total loop flow rate.
- New restricting orifice plates will be installed in the pa*allel j branch supply lines to to the llPCS Diesel Generator jacket water coolers. The new orifice plates will serve to reduce the flow to the Jacket water coolers in order to force additional flow to the l IIPCS room cooler.
I NPE90/SNLICFLR - 59 L__ _ _ _ _ _ . . __ _-_ _ __ _ - __ _ - - . _ _
Attachment to GNRO-91/00001 '
NPE-90-069 Page 2 Drain htles are provided in the modified header to provido passive freeze protection for the header piping. The drain hole sizo has been reduced from 3/4" (as previously dnsigned for the 16" and 24" pipe) to 1/4". The smaller drain hole sizn is designed to '
minimizo the amount. of hot return water bypassing the cooling tower nozzles. The 1/4" hole size is considered largo enough to prevent clogging under normal conditions since the LSW pump suction screen is fabricated from perforated plate with 1/8" perforations. As with the previous design for the existing loops, two drain holes are provided for redundancy in the remote event that a single drain hole should becemo clogged.
The nozzles selected for the modified headeV are hollow cono spray i
nozzles fabricated from brass which is similar to the existing type of nozzles in the loop A & B headers. Hollow cono spray nozzles create smaller size water droplets (..id have larger internal flow clearances than full ceno apray noe.zles. The smallest internal passage of the holicw cono replacement nozzles in 1/4", which is larger than the 1/8" perforations in the pump suction screen.
Two types of nozzins will be used in the modified cooling towers.
One nozzle type is designed for installation about the heador perimotor, while a dif ferent type of nozzle is designed for installation on tho internt. locations of the header. The perimotor nozzles feature a spray cono angle designed to minimize apray against the tower wall which could result in tower fill material bypass. The nozzles located in the interior of the header fonturn a wide spray cone angle designed to maximizo spray coverage overlap. The spray cone angin for both nozzle types is designed to provide sufficient spray area coverage of the fill material in the event of a loss of an adjacent nozzlo due to clogging.
A thermal performanco calculation has bonn performed for utilizing the existing tower enlis as natural draf t cooling towers. The performanco calculation indicater that the limiting maximum flow to the tower is approximately 800 GPH per en11, for a total loop flow rate of approximately 1600 GPM. With a total loop flow rate i of 1600 GPM, the cooled water temperature is calculated to approach 90 F. The design of tho modified Loop C piping will limit the total loop flow to approximately 1000 GPM, which is far less than the allowable 1600 GPM, and will result in cooled water tnmpnraturns loss than 90 F. Although the design will limit the maximum flow to approximately 1000 GPM, operation between 1000 GPM, operation between 1000 GPM and 1600 GPM is acceptable. The flow difference between cells is calculated to be balanced to within approximately 4 percent such that nnither cell will approach tho 800 GPM per enll limit under the designed loop flow rate of approximately 1000 GPM.
I NPE90/SNLICFLR - 60
.~- sn-- v- , .
, - s ..-.o,,,,wa. > - ~ . ,,n. - - - - , ,,m v. -,,,w - , - - , . _ , . - - , , - . - n--,-n- m. --.--.w>,
Attachment to GNIC) 91/00001 1
NPE-90-069 Pago 3 I
RfASON FOR CilANGE: Two separato problem nrens have been iderit ified conentning the llPCS SSW. The first problem dente, with l
. the potentini common malo failure of SSW Lup A niul Loop C return
., hender.to the SSW cooling tower. In n postulated LOCA scenario .
i where the single failure in ESP Electitcal Division 1. the 1.ow l Pressure Core Fpray (LPCS) system and the Standby Servien Water l (SSW) synten Loop A would not hn nynilabin. This would lenvo the 3 liigh Pressurn Core Sprny (llPCS) system (ESP Elect rical Division !
111) as the only nyallnblo corn spray, in addition to the two Low i i Pressuro coolant inject ion (LNI) pumps for long term core l cooling.
in the GGNS design, thn lipCS service wat er and the division I service wat er both ri ,uin to the SSW tooling tower through thn common Loop A spray hondor. The rointively small return flow of the llPCS service water, without the added SSV return flow from L Division I components, would ho insufficient to provide ef fect ivn !
sprn/ over the SSW Loop A cooling tower fill. After approximntoly ;
50 to 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />, ilPCS service water temperature ceuld exceed the '
desigu temperature of 90*P, and the nynilability of the llPCS .
system may not be assure. !
t GE has performed on evaluation which demmnst rates ECCS criteria .
are met assuning no credit for coro spray cooling after thn initini 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> of IIPCS operation. Entergy Operations considered this cynluation adequate for interim operat lon. Ilowever. Ent ergy .
Operations commit ted to implement system modifications to attain [
adequate long term llPCS service water cooling prior to start-up ;
from RPO4. i The second problem deals with the lack of design Loop C SSW cooling flow to thn llPCS room coolers. The llPCS room cooler and i the two llPCS Diesel Generator cooling water jackets are designed in onrnllol flow paths for SSW Loop C flow. The component with the highest flow resistance in the llPCS toom cooler. The high room cooler flow resistance is caused primarily by the 1cugthy run l of 2" anl 2-1/2" piping. The lipCS Diesel Generntor cooling water ;
jackets and branch piping create very little flow resistance thereforn too vast enjority of ihn hoop C flow passns through the t cooling wat er jacknts. ;
The llPCS room cooler in designed to operato '<lth a minimum flow r rato of 40 GPM. A Pre-Operational Test documented a measured SSW
,' Loop C flow rato to the !!PCS room cooler of 22.2 GPM during 1982. ,
i The Pre-Oper at ional test documerit ed a st art-up exception to the Inck of design SSW flow to the HPCS toom cooler based on an ovaluntion which used the Log Hean Temperaturo Differenco (LMTD) to characterize thn performance of the room cooler at the lower flow-rato. The LMin met hod ,,a:; nes-conversativc for ovalunting
- the ll.'CS room cooler performance. ,
NPE90/SNLICPLk - 61 i
Attachment to GNRO-91/00001 l
l j NPE-90 069 l Pngn 4 i The llPCS pump room temperat ure could rench 166 'T wit h nr. SSW I,oop !
1 C flow rato to the llPCS room cooler of only 20 GPH. Itased on n !
postulated post-h0CA room t emperat ure of 166 'r, the expect ed i operating life of the llPCS pump motor windings was detetmined to .j be npproximately 64 days. The docuroentation of the llPCS Hoom 3 Cooler flow problem wan reported to the NRC. '
The system modifications were implemented prior to start-up from l kr04 and will provide for thn original design SSW flow of 40 GPM ,
to t he llPCS toom cooler. '
. SAPETY EVALUATION: This safety cynluntion concluded that thn i chango did not involve an unreviewed safety question. The SSW .
system, containing the plant ultimat e heat nink (UllS), is nn !
essential nuxiliary supporting system which is designnd to removo !
, heat from plant auxillaries that are required for a safe reactor shutdown.
The modifications modo per DCP 90/0551 do not affect the integrity I of thn PSW mnkcup 1ine, nor do the modificntions af fect tho 30 day basin inventory. DCP 90/0551 does not alter the existing t
configuration of the 1,oop A or hoop C pump seni. DCP 90/0551 dnes ,
not nitor the existing configuration of the boop A or 1,oop C '
roturn valvo located in t.hn 1.oop A valvo room. Supplement. I to .
DCP 90/0551 does modify the bore of the previously installed -
flanged restricting orifice plates in the llPCS SSW supply and return line and provides for additionni flanged orifice plates in !
the supply lines to the HPCS dicsol generator jacket water cooler. '
ilowever t he orifico plant modifications modo are desigund in !
necordance with ASME Sect ton XI, which mont s the originni j const ruct ion code f or thn llPCS SSW piping. Modifications shall bo in accordance with the original const ruction code except that NA i symbol stamping is not required. The safety related portions of the SSW system were originntly designed and constructed to the ,
requirements of ASME Section 111, Division 1, Class 3, !
Hodifications required by DCP 90/0551 are designed, and ;
installation requirements are specified, to meet the requirements t of ASME Section III, Class 3. Comptinnco with .dHE Section XI and !
thn origins 1 const ruction code ( ASME Sect ion 111) ensures continued pressure boundary integrity of thn SSW piping ind Com potle ti t n .
(
{~
l- !
t i
NPE90/SNhlCrl,R - 62 L._ _ _ _ _ _ _ _ ___ _ . - - _ . _ _ _
At t nehme it to GNRD-91/00001 NPE-90-069 Page 5 ECCS clectric power loads are rigoiously divided into Division 1, Division 2 and Division 3 so that loss of any one group will not prevent t he mininium sa f et y f unct ions from being performed. No int erconnect ion tr'od i f icat s.ons nre be ttig vnde por DCP 90/0551 which could comprornise redundant power sources.
Separation within the ECCS is such that controls, irist r ument at ion, equipment and wiring is sagregnted into three separate divisions designated 1, 2, and 3.
1,oss of cont rol power or bypass of any piece of equiprnent in the SSV system is cont inuously itullent ed in t he cont rol room.
SSW syst em todutulancy will he maintained following implementation of DCP 90/0551. Inst rument at ion nrul cont rols nssocint ed wit h the three logic t rains of the SSW system are physically and elect rically separat ed and meet all separation requirements irnposed upon redundant sa f et y reint ed ci rcuit s. The separation criterin for redundant Clnss 10 circuits niul equipment within the ECCS nasures that. the failure of equipment of one redundant system ennnet disable circuits or equipment essent ial to the operat ion of the other redundant systems.
The modifications rnnde per DCP 90/0551 will piovide for complete separation of Loop A f rom 1oop C therefore providing for independence of operation between the two loops. The failure of either hoop A or 1,oop C pump will not affect the operat ion of the remaining basin A pump. DCP 90/0551 does not alter the conf igurat ion of SSW l.oop B.
DCP 90/0551 does not alter the existlug provisions for nonessential system intertie isoint ion vnives or the tedundant int ert ie isolation valves. Yalve nodifications por DCP 90/0551 are designed per t he t equirement s of ASML Sect ion F1 and per the requirement s of t he design specificat ion for the SSW system. DCP 90/0551 does not alter the design of the Loop A cooling tower fan.
The revised design of the llPCS SSW per DCP 90/0551 utilizes natural drnit circulation t hiough t he Loop C cooling tower, therefore, the revised llPCS SSW does not ut(lize fans for rnechanical draf t cooling of the hoop C return wat er. DCP 90/0551 does not alter the exist ing diesel generat or londing sequence, nor does the DCP ndd or delete any londs from nny diesel generators.
The incronsed llPCS SSW pump flow does n1ter the actun1 IIPCS SSW pump motor electritnl lond. however the inccensed electrien1 lond is bound by existlog Diviston 111 generator lond and fuel consortpt ion nna lys is .
The rnod i f icat ions por DCP 90/0551 have been evolunted for consequences of moderato energy line bronks.
NPE90/SNhlCFhK - 63
At t arbtnent to GNKO 91/00001 NpC-90-069 page 6 The c on ponent s niul suppor t inn st ruct ut es of all moclif leil equipment that nic not seismic Category i nnel whose collapse coulti tesult in loss of a requise<l functton of the SSW system t hiough eit her impact of f lootlitig wer e evnlunteil to ensule that they will not c ollnpso when subjortort to seismic l on tl i n g .
Railint. lon toon i t or s nie pi ov bleil on 1. oops A 6 11 t o clot ect.
cont am inn t 100 tesulting liom kilk heat exthnnger tube lonks. The provisions for rnrlint ion <let ec t ion in hoops A 6 11 will tiot be alteicil by DCp 40/05;1. Loop C <loes not itit er f ace ilit ect ly wit h cont aminnt ecl syst ems , theiefore 1.oop C is not inc lutteil in t he post ncrielent snnpling system, lloweve r , thn snoct i f l eil 1,nop C w i l l be prov bleil wit h a lorni snmple st at inn for cont aminat ion monit orit%
in the event that bns in A her ornos < ont aminnt e<l f:om Kilk hent exchanger s vin 1.oop A s equiring l oop A isoint ion while 1.nop C opeint ion is st ill requireil.
- l h o SSW cooling toweis , bnsins, nial pump houses a r c clos ignett t o w it hst ntul, w it hout a loss of functlonal cnpnbility, the following nnturni ph,n orn e n n :
- n. Entthqunko b, Mnximum piobable flooil elevntion of 4103 feet above menn son levol,
- c. Tornolo winct f oi ces anil tornntlo-boino missiles.
The SSW coolitig towers, basins, nrul pump houses are const ruc t e<l o f concteto walls atul i nof s at least 2 foot thic k. The SSW cooling tower f ans nre provirlert wit h ilebris piot ect or s t o prevent c inmi t aneous failure of the fans from t oi nntle ~nt ra lneel elebr is.
The 1 nop C cells will retain the missile protectton previously a f f orilert t o 1. oops A 6 11 f100<1 protoctlon is not alt et ett by implementation of DCp 90/0551 since the St atutby Seivice Wnt er pump llouses nio not n f f cct c<l by t he DCp.
Since DCp n0/0551 rnaint n ins the stiortura1 Int egr it y of t he SSW bnsins nnit toweis, thn f at tor or sofety for buoyancy f or so isin it Category I structutes is not a l t e t otl. The builtlings contnining ESF components hnvo bnen <les ignoci:
- a. t o wit hst atul a ll ci erlible tnoteorological events an<l torontlos
- b. seismically niel will inmnin f unct ionnl ilur ing an<l f ollowing a safe shut tlown ent t hqunke (SSR).
- c. to protert the hSi systoms in the event o f a pos t u la t e<l f i t e.
Np090/SNhlCFLR - 64
Attachment to GNRO-91/00001 NPF.-90 069 4
Page 7 )
I
- d. for protection against dynamic effects associated with the j postulated rupture of piping. ,
i j o. for protection against missiles Tailures associated with the SSW system do not result in tho l' initiation of accidents. The Ull5 st ructure is capable of withstanding the of fects of the most severe natural phenomena associated with the pinnt location, other applicable sito-related ,
events, reasonable probablo combinations of loss sevnte phenomena, !
j and any credible single failure of any man mado structural features without loss of the sink capability to provido the !
necessaiy heat rejection. Whern protective action is required under adverso environmental conditions during postulated ,
! accidents the SSW system components are designed to function c ;
under such conditions.
Control room annunciation in provided for leakagn from the SSW 1 system, i.eakage can also be detected by a high lovel alarm from ,
any one of the sumps located throughout the plant. Botn high :
alarms and standby sump pump operation signal' arn monitored by thn plant computer. '
Thorn are no existing multiple not points in the SSW eystem.
DCP 90/0551 does not create any mult iple set point s.
Access to all means for bypassing the SSW system is under administrative control. DCP 90/0551 does not alter the administrative control or automatic system design logic.
DCP 90/0551 does not alter tho total heat rejection loads to the UliS . j The SSW system is designed to perform its required function for all modes of system operation. Previous analyses of system operation for the various modes have determined that Hodo IV is the crit ical modo for evaluating the capnbility of the SSW system !
, to perform its safety function during singic unit operation. Hodo IV is defined as a LOCA in Unit I coincident with worst singlo y active failure and total loss of offsito power; with Unit 2 ,
non-operational. ,
The Safety F. valuation for the SSW is af fected by implementation of 4
DCP 90/0551 by changing the heat loads delivered to the existing loop A cooling tower. Mode IV cooling requirements for shutdown of Unit I havn been previously evaluated and are satisfied by SSW Loop B and ilPCS Servico Water Loop C. Thernfore t.hn modificat ion was ovaluated using mathematical techniques previously, used for modoling Mode IV heat rnjection.
i NPE90/SNLICTLR - 65
.4 --*e-r-,m- -,,~r--,-,-----m----.-,e,y-,-,
er eec,*.-.,, % w-,-rcwn,,-,-,-,-,,- ,,,,w,,.,,+,pw--yr---,-...wu,r,--r,--m-,www,-,--p-----e,w
Attachment to GNRO-91/00001 NPE-90-069 Page 8 The heat rejection evaluation for the Loop C natural draft cooling tower shows a cold water return temperature of approximately 89.3*r with an ambient air wet bulb temperaturn of 79'F and an ambient air dry bulb temperature of 100'F. The evaluation was based on a basin water temperature of 90*F. a Loop C flow rate of 1000 GPM, with a constant peak Loop C heat load.
The worst one day of UllS cooling tower demand occurs on the first day of the 30-day period following a LOCA. The actual basin water temperaturn is at its lowest on this day and will bn at a maximum initial temperaturn of 75'r, assuming highest PSW temperaturn.
The cooled SSW return water mixes with the large basin water volume, resulting in a lower actual SSW pump suction temperature than the 90*F assumed in t ho evalunt lon. The cold water return temperature will thnreforn not actually be as high as 89.S'F.
As the basin levn1 decreases dun to drif t and evaporat f ou, t he Loop C flow rate will decreann. The decreano in Loop C flow will result in a higher Loop C return water temperaturo which will improve the performanen of the natural draft cooling tower. The improved performance of the natural draft cooling tower will result in a lower Loop C cold water temperaturn returning to the basin. The llPCS SSW performance analysis verifies that with a gradually depleting basin inventory, the basin water will not exceed the design temperature of 90'r.
Losses from the SSW basin inventory result from cooling tower i
drift and evaporation. The evaporativo loss is a function of the meteorological conditions and the return water temperature. Since the applicabin meteorological conditions are identical for both the llPCS SSW and the Standby Diesel Generator SSW, the ovaporative losses determined in the Hodn IV analyses arn considered valid for thn separato utilization of tho - IIPCS SSW natural draf t cooling tower and the Standby Diesel Generator SSW mechanical draf t cooling tower.
The Mode lY estimate for a 0.02 pntcent drif t loss from the Grand Gulf SSW cooling towers has been supported by drift climinator performance tests conducted by an independent. testing firm. The tests were performed on both a test cell and an actual operating enll whose sizo is similar to those for t hn Grand Gulf SSW towers.
The drif t climinators used on the cells tested were of the same zig-zag design as those used on the Grand Gulf towers. The test results found that dri f t losces to be less than 0.000018 porcent.
Based on the results of the tests conducted on the drif t climinators, the 0.02 percent estimated drift loss f rom t he Grand Gulf towers is consnrvative.
NPE90/SNLICPl.R - 66
Attachment to GNKO 91/00001 NPE-90-069 Pago 9 The llPCS SSW natural draf t cooling tower will have a lower air velocity through the tower than the previously analyzed existing mechanical draft cooling toworn. The lower air velocity should result in a lower drif t rato, therefore the drif t rato preciously documented for the mechauf cal draf t cooling towern in cotinidored to be conservativo for the natural draf t cooling towor.
l'or a 30-day period of operat ion following plant shutdown a design conditions, the existing Modo IV total integrated water loanen resulting f rom cooling tower drif t (of approximately 103,000 gallona) are considered to represent the num of the lonnen from the 1,oop C cooling t ower tind the 1,oop A cooling tower since the basen due to the 1. cop C flow worn previously included in the Modo IV analysen.
The llPCS SSW natural draf t cooling tower design in such that variations in wind spend and direction any temporarily affect the tower's performanco, llowever variations in wind speed and direction are considered to be sufficiently random and dynamic thus precluding any consistent, extended degradation in tower performanco. During the peak heat load period of SSW cooling tower operation, the basin water in at a temperaturo lower than the maximum allowable water temperature. Bocanno the hanin contains a largo volume of water, any short period of wind af fects or air temperaturen higher than the design air tonneraturen in not expected to criuno the basin water temperaturn to exceed the design limit. There is thereforo no reduction in the margin of safety an defined in the bania for any Technical Specification.
NpE90/SNhlCI'hR - 67 J
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1 At t achinent to GNEO-91/00001 SRASN: NPE-90-070 DOC NO: CN-90-0391 SYSTEM: N64 DESCRIPTION Ol' CilANGE: This change provided for new top connections at the bottom of the 6" pipe E P.D- 2 5.
REASON FOR Cll ANGl'.: Due to pipitig interference, a new locntion wan fcund to install the fluid Components, Inc. (FCl) model hT-81-4 flow instruments.
SAI ETY FNA1.UATlDN: There is no increnr.c in the probability of occur rence or in the cone.equences of an accident or malf unct ion of equipment important to safety pieviously evalunted in the Safety Analysis Report. FSAR Section 15.7.1 states the equipment ntnl piping nre designed to contain any hydrogen oxygen detonat ion which has a reasonnhin prohnhility of occurring. A detonnt ion is not considered as a possible failure. The new flow elements have been purchased to the original design drawings and specificat ions, thetofore the original ren f ety analysis is not compromised. The piping designs issued by this change meet ANS1 1131.1 requirements and nre qualified for the appropriat e dendweight and thermal lands. The piping will function in its int ended innnner . No other accident precursons evnlunted in the UFSAR nre affected by this chnnge. Therefore, there is no crent ion of a possibility for an accident or malfunction of n different type than any evnlunted previously in the Safety Analysis Report.
The design change doen not offect t he opernt ion or funct Ion of thn of fgas syst em as defined in the bases for nny Technical Speci f icat ion. Therefore, all renigins of safety as dnfined for any Technical Speci ficni ion t hus remain unchanged.
NI E90/SNI lCFl.R - 68
At tachment to GNRO-91/00001
! i i
SRASN: NPE-90-071 DOC NO: HCP-89-1098-S00-R00 SYSTEM: M41 f
l DESCRIPTION Or CllANGE: The exhaust valves and ASCO solenoid
- valves of eighteen (18) Bettis air operated valve actuators woro
.I ' '
replaced with different Asco solenoid valves. The actuators are nn integral part of several butterfly valves. The associated isolation valves are identified as follows; (a) 1H41r007, roca, ,
T036, T037, (b) IT41r006, 007, (c) IT42r003, r004, T019, F020, roll, r012, (d) SZ$1 r001, T002, F003, T004, T010, roll.
REASON FOR CHANCE: Three secondary containment isolnt ton !
dnepor/vnives failed to close within their Technical Specification s limit of s 4.0 seconds. The cause of the failure was attributed .;
to the quick exhaust valve (Parker-liannifin Model OR50 OR OR50ll) t installed on the ficttis air operated valvo actuators. ,
J Soveral system M41 air operated valvo actuators have been successfully modified without the quick exhaust volves that hnd been the cause of the failure, liased on these modit'icntions, the installed ASCO solencol valves and exhnust valves were replaced with different Asco solenoid valves successfully used beforn. The new solenoid valves have a larger orifice than the original equipment and uso inrger actuator nir exhaust tubing than tho ,
original installation so that the exhaust nir flow rato is limited to n largo degroo by the pressure port size of tha nir operated '
valvo actuator.
SATETY EVALUATION: This safety evaluation concluded that the chango did not involve nn unroviewed safety question. No :
Isolat ion dampor/valvo cont rol logic / circuitry has becu changed by the MCP. The MCP implementat ton maintains the required mnximum operating time whfic climinating a potential valve failure source (quick exhaust valves). The modifications which consist of a l different. sizo solenoid valve and larger diameter tubing has been
- analyzed and determined to be seismically satisfactory. Since the remaining pneumatic components are standard items for this type of
- installation, successful implementation of the MCP will improve thn dampor/vnivo operational rnliability.
In order to compensato .
for the exclusion of the quick exhaust valve the resistance of tho l flow path through the tubing and soinnoid valve will be decreased
! by increasing thn tubing and solenoid valvo orifico diameters.
Thn use of a solenold valve with a larger orifico does not increase the likelihood of any f ailurn. The tubing modifications havn also been annlyzod for seismic concerns and are satisfactory.
l A potential sourco for malfunction of thn isolation dampor/valvns I has been Flentified by the actuator supplier via a 10CIR21 report concerning ti.3 uso of Mobile 28 grease and Ethyleno-Propylene j men ts within the basic actuntor assembly. The af fected actuat ors were rebuilt to oliminnt.c the problem. The solenoid valves, although a potential source for malfunction, havn not been NPE90/SNI,1CrlR - 69
,.-.,,-,..._,,.....,,_._____.,,m._ ,,-..__._m,,,-,_,m, _ . - , , _ . . , , , , _,,_y.,,y_.m,_. , _ . . . _ _ , . , , _ . _ . _ _ , , . , , , . . , . _ . . . . ~ , .
Attachment to GNKO-91/00001 NPE-90-71 Pnge 7 identifie<l with any getieric failure mode i ti the prehent applientions.
The isoln t ion darnper/vnives can only fall to isointo within t he 4 socoruls seguired if the initiat ion circuit r y f ails or the pneumntic niul rnechanical component s dir ect ly nasocint ed with the valva rnnifunction. The MCP affects only the pneumatic or electio pneumnLic components associat ed with t he vnive nctuntor, implernent at ion of t he MCp will result in the reniovnl of the qutck exhnost valves. The remainder of t hn pneumatic components will be the same na those presently installed except for internal d i n tn e t o r s and therefore their potenttal failure modes will be identital.
Since the only changes involve the pneumnt ic control flow path for the isolnt inn dnrnper/vnives niul no changes nre made in any isolat ion cont rol logic or elect rical component s (other than Asco solenoid valve sizn), only the performance of the isoint inn vali>
requires evaluntion. The MCp implementntion providen t he snine valve actuntor function while excluding n potential isolation damper /vnive mnifunction <,ource (quick exhnust valva).
A Calculation was perfoined to assure that the design does not increase t he pneumnt ic damper /vnive blow down t ime when cornpared with the present exhaust valvo installation. The not i f icat ion wjil result in a lass complex closure scheme which will maintain the Technical Spa cificat ion required closure t irne of < 4 seconds.
NpE90/SNhlCF1K - 70
At t at.hrnent ta GNRO-91/00001 SRASN: NPC-90-072 100 NO: !!C1' 1102 SYSTEM: C91 DESCRIPTION Or CilANGE: Thin MCP providen electrical details for l
the installation of a permanent power cable (routed in condult) i between Panni SC41aP890 and disconnect switchen 08-1Y91-24 & 26 in '
the 120-240VAC uninterruptible power panel 1Y91. This involven removal of the extating power cables and installation of new power ,
cabica routed in conduit betwcon the two pancia.
REASON TOR CllANGE: To provido a permanent power supply in thn f place of the temporary supply previounty being unod. l SAPETY EVAL,UAT10N: This nnfoty evaluation concluded that the chnngo did not. involvo an unreviewod anfoty question. This design ;
change inst alin permanent power to non-nn foty relat ed comput er "
panol SC91-P890 fod f rom disconnect switchen 08-1Y91-24 626 in tho 120-240 VAC fl0P uninterruptibio power nupply (UPS) panol 1Y91.
Distribution panol 1Y91 in fed f rom the station 125 VDC Non-Clann ,
1E battcry and battcry chargen which are connectod to one of tho !
clann IE bunnen. Tallure of any of the equipment in the 125 VDC l aupply circuit enables the static switch to transfer the power nource automatically to an alternate source fed from a 480 volt i Clann IE AC bus through a transformer. When a 1,0CA occurs, the Clann 1E feed from the lond center that foods thn chargern in l tripped. Thoroforo tho malfunction of loads to lY91 will thus bo bounded by- a 1,0P-I,0CA.
l The implementation of this design change will not af fect any !
equipment ident ified as the basin for any technical specification, t The design adds permanent pc< wor to It0F computer panol SC91-P890 from Non-Class IE uninterruptibic power dist ribut ion panel 1Y91.
The 110P computer and the 120-240 VAC 110P uninterruptible power -
nupply system nio not addressed in the GGNS-1 Technical ,
Spectfications, i l
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I NPE90/SNLICFLR - 71 1 i
Attachment to GNRO 91/00001 1
SRASN: NPE-90-073 DOC NO: HCP 1103- 800- R01 SYSTEH: P81 (
t DESCRIPTION OF CllANGE: The installed air regulators for start ing i air regulator valven P81 PCV-T505 A [li) and PCV-T506 A (P) for tho ;
ilPCS diesel generator are Norgren model (1 R02-200-RGS-AU. This '
amini regulator has been replaced by a model Ril-200 RGSA or a modn1 R08-200-ROSA. !
REASON FOR CilANGE: The old model has been discontinued.
i SAFETY EVAbUATION: This safety evaluntton concluded that the '
chango did not- involvo an unreviewed nafety question. The replacement regulators maintain the namn form and functloti as the '
original model. Mounting hardware will be modified to allow the new model to bc . inst alled. Evaluatton han shown that the modola :
and the revised mounting hardwarn will not compromise the original soimmic qualifications. The new modol number and installation will meet the original form, fit, and function. Thorofore ihn i start air pressure regulator valvos P81 PCV F505 A[B] and PCV F506 AlB] will function an originally designed. Evaluation han shown that the models and the revised mounting hardware will not compromiso the original snismic qualifications.
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6 NPE90/SNI.lCFl.R - 72 P
. - - . . . . . . . , ._-,._.-r- , . ., ,.4 ., _ . _ . . . . , ,w_,,mm,,,,,,.g.,~,--m,e.. m.m.,, e,.ww..r...vwy,-,-,r,-- ...,,.wf- ,.w-.y,o , - .c..vcy---,w-
Attachment t o GNRO 91/00001 SRASN: N Pl;- 9 0- 0 74 110C NO: MC P- 8 9 - l l 3 'i- 5 0 0 - K 0 0 .u STi;M : 113 3 DFSCRIPTION OF CHANGl;: 'This MCP disables the non loop mnnunt modes by hasdwiring the " Reset flow Control to Manual' trip. No erpil pmen t is to bn physically removed or elect rically disconnect ed from it power source.
The only mode of opeint ion of t he Retire flow Control vnives used at GGNS is 1.oop Manunl. This MCp disables all other "non loop manual" modes.
SArl:TY INAl.UATION: This sa f et y evn'ont ion c oncluded t hat the change did not involve nn unreviewed safety questton. This MCP only disnbles the non loop mnnuni modes. Cont inuous opernt ion in loop manon1 mny reduce t he probabilit y of n Koclic flow Cont rol rnilure (Increnning Flow) and n Kocirc flow Control Uniture (Decreasing Flow). This is because a inrge proportion of the flow cent t ol inst rument at ion is bypassed when thn system is in loop mnnunt mode. With less active equipment, there should be a mathemntIcnl decrease in the probnbility of an equipment failure that could enuso these events to occur. The classifient tons of t bene event s (incident s of modernt e frequency) wi11 not be changed. No other evnluit ed nccident s nre predicated on a failure of the recirc flow control system and no other systems / system components are affected by this change. Operating in loop manual may decrense the consequences of t hose events beenuse only one loop is postulated to fnil instend of two as in non loop mnnunt modes.
No saf et y relat ed equipment is affected. Thin MCP simply fortes thn recirc flow cont rol system to loop manual by placing jumpe rs neross thn :ontacts of non safety scinted relnys. There is no addit lon, delet ion or modif f rat ion of any ASMl; companent or pressuen boundary involved. There is no addit lon, delet ion or modi ficnt ion of any clnss 10. component or circuit invoiva l, l.oop mnnual and nnn loop mnnual modes of recirc flow cont rol have been sepnrnt ely evnlunt ed. Iloth modes of operntion have been approved. This MCp simply disables the non loop manuni modes, e Therefore, there is no reduct ion in the mnrgin of safety as defined in the basis f or nny Tor.hn ical Speci f icnt ion.
e s A s 1
Attachment to GNRO-91/00001 SRASN: NPE-90-075 DOC NO: HCP-89-ll26-800-R0-R1 SYSTEM: 112 1 DESCRIPTION OF CHANCE: HCP 89-1126 replaces the solenoid pilot i valves used on the inboard and outboard Hain Steam Isolation Valves (MSIV) H21-1022A, li, C. D and B21-r028A. B, C.11 The ASCO duni solenoid valve model NP8323A20E is boing replaced with two ASCO model NP8320A185V solenoid valves.
REASON FOR CHANGE: ASCO no longer manufacturers thn old model solenoid valves. In addition. GONS has experienced problems with this solenoid valyn model not going to their drenergized posit ton.
These f ailures have been caused by ext rusion of the EPDM seat ing j material into the valvo body. The mechanism for thf N failure has
. been at t ributed to degradation of t he EPDM due ta clovated i temperaturn. The valves are not only subjected to a high ambient temperature thoro are also exposed to a highnr tempornture riac becaano both coils are continuously energized. Adding to this failuro mechanism is the high seating force continually applied to the sent ing material by the 11 solenoid.
A calculation was dono and the results indicate that an ASCO NP8320A185 with a 3/32" orifice would minimize the impact on MS1V response tima, The new model solenoid valves uso viton seating material. Replacing the dual solenoid valves with two singin soinnoid valves will reduce the expected heat riso by 30 degrees centigrade.
SAFE 1Y EVALUATION: This safety evaluation concluded that the chango did not involvo an unrnviewed safety question. The chango performed by this HCP will not alt.er the HSIV trip logic.
Thoroforo all safety action required by thn MSIVs will not bn altered. The evaluated event in the FSAR is an increase in roactor pressuta dun to a MSIV closure. The influro mechanism within the ropiacement valves for a HSly closurn would be a !
failure of the seating material to maintain it's pressurn seal !
thus allowing the HSIV to go to the closo position. It would bo ;
expncted that. this kind of failuro would be similar to the failures of the snating materini experienced throughout tho ,
industry and in particular here at GGNS. One of the causes for '
this failure of the seating material in it's exposure to clovated temperatures. Replacing the singin dual solenoid valvo with two ;
single solenoid valves will reduce thn expected valve temperature i rise by v 30 *C. An expected cont ributing fact or to the failurn -
of the seating material to maintain it's seal would bn the seating !
forced experienced by the seating material dun t o t i.e "B" i solenoid. Replacing the NP8323A20 valves with two NP8320A185 valves will greatly reduen this seating force. A reduced qualified life for thn replacement SVs is expected. This new qualified life is based on the unn of viton seating material.
l Analysis shows that tho use of viton will not impair the valvo's l
3 NPE90/SNLICFLR - 74 .
Attachment to GNRO-41/00001 !
1 p
NPE-90-075 Page 2 :
ability to perform it's safety function when expo 6ed to the expected radiation dose over the qualified lifo. In addition tho l new solenoid valves and tubing arrangement has been analyzed to l 1
ensure that scismic qualifications have not been tempromised. A i Stress calculation has boon performed to ensure the now tubing '
configuration will not loose it's pressuro retainin3 capabliity [
beforn during and af ter a Safe Shutdown Earthquako (SSE). ;
The absence of the second coil will also reduce the seating force on the. valvo, in addition thn new SVs will use viton not EPDM seating material and have a roduced qualified life. Ililler has noted that the closing spring force is slightly greater for the NP8320 valvo. Ilillor has also noted that the NP8320 has been !
successfully used on lillier operator npplications similar to the !
MSlV. The NP8320 valves are fully qualified and thn revised tubing configuration has been fully analyzed to ensurn it will ;
function bnfore during and after an SSE. The probability of a 1 failure of the pneumatic /ilydraulic unit to operatn their l respective MSIV has been decreased by this changn. Thornfore, ;
there is no creation of a possibility for an accident or i malfunction of a dif ferent type than any evaluated previously in -
the Safety Analysis Report .
The operating time of the MSly is a minimum of 3 secondr. and a maximum of 5 seconds. A calculation was performed and has ;
demonstrated that the installation of two Np8320A185 valvns with a 3/32" orifice will provido a sligi tly faster response tire of tho ,
pilot pneumatic circuit then the presently installed NP8323A20 valvo with a 1/16" ort f f en. Thernfore, the maximura time of 5 snconds will still be achievable. From review of past MSIV time response data it can bn determined that suf ficient adjustment is ,
available to compnnsatn for the slight increano in pilot operating '
time. Therefore, the minimum operating timo in achiovable. l i
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NPE90/SNLICThR - 75 j i
. . . . - - . , . . - . . - . . . - - - - ~ . . - - . . . _ . . . . - . _ . - - . - - . . , , - . . - , , . . ~ . - - , , - - . , - , . . - , - . -
Attachment to GNRO-91/00001 l I
SRASN: NI E-90-076 DOC Not MCP 90-1004 800-R00 SYSTEH: P41 ;
i DESCRIPTION Or CHANGE: The purpose of this HCP is twofold:
provido a removable spool piece in the SSW makeup water line to nanin "A" and install thn injection line in flasin "A" for thn ;
future SSW chemical injection system. ,
The SSW n...keup supply lion, 8" JDD-174, will bn modified by the I installation of 2 pair of flanges juat downsteam of valve !
NSP417504A. Also, line 3/4" JHD-1205 will require minor design :
changen to allow inst allation of the flanges. I l
a The installatfon of the injection linn, JZD-40, for the futurn chemical injection system, in 2" diameter pipe mado of carpenter 20 alloy. It originates outside of the pump house on the north i side. This outside portion consists of a blind flange and a plug i
valve. The line enters and exits the pump house through two new ;
penetrations. It descends to elevation 76' passing through the ;
e debris screen to a point between the SSW pump Q1P41C001A-A and the llPCS SW pump QlP41C002-C. It is located and supported as to preclude any possible failure that could affect the operation of i the SSW system. !
REASON FOR CHANGE: The removable sp ol piece in the basin makeup water supply line was provided to allow for the inntallation of a 7 temporary filter system durita the refill!r;g of the basin following a drain down.
Provision was made for tho inntillation of an SSW system.
Chemical inje, tion system by a future DCP. The change was made at this t ima becausn t he "A" basin was drained.
SAFhTY EVAhUATION: This safety evaluation concluded that thn i
chango did not involve an unroviewed safety question. The safety function of the SSW system, containing thn plant ultimate heat sink (UHS), is to provide a reliable source of cooling for plant ,
, auxiliaries that are essenttal to a safe reactor shutdown. Thn t S8W system is designed to perform this cooling function following -
a design basis loss of coolant accident (h0CA) automatically and '
, without operator action assuming a single activo failure l coincident with a loss : offsite power.
The SSW system original design as described in the UFSAR has not ;
changed as a result of the installation of the described flanges or the injection line. The piping and pipe supports installed by ;
this MCP have been designed to ANSI B31.1 requirements and are t qualified as seismic category 11/1. Plant operation with this piping installed in the SSW system will have no adverse effect on ,
the functionality of system required to mitigato the consequences of postulated accidents ovaluated in the UFSAR.
NPE90/SNhlCFhR - 76
_ _ . . _ _ _ _ _ ._ _- _ _ _ .. _ .. _ _ .__ _ __ _ . - . - ~ _ -
Attachment to GNRO-91/00001 I
i NI'E 0 76 i page 2 [
The modification of the SSW makeup lino by the installation of the {
flangen will not iequire a change in operating the syst em. The
) installat ion of t he injection line will not impact operation of l the system. The addition of the flangen to the SSW makenp lino '
will not chatigo or af fect its function. The design of the chemical injection lino's dinchntge sparger, which will be located j in the SSW hanin sump, is consintent. with the domign of the debris l l ncreen over thu sump with respect to prevent ing part icleN greater !
than 1/8" dinmet er from entoring the SSW pump muct ion. Also, the dischargo sparger is located and supported an to precludo any 3 possible failuro that could affect the operation of the SSW I system. ;
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i NPE90/SNLICM.R - 7 7
.......,..._..m, -. . , _ - __ . _ . . . - . _ _ _ . . . _ _ . . _ . . , , . _ . - . .. ,m_ - ., . . . , ._m.,-- ..,_,,_.c,_,_.-,,--.__..,. ,,-,
2 Attachment to GNRO-91/00001 i
i SRASN: NPE-90 077 DOC NO: HCp-90-1007-S00-R00 SYSTEH: E21
- l i t
] DESCRIPTION OF CllANGE: This HCP changen the makeup water supply
- to the refnrenco leg of suppicanlon pool level transmittern I C61-1.T-N402A, 030-LT-N003A, 30, and 4A. The new supply in from !
instrument valve E21pX020 located on the bow prennurn Coro Sprny (LPCS) jockey pump dischargo lino, The old nupply was from E21FX013 located on t he LpCS pump discharge lino. The supply ;
, tubing was rerouted to the new location. 021FX013 will be capped off and nbandoned in placn.
{
i REASON p0R CllANGE: A !.pCS punip start would couno the suppression [
poci invel monitoring t ransmit ters to go into an alarm stat o, thun making up hnif t he logic required to dump the upper containment pool. This was caused by the pressuro nurgo in the supply water ;
lino when the pump started. The new nupply la not susceptibio to j this problem and will keep the referenco legs full. -
6 sap 0TY EVALUATION: This safety ovaluntton concluded that tho l change did not involvo an unroviewed safety quentfon. prennutly, !
1 t he sent pots can receive makeup water f rom either the LpCS pump !
or the LpCS jockey pump. When implemented, the design change will prevent thn 1,PCS pump f rom being used for this purpose. Thin in i acceptable, because of 1.PCS jockey pump in environmentally nnd j scismically qualified. Also, thn suppression pool maknup system ;
connists of two independect , 100 pntcent cnpacity subsystems which arn divisionally separated. Thus, tho influre of n single activo .
c ompon.in t (including the I.PCS jockey pump) in either subsystem !
will not cause a loss of suppression pool makeup capability. The ;
valvo, tubing and t ubing support changes meet all applicabin !
! snismic/ASME Section 111 Clann 2 design requirements. L 1 i The E21. E30, C61 syst em oparnt ton and function will not chango. ;
the ins t rument valve, tubing and t ubo support s supplied by t his ;
MCp meet all applienble noismic/ASME Section 111 Class 2 design f requirements and will function in thnir intended manner.
Thn modj ficat.f on of the volves, tubing and tubn supports docq not l chnnge the limit ing condit ions for opnration applicability or surynillance requirements. Tiso setpoint s of the suppression pool
, level t ransmitters are not af fected. i 1
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. i NpE90/SNLICphR - 78
I I
Attachment to GNRO-91/00001 )
i 1 SRASN: NPE-90-078 DOC NO: HCP-90-1017-800 RO R1 SYSTEH: N71 i i DESCRIPTION OF Cl!ANGE: This design changn will fustall mnnunliy-operated " Mud Yalves" (NIN71T384 through NIN71P395) in {
the lateral flumes of the naturni draf t cooling tower. l
- j. !
REASON FOR CHANGr.: The new design f acilit ates on-st ream 11ushing j of the flumes by allowing accumulated medimentn to be flushed ;
directly into the cooling tower basin during stntion operation. '
Additionally, this capability will help to allovinte st ructural concerns relative to the flumes and supports structures due to necumulated sediments. Tho implementation of this design has provided a significantly less laborious and time consuming method ;
of draining and clenning thn flumen during mnintennnco outngen.
SATCTY EVA1.UATION: As postulated in the UFSAR, greas failurn of
, thn circulating water system butterfly valves and/or expansion joints results in flooding inside the Tuihinn ilulldings, Radwnste llulldings, Control !!uildtug, and thn Unit I rodwnsto pipe tunnel. ,
Thn Circulating Water System is a closed loop system, and failure of thn " mud valve" design would result in water passing directly from the cooling tower flumes into the tower basin, and no i additional water would be added to tho system. Thereforo, no ;
increase in aren ilooding would occur should these system componentn ini). The GGNS UTSAR niso evaluates the Circulat ing Water System for potential flooding of safet.y-rninted equipment i
due to failurn of a system component. The only safety-reinted +
equipment in the vicinity of thn condenser room below clovation *
, 116 font is valvo Ql-P44-F116, a secondary containment isolation
- i. valve. Failurn of this valve due to aren flooding will not ,
adversoly affect attaining and maintaining a cold safe shutdown. '
railure of the fiumes or the " mud valves" f uside the cooling tower ,
will not increase the probability of flooding, and consequently I cannot. Increase the probability of valve Q1-p44-Fil6 ,
mn1functsoning.
Design installation will be in accordanen with required standards l n:nl specifications and will enhanen flume clenning and drainnge.
l Aron flooding due t o syst em or component failure in the only postulated accident evalunted for thn N71 system. Gross failure of this dnsign will result in water and debris being deposited in the tower basin. Thir situat ton could causn n decrease in system ;
performance, but will not creat e a possibility for an accident or l l malfunction of a different type t han any cynlunted previously in the Safety Analysis Keport. ;
I -
In addition, the design does not involve .installnd inst.rumentation that is used to detect, and indiente in the control room a significant abnormal degradntion of the renctor coolant pressurn boundary. The design does not involve a process varinhle that is l
l NPE90/SNI,1cFl.R - 79 l
l l- _ ,-.._.._m.-, .,... _ .m. , _ . . , , . . , . . _ . . _ _ , , _ _ , _ . _ . , , . _ , . - . .,-,._....,,-r,..,_. . , . . , , ,_.,.mr., ,,,,.___a
l Attachment to GNRO-91/00001 j 4
i NPE-90-078 Pagn 2 ;
1 nn init int condition of a Dnnign Benia Accident or Transient l Analysen that althet assumen the influre of, or presents n challengo to the integrity of a finnion product barrier. The !
design does not af fect n st ructurn, system, or component. that. is !
part of the primary succean path and which functions or actuates !
to mitigate a Design 13nsis Accident or Transient that either !
, assumes the failure of, or prosents a clin 11orige to the integrity ;
of a fissfori product barrint. +
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NPE90/SNhtCFI,R - 80 l
i
-, . __.._._ ,--.., _ . _ ,...,. m _ ___.. , _ , - . . , _ _ _ _ - , _ _ _ . _ . _ . ~ , . _ _ . . - _ , , _ . _ _ - _ . . _ - _ _ _ _ . . _ . _ _ _ _ _ _ _ - _ _ - -
Attachment to GNRO-91/00001 I
l SRASN: NPE-90-079 DDC NO: MLp-90-1020-500-R00 SYSTEM: N22 j I
!)ESCRIPTION OF CHANGE: This MCP replaces a leaking elbow downstream of valve N22F098 and relocates a restricting orifice into a straight-section of piping.
REASON FOR CilANGE: The leak in thn nlbow was caused by crosion
, from thn restricting orifice which was located between the cibow
, and valvo N22F098. Moving the restricting orifico into a straight section of pipe will elimiliate the erosion effect on the piping s y s t.cm .
SAFETY EVALUATION: The modifications provido for the repair of a leaking cibow and the relocation of a rostricting orificn to :
c11minatn the existing crosion problem, The piping in supported l
to dead weight loads only, since it in installed in the portion of the Turbino fluilding, containing no safety related equipment. The '
i Condensato Clonnup Syst em sorves no s.1fety funct ion. Systems analysis has shown that f ailurn of t he Coodavisate Cleanup System will not compromise any sa fot y relat ed systems or prevent reactor l shutdown. The operation or funct ion of t he Condensat e Cleanup ,
system, as analyr.ed in the USAR. is not affected by the l modi ficat ions of this MCP. The design change by this MCP is .
non-safety related. The modifications made by this MCP vill not l af fect the N22 syst em. The piping designs havn been designed to ANSI B31.1 code requirements. The system will function in its intended manner, t
( !
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i NPE90/SNLICFLR - 81 ,
L___-__-__._..__- _ _ . . . _ - . _ , , - _ . - - . . _ _ _ - . _- - . - - - - -
SkASN: NPE 90-080 DOC NO: MC P 104 2- S0 0- k00 S Y STI.M : E22 DESCRIPTION OP CllANGF- An annunciator was installed for niternntivn vison] 'ent t on of bot h "llPCS Init int ed" nmi "llPCS liigh Wa t er 1.cVel Se, .In", lhese 111 Pressut o Core Sprny nonuncintors will ut ilize spare cont act s of exist ing Divis ion 'l relays ns sigoni sources. The power through the relay contacts will bn 125 VDC f rom n Division 3A circuit owl will be wired lu o ex ist ing division 3-t o-Nomlivis ion isolators. All wiring on the Q-s hle of t he isolntor will be mntleed as Division 3A when inst alled ami will be routed nrnt sepnrateil the snoe as Division 3 c i rcuit s. The power snuice for the output side of the isolntor amt the annuncintor input is non-divisionnl. This wiring will be designat e i ns non-div is tonal upon innta11ntion and eilI be sepnt at eil f rom a ll div is ionn l c it cu i t s.
REASDN l'OR CllANGP.- The only imlie nt ions avriinhle to identify these two plant comi t t ions ( i . e llPCS i n i t Q t ed , llPCS h i wa t e r level scal-in) were single element inen< ,tescent lamps. Should eit her of these comlit ions have ouurt e<l nmi should t he imlicator lamps be blown, it could not be ens'ly determined if the com!it ion (s) had been r eset .
SAIT.TY EVAL.UATION : This nonvicintor will be u t i l ha.ed a s alternative visual f rullen t i .;n of "lIPCS Init inted" nmi "IIPCS liigh Water 1.evel Senl-In". Oriy existing equipment will be utilize <1 for this dealgn change Only spare c omloc t or s of existing cables et jumper wires n dd er' inside control room panels will be utilized.
The appropriat e di' isions of electrical powrr have been utilized for both the inp'.c and output cf the electrica1 isolntor. The q<lded elec t ric*.1 londs will be int ermit t ent in nature amt t he input to t he divisional side of the isolator is current limited.
Failure of the Division 3A circuit ut i l ized in this design change has boon piovlourly concluded to have no adverse affects or, the sa fet y per f ormance of t he llPCS System.
All wiring will be rout ed , sepa rnt ed , ami ident i f ied in accordance with the npproprinto reg guide. A failure in this pownr circuit can not propngnte in Division 1 or 2 m r can electrical failure in Division 1 or 2 propagate into this power circuit due to this design thqnge.
"his design change will not affect the llPCS System in considointion to items such as flow, chemistry, setpoint, capacity, level, or pressure. Pnilure of the Division 3A circuit utilized for this MCP has been previously concluded to have no adverse a f f ects on t he sa fet y per formance of t he llPCS Syst em.
None of t he n r iected equipment will be requ i re<1 to opernt e outside of theli designed rat ings.
NPE90/ SN1.lCFl.R - 82
A t t a chtnen t t o GNRO-91/00001 SRASN: NPE-90-081 DOC No: MCP-90-1054-S00-R00 SYSTl:H : E12 DESCRIPTION OP CilANC.E: MCP 90-1054 ndds 0-1001 valve _ posit.lon indication in the Control Koom for valve IE12-P424. This volve is the flow control valve for t he Alt eront e Decay Hent Wernovn l Sys t em
( ADilR) . This MCP niso ptovides f or rout ing/ t et minnt ion of 11ainnce of Pinnt (It0P ) cables for future flow monitoring ins t t unient a t ion for ADilk with itutiention iti the Control Roon,.
REASON 10R CHANGl;' To enhnnce the long t e a m v inhilit y of t hn ADilR system, SAlETY EVAL,UATION: Connecting the positinn itulicat ion on t his valve nial rout ing/ t erminnt ion of 140P cables for system flow innlientton do not change the originnt design intent- of nny component, system or st i nct ure. The athlit ion of t he posit ion itulient ion is not requit ed t o suppot t the snfo shntdown of the reactor or to perform in t he opeint ion of react or sn f et y fentures nor is the nthlit ion of flow linlita t ion. The changes ronde by t his MCP do not prevent any e(piipment relied upon to mit ignte the consegnences of any evnlunted transient os accident f t om performing it s sa f et y funct ion.
All structures, systems ntnl component s added or rnod i f led by t his MCP have been designed to meet all applicable requirements and thus no new failure modes are created.
The inntgins of sa fety as defined in t he bases for the Technical Specificatlons are not changed by the addition of this position itulient lon. The addit. ion of t he pos i t inn indient er t o the valve does not cht.nge the originn1 desinn intent of any eqnipmen',
i NPE40/SNI,1Cil,R - 83
i Attachment to GNRO-91/00001 SRASN: NPE-90 082 DOC NO: MCP-90-1064-800-R00 SYSTEH: E12 l i
DESCRIPTION OF CllANGE: HCP 90/1064 adds the equipment necenanty i to provide monitoring capabilities for Altornat.c Decay llent !
Removal (ADilR) syntem flow, pump suction prennure, and pump ,
dischargo pressure. System flow will bo provided in the control !
l room. The control room indication wiil be driven f rom n !
differential prennuro transmitter connected to an annubar flow I acunor. Tha pressure indication will be provided by local ;
instrumentation. All int,trumentation in powered from non i divinfonni Dalanco Of Plant (BOP) pownr.
1 REASON FOR CilANGE: To enhance the long term vinbility of the ADilR !
system.
SAFETY EVAbUATION: The instrumentation installed will provido n l monitoring function only. The instrumentation . installed in a prosaurn boundary npplication in noinmically analyzed for structural integrity and in acceptable for une in anfety reinted pressure boundary. All instrument tubing in installed to sciamic category 1 design requirements. The modifications implemented by I thin MCP will not change any design criterin or functions of tho :
ADilRS. Tnflure of thn components modified or added by this chango '
will not. Initinto or prevent initiation of any at'amic category 1 ,
component, nyritem, or structure. [
The technical specification contains the administrative controls for the operation of the ADitRS. The addition of the flow and pressure indication will not impact the operating controls of tho ADilRS. '
Therefore, the margin of safety as defined in the basis for any technical spncification will remain unchanged, i i
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t-NPE90/SNI,1CFI,R - 84
,,n- . - , . , .- - . - , , - , , - .. , - .- .--~ne-,~.-,.--,-,~.--,,.,..._e, _--_._-,nn.-...-,-- .,--...-..,~,,,.-,n. ~ , , , .,,n-,
Attathment to GNRO 91/00001 SRASN: N i'r.- 9 0- 0 8 3 DOC NO: HCP-40 1055-800-R00 SY STI.M : F.12 Dr.SCRipllON OF Cll ANGF.: MCP 90/1055 adds two manunily operated valves which in effett, relocates the Alt einnte Detny llent Removal system high point vent valves from an elevation of approximately 17' above floor elevat ion t o approximat ely 6" nhove floor 1cvel.
The new valves will be the new sniety to non safety boundary of the vent system. The old valves (i;12r418 and T.121427) will he lef t in pince but will remain open dating system operatton, RF.ASON 00k CilANGI;' To increase accessibility of the manually opernted vent valves, thereby enhnncing ADilRS opernhility nnd increasing pinut worker safety.
S AIT.TY FNA1,11 AT10N : This sn f et y evaluat ion concluded t hat the change did not involve an unroviewed snfety question. The installation of the vent. valves at an accessible locat ion will not affect the operat ion or funct ion of the F.12 syst em as described in the USAR. This design change provides an onmier method of vent ing the system and will not iesult in the crent lon of any new f ailur e mcule s . The piping and pipe support changes will functton in their int enuted manner.
The safety reinted piping and pipe support s derrigns meet ASMr.
Section 111 requirements and are qunlified as seismic category 1.
The non safet) reinted piping and pipe snpports meet ANSI 1431.1 requirements and are qualified as seismic cat egory 11/1. The addition of the pip.ing and pipe supports door not affect the integrity of the int erf acing piping systems or any saf ety system.
The piping and pipe supports will function in their intended mnllfie r .
The installation of the piping nnd pipe supports made by this MCP to the E12 system will not chnogo the functton or operatton as defined by any bases for the Technical Specificat.fon, therefore, the mntgin of safety is not reduced.
NPE90/SNhlCrhR - A'i
Attachment to GNRO-91/00001 SRASN: NPE-90-084 DOC NO: HCP-90-1056-S00-R00 SYSTEM: E12 i
DESCRIPTION OF CIIANGE: HCP 90/1056 Rev. O will allow liquid samplDig capabilities on the RilR and PSW (P44) portions of ADilRS. )
Sarple niement, P44-SE-N093 and samplo element, E12-SE-N195 will i be routed to a new mamplo sink located in the Auxiliary Building. l Elevation 93'-0", Area 10. The E12 sample will require a new penntration AJ-86A, in the north wall of Room 1All6 and a samplo cooler to be utilized. Cooling water to the samplo cooler will be supplied from the CCW system. Thn sample sink drain will bc :
routed to an existing DRW drain.
REASON F)R CHANGE: To make the sampling required by Technical Specifi lons cas ter to obtain.
SAFETY EVALUATION: This design change provides a method of c'>t sining liquid samples of the E12 and P44 portions of ADHRS.
Thit: Sange to the systems (E12 and P44) will not af fect their norn. ,
operation or function, The safety related piping and tubi a designs meet ASME Section III requirements and nrn qualified as Seismic Category 1. Tho non-safety related piping, pipo support and tubing meet ANSI 1131.1 requirements and are 4
qualified as Seismic Category 11/1. The sampin sink support has been dnsigned to withstand tho applicable seismic loads to i preclude any 11/1 hazards. No snismic 11/1 hazards or pipe break concerns will be created by the implementation of this. MCP. The addition of the pipe, pipo supports, tubing, and tubing supports does not affect the integrity of the interfacing systems or any 8afety system. The piping, pipe support s, tubin',and tubing supports will function in their intended manner.
No seismic 11/1 hazards or pipo break concerns will ho created by the implementation of this MCP. No new failure modes are being 1- crented. Thereforn, thern arn no unresolved safety questions associated with this change. *
~ .. bases for Technical Specification 3/4.7.7 is to limit ifro l unmaga by proventing a singin firo from involving morn than one i
safety related firn arna prior to detectinn and extinguishment.
The aforementioned penetration provides a 3-hour fire rated '
closurn wh.ich is an equivalent rating to the affected barriers.
l The implementation of MCp 90/1056 involving E12, P44, and p42 systems will not change the function or operation as defined by any bases for the Technical Specifications, therefore, the margin
- c. safety is not reduced.
NPE90/SNLICFLR - 86 y- w w- -,= -mr,-,- , -w - - - - - ,_r ,,,,-.-,_,--y._.y ,,.,,.-,...-yy,,.,_,,,.-,-.-,,--y--w,,f_,,-.=m..c-a..me,. - . ,-....g. ,.--,--y.. ,=.y---vu, w =w.+P
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Attachment to GNkO-91/00001 SRASN: NPE-90-085 DOC NO: HCP 1061- S00-R 00 SYS'rgH : p44 i I
l DESCRIPTION OF CllANGE: HCP 90/1063 increased the siro of thn I piping immediately downstrenm of Plant Servico Water riow Control !
valve P44F513 from 14" to 24". This valvo is the temperaturo {
control valvo for Lt.n Turbino Du11 ding Cooling Water System ,
(Tf1CW) . l REASON FOR Tilh CllANGE: A pin holo lenk had developed downstronm 4 of the valvn and significant nrosion was discovernd upstream and l downstream of the valvo. This crosion appears to bn a result of l high velocity flow beenuso of thn line sizo reduction. The increase in piping sien from 14" to 24" was made to reduce the i flow vnlocity to an acceptahic level. l i
SAFETY EVALUATION: The mmif ficat ions provide for the repa'r of n lenking reducer and to increase the pipo line sizo to nliminate ,
the existing crosion problem. The Plant Servico Water System ,
serves no safety function. System analysis han shown that failurn "
of thn Plant Service Water System will not compromise any safety reinted systems or prevent reactor shutdown. The operation or ,
funct ton of the Plant Service Water System, as analyzed in the '
FSAR, is not affected by the modification of this HCP. t Tho designs-installed by this HCP meet ANSI B31.1 code '
requirements. The piping is supported to dead weight loads only ;
since it is installed in the portion of the Turbino Building '
cont.nining no safety reinted equipment. Increasing the pipe size '
will not impact operation of the Plant Servien Water System (P44) and will 011minato thn crosion of fect, on t he piping. The system will function in its intended manner.
f l
NPE90/SNLICFLR - 87
, ewww m w e e v e o- my- v w w =-wr-wwm w y im , . -meg--.orww,,.-ww-., 2.r%, - - - e.wr.-.. , -se +=w -+ e-w e e b+
i Attnchment 1o CNRO-91/00001 I SRASN: NPE-90-086 DOC NO: HCP-90-1073-500-R00 SYSTI'H: P44 l DESCRIPTION OF CllANGE: The object ive of this MCP will hn to l removn vnive N1P44F925 nnd replace it with n flanned branch line for hydrolyzing.
REASON FOR CllANGE: Thn fourwny valve, P44F925 on the supply / return plant Service Water (PSW) piping to the Daywell Chillors in obstructing flow. This valvo won originally installed to provido on-lino flow reversal capah!!ition for an automatic tube clenning system on the cold side (PSW) of the Drywell Chillern. Ilowevnt , performance of the clenning system was suspect and the cleaning system was subsequently removed and the valvn was ahniidoned in place.
SAFETY EVAh0ATION: This safety cynluation concluded that the change did not involvo an unreviewed snfety quention. The removal of valve P44P925 will not af fect the operation or function of the syster., sinco thn nutoma,1c t uhn cleaning system for the drywell chillers han been previously deleted and the valvo's flow reversal function in no longer required. The affected system in this HCP '
is non-safety reinted. The failurn of the af fected system will l not comproinine any safety related system or component and will not l prevent reactor shutdown. The mcxlification mndo by this HCP will i not affect the annlysis of the system as doncribed in the PSAR.
The design installed by inis HCP meets ANSI D31.1 Ccxic requirement.a. The piping in supported to dead weight lon'. only, i since it is installed in a portion of the Auxiliary Building whern i no II/I hnzards exist. Removing the valve (N1P44F925) will not ;
impact operation of the Plant Service Water Systra (p44) and will '
ollminata the flow obstructions on thn syntem. ,
i The modificat.fon madn by this HCP to thn P44 System will not ;
change the function of operation as defined in any Itases for the Technical Specifications; therefore, the margin of safety is not reduced.
f b
NPE90/SNhlCFl.R - 88 ,
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Attachment to GNRO-91/00001 ;
SRASN: NPE-90-087 DOC NO: MCP-90-1097-S00-R00 SYSTEM:
DESCRIPTION OF CHANGE: MCP 90/1097 provides for the inspect. ion and repair, as necessary, of pipe supports in SSW "B".
Additlocally, this MCP provides for the removal of non-essential -
basta piping. MCP 90/1097 was being developed to: 1) providn '
inspection requirements and required repair procedures for pipo -
supports in SSW Basin Bt and 2) an an alternate, removn piping and supports in SSW Basis B which do not impact Unit 1 operat.fons.
More specifically, thn piping to be removed is as follows:
1). Portions of the following Unit 2 SSW Basin B piping and associated supports:
- a. Loop C supply froa pump dischargn to basin wa11
- b. l. cop B return from basJn wall to Q2P410014A01
- c. Loop B return from Q2P41G014A01 to cooling tower coll
- d. Q2P410014A01 can only be removed if bo'h part.f als listed as "b" and "c" above are removed.
2). Port.fons of the Unit 2 SSW Basin B small piping, instrumentation and associated non-standard supports.
3). Basin B Sodium llypochlorite and acid piping, supports, and
- spray headers downstream of valvos SP41 AVF505B and l SP41AVF506B.
REASON FOR CHANGEt luspection of thn SSW "A" basin Indicated the potent.fal for corroded pipo hangnrs in "H" SSW hasin.
SAFETY EVALUATION: This safety evalunt.Jon concluded that the l' change did not. involve an unroviewed safety question. The Sodium flypochlorito System is not safnty related and has never been u t.1 i f r.ed . Thn remova1 of components as identifled in MCP 90/1097 will not. compromise any safety related syrnem or componnnts or prevent a safe reactor shutdown.
l The chlorination system (N72) is not safety related and the only safet.y related system which it is connected to is the SSW syst em. !
The design function of the SSW system is not changed by the implementat. ion of this MCP and no new failure modes are created.
The GGNS Unit one Technical Specifications do not mention the Sodium ilypochlorito System and the requirements specified in tha Technical Specifications are not . impacted by the implement.at ion of this MCP.
l NPE90/SNLICFLR - 89 I
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l Attachment to GNRO-91/00001 SRASN: NPE-90-088 DOC NO: MCP-90-1098-500-R00 SYSTEH: E51
- DESCRIPTION OF CilANGE: HCP 90/1098 replaces IE51N052 due to equipment malfunction. Originally, this dnvice was a Rosemount T ll51GP7D52T0003PD transmitter. A Rosemount il51GP7D22T0003PB transmitter is being installed in its place. These transmitters have all the same characteristics except the "D22" has a stainless steel process flango versus a nickel plated carbon steel process flange on the original.
Rosemount 1151GP7D22T003PB transmitters have been qualified for uso inside or outside containment. The 1151 transmitters are commercial grade transmitters purchased by General Electric who dedicated them for nuclear power applications. Qualification of these transmitters was accomplished by testing performed on 1151 transmitters and similarity arguments to 1152 t ransmit'ers.
REASON FOR CHANGE: Tho old transmitter is no longer available.
SAFETY EVAI.UATION: An engineering evaluation was done which concluded that the device cannot fall in such a manner as to negrade the Class lE power sourco. Therefore, those devices can -
be classified as Category C (equipment that will experience environmental conditions of design basis accidents through which it need not function for mitigation of such accidents and whose failure is deemed not. to be detrimental to plant safety or accident mitigation; it need not bo qualified for any accident environment). Further, there are no unresolved safety questions associated with this chango.
Tho engineering evaluation dono indicated that the electrical failure modes and ef fects and concludes that no electrical failure of this device would degrade the Class 1E pow supply.
Therefore, there is no reduction in the margin of safety as defined in the basis for any Tnchnical Specification.
NPE90/SNLICFLR - 90
At.tachment to GNRO-91/00001 SRASN: NPE-90-089 DOC NO: MCP-89-lll2-S00-R00 SYSTEM: P41 DESCHlPTION OF CilANGE: MCP 89/1112 caps the SSW basin overflow drain lines and documents the acceptability of the slight. movement.
of the misalin shinld wall.
REASON FOR CilANGE: The missile shield structures on the SSW pumphouses and valvo rooms had snttled allowing the shield i
structures to separato from the main SSW structuro. This movement damaged the overflow drain lines.
SAFETY EVAhUATION: This safety evaluation concluded that the chango did not involvo an unroviewed safet y quest.f on. The movement of the shiold wall is minor and the shield walls worn designed as separato structures to provido missilo protection for the doors. The angin required for any small missile to enter through the crack la such that the missile would hit thn concreto wall or slab and pose no safety concerns. Since the shield walls have been dnsigned as a separate structure, this movement does not impose any additional loads to the SSW structures. Capping of the overflow line will not, by itsel f, creat n the possibility of an accident since the basin levo) is automatically maintnined, lloweve r , if a malfunction of the basin lovel controller was to occur causing excessive make-up, the basin could overflow. This condition is bounded by the probabin maximum precipitat.lon (PHP) event.
Since shield walls function is maintained, there is no reduction in the margin of safety. Capping of the basins overflow linn will not affect the minimum basin water invol of 130' 3" MSh as required by Technical Specifications becausn the only path new availabic for overflow on thn basins will be at. the basin slab of 133' MSh.
NPE90/SNhlCFhR - 91 l
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Attachment to GNRO-91/00001 SRASN: NPE-90-090 DOC NO: NPEAP-807, 320, 332 SYSTEM:
DESCRIPTION OF CllANGE: Nuclear Plant Engineering Administrative Procedure (NPEAP) 807, 320, and 332 will govern the dispositioning and corrective action via the 10CFR50.59 Safety Evaluntion/ Applicability Screening for all drawing changes mndo in response to a QDR, Drawing Revision Noticos (DRNs) and Drawing Revision Requests (DRRs).
REASON FOR CllANGE: The categories for the drawings changes addressed by this safuty evaluation for thn above documents are as follows:
- 1) Editorial changes
- 2) Device numbers (valves, brenk o rs , penntrations, etc.)
except for those specifically addressed in Technical Spectfications.
- 3) Valvn position Identifiers - except. for those specifically indicated in Technical Specificntions.
- 4) Electrical contact position identifiers - except for those specifically identified in Technical Specifications.
- 5) Increase in level of detail shown on drawings, i.n.,
addition of inst rument root valves of Piping and
'.nstrumentation Dingrams (P&ID[s]).
SAFETY EVAI,UATION: This snfety evaluation concluded that the change did not involve an unreviewed safety question. The actions dascrlhed are drawing changes only and have no physical impact on plant components, structures or systems. These changos have no affect on the operations or functions of plant. f ac i l i t. ies nor its reliability.
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NPE90/SNI,ICFI.R - 92
Attachment to GNRO-91/00001 SRASN: NPE-90-091 DOC NO: NPEFSAR-90-0044 SYSTEM:
DESCRIPTION OF CHANGE: NPEFSAR 90-044 corrects the specified maximum clos. urn tiens in UFSAR tablo 5.2-5. This chango deletes i the specified maximum closure times which are not based on an I analytical limit. Specifically the change to tablo 5.2-5 will bring it into agreement with Table 6.2-44 of the UPSAR and with Tabin 3.6.4-1 of the Technical Specifications. The maximum stroke times for valves without analytical limits arn governed by the ASME Section XI Innervice Testing (IST) program.
REASON FOR CHANGE: UFSAR Table 5.2-5 gives a description of pumps and valves which arn part of the reactor coolant pressuto boundary (RCPB), Maximum closurn t imes are listed for those valvns equipped with motor operators. UFSAR Tablo 6.2-44 gives a description of containment isolation valves and lists thn maximum closure time if based on an analytical limit. Valvo strokn times with no analytical limit are not included in tabin 6.2-44.
Valves which are containment isolation and part of the RCPil are listed in both Tablo 5.2-5 and Tabic 6.2-44. Some discrepancies existed between these tables with regard to the strokn tis es listed. Specifically, the " maximum closurn Lines" fu Table 5.2-5 did not agron with thn " analytical isolation timos" in Table 6.2-44 for the following valvns:
l RilR Shutdown Cooling E12F009 Suction E12F008 l Main Steam Isolation B21F022 B21F028 RWCU G33F001 G33F004 f Tablo 5.2-5 contains two valves which hnvn incorrect maximum closure times listed. Those valves arn not containment isolation valves and are not listed in Table 6.2-44. The subject valves arn:
l RWCU G33F250 G33P251 Tabin 5.2-5 also contains maximum closure timos for certain valves which havn no analytical isolation time. The following valves havn non-analytical closure times listed in Table 5.2-5:
RilR licad Spray E12F023 l E12F394 l Main Steam Drain 021F016 i B21F019 l B21F067
(
RCIC Steam Supply E51F076 RWCU Pump Dischargo 1133F019 li33F020 l
l 1
NPE40/SNLICFnR - 93
,r -- -
---r - - ---.-m,- - - .,.-wma -w-.,w aw,-w.-.m---e s,,,.w-e~--.m- ,,,-p - ~,,---sp ,4y9 --rw-- --
At t achrnent to GNRO-91/00001
. NPE-90-091 Page 2 These valves are olso listed in Table 6.2-44 but are not included in Note (d) as having analytical isolation times. Since the stroke times for those valves do not represent an analytical limit, the stroke times should not bn listed in Table 5.2-5.
SAFETY EVAhUATION: This safety evaluation concluded that the change did not involve an unreviewed safety question. The changes to PSAR Table 5.2-5 will not result in any change to the design or funct ion of the associated valves. The analytical stroke times are revised to reficct the correct values as per UFSAR Table 6.2-44 and TS Tr.ble 3.6.4-1. Non-analytical st roke times which are being deleted from Table 5.2-5 are not used in any accident analyses. No physical rnodi ficat ion to any plant equipment is involved.
The changes to FSAR Table 5.2-5 do not. require any new safety analyses or impact. any existing safety analyses. The analytical stroke times which are a f fected by this change are revised in order to reflect the correct values which are listed in TS Table 3.6.4-1 and FSAR Jable 6.2-44. The non-analjtical stroke times which are being deleted from FSAR Table 5.2-5 are not used as a basis for any safety analysis and are governed by the IST Program.
The ability of the valves to perform their required active function will continue to be verified by testing in accordance with the IST Program. Therefore, this change will not reduce the margin of safety as defined in the basis for any Technical Specification.
NPE90/SNhlCFhR - 94
Al t n chment to GNRO-91/00001 SRASN: NPE-90-092 DOC NO: MNCR-89-00293 SYSTEM: U17 DESCRIPTION OF CilANGE: This sa fety evalunt ion reevaluat ed the environmental conditions for certain post accident monitoring equipment. The equipment consist s of the Eberline AXM-1 accident.
range monit ors and the Air Monit or Corporation ( AMC) redundant stack flow monitors. The equipment was established to be in a mild environment and are therefore exempt from being environmentally qualified per 10CFR50.49/NUREG 0588. The equipment was deleted from the GGNS Environmental Qualification Program on this bnsis. No physical change was made tonny l equipment.
REASON FOR CllANGE: To correct the environmental condition designntton of this equipment and delete the equipment from the GGNS Envit onment a l Quali fication Program.
SAFETY EVALUATION: This safety evaluntIon concluded that the change did not involve nn unreviewed safety question. No change to the pinnt or plant procedures is being made. The equipment.
performs no safety function and is merely being deleted f rom the GGNS EQ Program because the environmental conditions have been determined to meet the definition of " Mild Environment". No chnnge is being made to the equipment and apprcpriate separation from safety systems already exists.
Deleting these items from the GGNS EQ Program will not reduce the margin of safety as defined in the basis for any Technical Specifications since the items perform no safety function, are separated from Class lE power, and are located in mild environments post accident.
NPE90/SNLICFLR - 95 !
l Attachment to GNRO-91/00001 SRASN: NPE-90-093 DOC NO: NPEFSAR 90-0056 SYSTEM:
DESCRIPTION OP CllANGE: This changes UFSAR table 3.2-1 to accurately reflect the as-built configuration of liigh Pressure Corn Spray (llPCS) Diesel Generator auxillaries.
REASON FOR CllANGE: During the licensing of the Division Ill Diesel Generator the NPC established that, as a minimum, all piping and valves in the engine skid mounted portions of the lube oil subsystem, the jacket water subuystem, the starting air subsystem and the fuel oil subsystem which were not dasigned in accordance with ASME Section 111 (i.e. Quality Group C) be upgraded from Quality Group D to Qunlity Group D Augmented. MP6h committed to impose Qunlity Group D Augmented design requirements on the engine skid mounted components (i.e. piping, valves, pumps, etc.) associated with the llPCS Diesel Generator starting air, lube oil and fuel oil subsystems. MP6L also committed to hydrostatically lonk test the engine skid mounted piping in the lube oil, fuel oil and starting air subsystems in accordance with ANSI B31.1 (i.e. 1.5 times the design pressure) even though the NRC required a hydro of only 1.25 times the design pressure which is consistent with Section 111 of the ASME code, in addition MP&L committed to impose the design requirements of ASME Section 111 on: (A) the enginn skid mounted components associated with the llPCS Diesel Gennrator jncket water subsystem and (B) the of f-engine piping and accessories (i.e. exhaust silencers, intake air silencers and intake air filters) in the combustion air intake and exhaust subsystem. Althot.gh MP&L commit ted to design the jacket water and combustion air intake and exhaust subsystems to the Codes applicable to Quality Group C (i.e. ASME Section III, Class 3), the NRC had agreed to accept Quality Group D Augmented (Ref. MAEC 75/36). The design of the piping in jacket water subsystem hns been evnlunted ogninst the guidance of ANSI B31.1 and it has been concluded that the ANSI B31.I requirements have been satisfied. The piping in t he combustion air intake and exhaust subsystem nre designed and installed in accordance with ANSI B31.1 as Seismic Category I piping.
SAFETY EVAhUATION: This safety evaluation concluded that the change did not involve an unreviewed safety question. The engine skid mounted auxiliaries (1,o. piping, valves, pumps, heat exchangers, tanks, etc.) on the HPCS Diesel Generator have been evnlunted/ analyzed ngninst each of the Quality Group D Augmented requirements, as specified in MAEC 75/36, nnd it has been concluded that the design of the subject systems comply with thn intent of the Quality Group D Augmented requirements. Changing UFSAR Table 3.2-1 to accurately reflect the as-built configurations of IIPCS Diesel Generator auxiliaries will not jeopardize the ability of the llPCS piesel Generator to perform its design safety function. Nn new failurn modes have been introduced j since the skid mounted auxiliaries are in compliance with tho ;
Quality Group D Augmented requirements. l l
)
NPE90/SNLICFlR - 96
. ~ _ _ _ . _ _ . _ ~ _ - _ .
Attachment to GNRO-91/00001 ;
NPE-90-093 Page 2 The GGNS Unit One Technical Specifications do not address the quality group classification nor the codos and standards used to design and install the skid mounted nuxillarins on the llPCS Diosol Generator. Therefore, thorn is no reduction in the margin of snfety as defined in the basis for any Technical Specification.
f r
NPE90/SNLICFI,R - 97
Attachment to GNRO-91/00001 SRASN: NPE-90-044 DOC NO: NPE'8AR 90-0021 SYSTEM:
] DESCRIPTION OF CilANGE- UFSAR 6.2.1.1.5.8 addresses a failure in the 1.pCI injection check valve (E12F041 A/B) during t ransfer f rom inject ion into the vessel (hPCI) modo to the containment spray mode. The nnnlysis was conducted to determine the amount of containment pressurization which could occur due to postulated back flow through a failed open hPCI check valve into the containment spray piping. 11ack flow can only occur during the time it takes the 1.PCI injection valve to close and while the containment spray hender valve is opening. Thn nonlysis assumed
?8.5 seconds for the E12F042A/B to close and resulted in n 0.8 psi increase in containment pressure, initial containment pressure was assumed to be 9.0 psig, therefore, the total pressure as n result of the check valve failure is 9.8 psig which is well below the containment design pressure of 15 psig.
An nualysis was done using the same methodology, but assuming a stroke t ime of 30 seconds to ensure that the conclusions reached in the original analysis were still valid.
REASON FOR CilANGE: Valve stroke times associated with the Inservice Testing Program potentially conflicted with times in the UFSAR.
SAFETY EVAh0AT10N: The revision of UFSAR 6.2.1.1.5.8 clarifles that the 18.5 second closure time is an assumption based on GGNS startup data, and specifice that this analysis has been further evalunted up to a stroke time of 30 seconds. The analysis in s ec t. lon 6.2.1.1.5.8 is an evaluation of the f ailure of the LPCI injection check valve (E12F041A/B) during transfer from injection 1,+ n the vessel t o the containment spray mode. An increase in the time qllowed to close E12-F042A/B from 18.5 seconds to 30 seconds vill increase contninment pressure f rom 9.8 psig t o approximately 10.6 psig, which is bounded by the maximum calculated accident pressure of 11.5 psig as listed in UFSAR Table 6.2-13.
The revision to UFSAR 6.2.1.1.5.8 does not reduce the margin of safety ns defined in the basis for any technical specification.
The margin of snfety is established by a more bounding analysis resulting in a higher accident contninment pressure as specified in UFSAR Table 6.2-13. Technical Specifications do not include n time requirement for the close direction for E12-F042A/B since those valves do not recolve nn nut oma t ic cont a inment isolation signal.
NPE90/SNh1CFhR - 98
Attachment. to GNRO-91/00001 SWASN: NPE-90-095 DOC NO: CN-90-0268 SYSTEM:
DESCRIPTION OP CllANGE: CN 90-0268 provides for nn interface device between the Plant Pnging System and the telephone system in the second level of the M&E hullding.
REASON FOR CHANGE: The installation will allow M&E hullding personnel the use of Plant Paging System through their telephones.
SAFETY EVALUATION: This snfety cynluntion concluded that the changn did not involve nu unreviewed safety question. All work associnted with the Plant Paging System and the telephone system nre non-safety related. All installation in M&E Building nre non-seismic since this building does not contain any safety related equipment. Power to t he P. A. system nie from BOP hatteries D & E. The Pinnt P. A. and t elephone systems nre independent and elect rically separated from all other class lE circuits.
NFE90/SNLICFLR - 99
Attachment to GNRO-91/00001 SRASN: NPE-90-096 DOC NO: NFEFSAR 90-0042 SYSTEM:
DESCRIPTION OF CilANGE: The stroke times for the motor operated valves in the Main Steam Isolation Valve Leakage Control System (MSIV-LCS) are given in the UFSAR section 6.7.1.3.1 as about 5 seconds. The actual stroke times based on operating history are between 7 and 10 seconds. A revision of the UFSAR is required to correct this discrepancy. The specific valves covered by these requirements are:
E32-F001A, E, J, N E32-F002A, E, J. N E32-F003A, E, J, N E32-F006 E32-F007 E32-F008 E32-F009 REASON FOR CHANGE: The maximum stroke times of 15 and 30 seconds are based on a system process limit which will cause the inboard system to trip if adequate flow is not established within 3015 seconds. The outboard system has no low flow trips associated with it's control circuitry. Therefore, a maximum valve open stroke time of 15 seconds will allow flow to develop in the system l before the minimum trip setpoint of 25 seconds is reached. The 5 second stroke time currently in subsection 6.7.1.3.1 of the UFSAR has no analytical basis. Therefore, revision of this subsection is necessary to eliminate this incorrect valve stroke time.
SAFETY EVALUATION: This change revises the MSIV-LCS motor operated valve stroke times from about 5 secords to 15 to 30 seconds. The correct valve stroke times of 15 to 30 seconds have always been specified in the MSIV-LCS Design Criteria and General Electric Frocess Diagram. For this reason, this UFSAR change request nnly corrects an error in the UFSAR and does not change l any operational parameters or design requirements of the MSIV-LCS.
This change does not introduce any new operational parameters or design requirements to the MSIV-LCS or any other system. No change to any physical system will be made.
Actuation of this system will be by operator action no sooner than 20 minutes following a postulated design basis LOCA. In addition, Table 3.6.4-1 of the Technical Specifications has not included any maximum valve isolation time for any of the motor operated valves in the MSIV-LCS. This change t.o subsection 6.7.1.3.1 of the UFSAR also will not result in any change to the required valve stroke l times specified in the MSIV-LCS Design Criteria and General Electric Process Diagram. For these reasons, this change will not l
reduce the margin of safety as defined in t he basis for any technical specification.
I l NPE90/SNLICFLR - 100 1
1
Attachment to CNRO-91/00001 SRASN: NPE-90-097 DOC NO: EER-90-6388 SYSTEM: B21 DESCRIPTION OF CllANGE: Engineering Evaluntion Request (EER) 90-6388 requested that temporary lead shielding he attached to certain portions of the packing lenk-of f lines from tt 3 B21F028 valves going into the standpipe drain which runs into the Reactor Core isolation Cooling (RCIC) room. Calculations were performed on the subject piping with the added weight of the lead shielding.
These cniculations show that the structural integrity of t he subject piping with the temporary lead shielding will he maintained in the unlikely event of an operating basis carthquake (OBE) or a safe shutdown earthquake (SSE). All applienble ANSI code stress allownbles are met. Therefore, the operability of t he system in Operating Modes 4 and 5 is not a f f ected by the to iporary lend shielding attached to the pipe.
Based on the above analysis, the temporary lead shiniding was installed on the pipe during Oparating Modes 4 and 5. No other lead shielding or any other additional weight could be attached to the piping out. to the first anchor while this shleiding was attached. This temporary shielding was installed during Operating Modes 4 and 5 only, and was removed prior to restart a f ter RF04.
Tempornry addition of lead shielding does not result in any permanent changes to locatlon, routing, or type of supports, nor does it alter any component performance characteristics, design parameters, or operational parameters of the a f fected system af t er thn temporary lead shielding is removed.
REASON FOR CilANCE: To reducn radiation exposure to personnel performing work in this area. The lead shielding will be installed during Operating Modes 4 and 5 only, and must be removed prior to restart.
SAFETY EVAhUATION: This sa fety evaluat ion concluded thnt the change did not involve an unreviewed safety question. Since these temporary changes do not affect the structural intagrity of the subject piping during cold shutdown, since all applienble ANSI code allownble stresses are met, the probnbility of occurrence of an nccident resulting from a seismically initiated pipe break is not increased. There will be no chnnge to existing designs after the lend shielding is r e.. ov ed . No new fr l!"re modes a re created.
Structural integrity of the subject piping has been confirmed with temporary lend shielding for Operating Modes 4 and 5.
Installation of lead shielding temporarily does not change the limiting conditions for opnration, applicability, or surveillance requiremants as defined in tle hasis for the Technien1 Specifications. There wi11 he no permanent changes made to ex isting des igns or opnrat iona l pa rameters a f ter t he affected shielding is removed.
NPE90/SNhlCFLR - 101
l ALtachment to GNRO-91/00001 l
SRASN: NPE-n0-098 DOC NO: Engineering Report SYSTEM:
GGNS-90-0028 R00 DESCRIPTION OF CilANGE: Engineering Report GGNS-90-0028 R00 ovaluated upper conta ament pool singic failure and siph,n protection requirements. This report determined that the siphon protection vacuum breakers 'G41-F042A through 11 and G41-F060A through D) in the Fuel Pool Cooling & Cleanup (FPCCU) System return lines to the upper containment pool (UPC) need not he classified as active safet.y-related components.
REASON FOR CilANGE: Reclassification of the vacuum breakers allows safe climination of ASME Section XI testing requirements in association with overall efforts to replace the FPCCU system siphon breakers with a more reliable, passive form of protection.
The existing design meets the intent of the siphou protection requironients specified in GE and GGNS design document s. These requirements do not specify a degree of protection which will prevent any drop in UCP water level nor are they intended to maintain the levels above the T/S minimum limits following single failures. All postulated single failures resulting in a UCP draindown below specified minimum levels have been evaluated to be acceptable in that the capabilities of the plant systems to perform and maintain a safe reactor shutdown or mitigate an accident are not reduced. Therefore, the existing piping design meets the applicable rnquirements and the active function of the UCP siphon breakers is nonsafety-related. Although the design requirements for the act ive function of these components are being changed, implementation of these changes do not constitute a design change as defined by GGNS procedures.
SAFETY EVAL.UATION: This safnty evaluation concluded that the change did not involve an unreviewed safet.y quest f ou. The Engineering Report. evaluated thn following types of events which could lead t.o an UCP draindown: 1) actuation of the SPMU system during a LOCA; 2) an inadvertent UCP dump; 3) a moderate energy line crack in piping connected to the UCP; 4) siphoning of water from the UCP; and 5) an operator error which results in loss of water through piping connncted to the UCP. Each of these potential causes of a draindown was considered as a single initiating event and was thoroughly evaluated agajnst the UCP design criterin. The report concluded that there will be no impact on the theoretical r inimum UCP wat er levnt i f no credit is taken for these siphon breakers. The only scenario which would result in lower water levels is that of an operator error which results in thn isolation of the fuel st.orage area dif fuser line by closing valve G41-F254 The theoretical minimum UCP water level following this event is at El. 195'-8" which is 4'-4" lower than the minimum theoretical water level if crndit is taken for the siphon breakers (i.e.. El. 200'-0"). The probability of this operator error is extremely remote and need not be postulated to occur coincident with a passive piping failure since there is no l
l NPE90/SNI.lCFhR - 102
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Attachment to GNRO-91/00001 '
hPE-90-098 '
Page 2 mechanistic relationship between these two failures. Although certain failures resulted in UCP water levels below the T/S minimum requirements as previously described, the consequences were evaluated to be within the existing licensing bases for all reactor OCs. For a "DDA" 1.0CA concurrent with the design basis assumptions including a loss-of-offsite powor and a single limiting failure, the climination of the vacuum breaknra would have no impact on the capabilities of the supprossion pool Jakeup system in performing the required safety functions. In addition, the evaluation concluded that thorn are no UCP draindown event.s involving the linos containing these siphon breakers which may occur concurrent wit h a funi handling accident which would prnvent the UCP from performing the required fission product removal functions within the applicable timo framo as currently analyzed.
The only activo function of those siphon breakers is to limit the '
severity of an inadvertent UCP draindown. The reclassification of this function as not safety-related by tSn Engineering Report has no effect on the probability of a draindown event. The passive safety function of these siphon breakers for maintaining t.he associated safety-related pressure boundaries is not changed by this reclassification of the act. lyn vacuum relief function. Thus, the probability of a passive siphon breaker failuro which could lead to an inadvertent draindown event is not increased. All other applicable design requirements are not changed by this report. The results of this report and the changing of the active safety function rnquirements for these siphon breakers will not cause any system or component to operate beyond its design limits nor will it affect overall system performance in a manner which could lead to an accider.t. No accident precursors evaluated in thn UFSAR are affected by this chango.
The design requiroments for thn passive function of thesn siphon i breakers are not changnd by this report. The report results support thn elimination of ASME Sect.fon XI testing requirements and thn eventual removal of the valves by establishing that the activo function to prevent siphoning of the UCP is not sa f e ty- re l a t ed . As evalunted in the re}. ort , all appilcable design, analysis, and installation requiremnnts are mot and that no new equipment failure modes are introduced by the elimination I
of the active function of thnso siphon breakers. The changns in thn class;fication and testing requirements for the UCP siphon brnakers do not a f fect any existing bases for the Technical Specifications and do not introduce any new requirements. By the evaluation presented in t.his Engineering Report , all applicable requirements for the existing UCP water levnt specifications are mot. The margin of safety provided by the minimum UCP water l Invels spncified in the Technical Specifications arn applicabin to e a 1.0CA and a fuel handling accident. No siphoning event l postulated to occur following a I.0CA or a funi handling accident i
s NPE90/SNLICFLR - 103
. . . . _ _ _ . _ . . _ . . _ _ _ _ _ _ _ . . _ . - .___ . _ _ . _ . _ . - - _ . . _ . . _ . . . _ . _ _ _ , _ . _ _ . - _ _ _ _ _ _ . . _ _ _ _ _ . _ _ _ _ _ ~ ~ . _ _ . . _ _ . . .
Attachment to GNRO-91/00001 i
NPE-90-98 Pago 3 would result in any significant reduction in (JCP water inventory during the period when this water level is required to achiovo safe shutdown or to limit. the relonso of radioactivity.
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NPE90/SNLICFLR - 104
Attachment to GNRO-91/00001 SRASN: NpE-90-099 DOC NO: EER-90-6231 SYSTEM: G33 1
l DESCRIPTION OF CllANGE: EER-90-6231 ovaluated tho addition of I temporary lead shielding to certain portions of the Reactor Watnr Cleanup (RWCU) system. The 1 cad shiniding was installed only during Operating Modes 4 and 5. Calculations were performed on tho subject piping with the added weight of the load shielding.
These calculations show that the structural integrity of the subject RWCU piping with the temporary lead shiciding and supports will be maintained in the unlikely event of an operating basis carthquako (013E) or a safe shutdown carthquake (SSE). All applicabin ASME code stress allowables are met. Therefore, the operability of the RWCU system in Operating Modos 4 and 5 is not af fected by the teniporary lead shielding attached to the pipe.
Temporary dead weight supports worn installed on the system before the lead shiolding was added and was not removed until all the shielding was removed. During the timo the temporary support s are being utilized, the changn in temperature of the RWCU system was not to exceed 50'F. Also, no other load shielding or any other additional weight can be attached to the piping out to the first anchor while this shiolding'is attached. The temporary lend shielding and supports woro installed during Operating Moden 4 and 5 only and were removed prior to restart. a f tnr RF04.
REASON FOR CilANGE: To reduce radiation exposurn to personnel performing work in this area.
SAFETY EVALUATION: This safety evaluation concluded that the change did not involvo an unreviewed safety questlon. Structural integrity of the RWCU piping has been confirmed with temporary lead shleiding and supports for Operating Modes 4 and 5. Thorn are no permanent. changns made to existing designs af ter the af fncted shielding and temporary supports are removed.
Since all applicable ASME code allowable stresses are mot, the probability of occurrence of an accident resulting from a seismically initiated pipe break is not increased. No new failure ,
modos are created.
Installation of lead shielding temporarily does not change the limiting conditions for operat ion, applicability, or surveillance requirements. Therefore. thorn is no reduction in the margin of safnty as defined in the basis for any Technical Specification.
NPE90/SNhlCFhk - 105
Attachment to GNRO-91/00001 I
I SRASN: NPE-90-100 DOC NO: MNCR-90-0176 SYSTEM: !
l DESCRIPTION OF CHANGE: MCP 90/1095, Rev. O was issued to repair damago reco1ved by tho shroud head in RF04. In addition, a previously issued design, MCP 90/1090, Rev. O was used to replace the locking bolt at location 34 due to spline wear. This safety evaluation addresses the modifications made in the above design documento. Also, this safety evaluation addresses those areas of dnmago where it was determined that the as found condition was a ccep t.a b l e. The repairs included removal of a damaged separator assembly, removal of existing shroud head bolt locking assembiles at bolt locations 12 through 28, installation of new bolt assemblies at locations 12, 14, 16, 19, 21, 23, 25, 27, 34, and a j wold repair to a gusset .in the vicinity of bolt 14..
An engineering ovaluation was performed to evaluate the damage and repairs performed on the separator. A summary of the evalur. Lion resulte la provided below:
The upper guido ring need not be restored to its original condition. The function of the ring is to provido alignment and support for the locking bolts. The design loads for the guido ring are small and well below the capability of the ring. The structural integrity of the ring is maint ained in the bent position. The extension bolts and the retainer cans lu the damaged areas are being removed and repinced, as required, to meet minimum bolting requirements.
The fanction of the tio bars is to interconnect the separators in order to reduce flow induced vibration, and to provide support against horizontal loads during a acismic event. The tio bars can i
still perform this function in the deformed condition. The i structural adequacy of the tio bars is maintained.
A weld repair w1s performed on the gusset torn from the separator.
Gussets that pushed into separator tubos are acceptable.in that. l position. The dimpling is minor and does not adversely affect the l structural integrity of the gusset a* separator and does not adversely affect the performance of .ie shroud head / separator. '
The locking bolts that were bent in the damaged area were removed t.o facilitate stud detensioning.
The retainer cans that woro damaged will be removed and replaced, as required. The associated bolts will also be removed as stated above. The retainer cans perform no function if the bolt is '
removed.
The elevated separator assembly will be removed. Thnro are 301 separators on the shroud head with the requirement to have only 280 to ensure adequato separator performance. Thorofore, removal of the separator assembly will have no adverso af fect on the shroud head.
NPE90/SNb1CFLR - 106
Attachment to GNRO-91/00001 NpE-90-100 page 2 REASON FOR CllANGE: The GGNS Unit I reactor steam separator roccived damngo during RF04 vessel disassembly. The damage occurred when the upper guido ring was contacted during removal of the dryer f roin the vessel . The damago wns confined to the nron from approximately Azimuth 110' to azimuth 200'. A summaty of the damage is as follows:
The upper guido ring was bent vertically upward n maximum of approximately 30*
Several tio hats were slightly buckled Several gussets used to attach the guido ring to thn separator had pushed into tbn separctor tubes Soveral locking bolts were hent Snverni retainer can worn part.lally detached or hent Ono_ separator was elevated approximately 1.5" higher than the others The reactor shroud hond consists of a flango and a domo onto which is welded an array of st.andpipes, with a steam separator on top of each standpipe. The shroud head mounts on the flango at the top of tho top guido and forms the cover of the core dischargn plenum region. The stainless stool fixed axial flow type steam sepnrators havn no moving parts. The shroud head is bolted to the top guido flangn by shroud head studs that have an extension to the top of t"ie separators for access during refueling. The separator /shtoud hond is not a pressurn retaining component. It is nonsafety-related, safety class other, and nonscismic.
SAFETY EVAL,UATION: This snfnty evaluation concluded that the chango did not involve nn unreviewed safety question. The damage that is "necept as is" and the repaired damage does not advnrsely affect the structural integrity or performance, nor does it create i the potential for a lonso part. The bolting requirements for tho shroud head are maintained within the design limits. The actions l
l described will not cause n decrenso in reactor coolant temperature, an increase in reactor pressure or a decreano in reactor coolant system flow rate. The actions will have no af fect on reactivity or power distribution, in addillon, the act.lons will not enuse on increase or decrease in ronctor coolnnt-invent ory, a f fect. the radioactive reinaso from a subsystem and component, or affect the control rods from performing thnir ,
function.
The function and structural Integrity of thn separator / shroud hond is maintained. Thn separator / shroud head does not serve a safety function nor will the a .tions described adversely affect any safety related systems cr components, or prevent safn shutdown.
NPE90/SN1ICFhR - 107
Attacheint to GNRO-91/00001 SRASN: NPE-90-101 DOC NO: EER-90-6385 SYSTEM: F41 DESCRIPTION OF CHANGE: EER-90-6385 evaluated the possibility of deferring the removal of certain reactor internal vibration instrumentation until RF05.
The UPSAR lists the equipment used in the Reactor Internal Vibration Monitoring System (F41), along with the location of equipment insido the reactor vessel. A partial description of this startup test equipment is included in GE Specification 21A3854, which states "it is intended that the equipment above the shroud support plato and above the corn support plato be removed during the first refueling outage". Most of the incoro vibration instrumentation was removnd during RF01, RF02, and RF03. The vibration instrumentation remaining in vessel at the start of RF04 is listed bolow:
Group 1 Guide rod with associated vibration instrumentation string Group 2 Four (4) transition blocks as foll,ws:
I at 90 associated with Jet Pump 6 1 at 150 associated with Jet Pump 12 1 at 200 assoc. fated with Jet Pump 14 1 at. 270 associated with Jet Pump 19 Group 3 Vibration equipment as follows: (90 to 180')
Fourteen (14) clamps One (1) coupling Seven (7) conduits approximately 14 feet Jong with lead offs Group 4 Vibration equipment as follows: (180* to 270')
Fourteen (14) clamps One (1) coupling Four (4) conduits approximately 14 feet long with lead offs Group 5 Vibrolon equipment as follows: (Bottom grid to top of shroud support plate at 180')
Ten (10 clamps One (1) coupling Four (4) conduits approximately 20 feet long with Icad offs NPE90/SNLICFLR - 108
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Attachment to GNRO-91/00001 NpE-90-101 Pago 2 A review has been performed to allow vibration instrumentation Groups 2-5 (or any combination of Groups 2-5) to remain within the reactor vessel until RF05. The review concluded that this deferral is acceptable. Thn basin for the acceptance is thn results of a vibration instrumentation residence timo evaluation which concluded that tho degradation of the vibration instrumentation equipment would be unlikely for up to 130 months of operation. >
REASON FOR CilANGE: To reduce the impact of reactor vessel vibration instrumentation removal on the RF04 schedule.
SAFETY EVAhUATION: This safety evaluation concluded that the chango did not involvo an unroviewed safety question. The presence of the subject equipment within the reactor vossal in Cycle 5 will not havn any af fect on the response of the plant to any of the analyzed accidents. Thoro is no credible mec.hanism to force any of the subject parts off their mountings. It was shown that the only conceivable mechanism for detachment of this '
equipment (stress corrosion cracking) is not a credible event during Cycle 5.
Because the equipment coming loose and circulating in the reactor vessel has bec.n evaluated not to be a credible event, there is no concern for interference with control rod operation or fuel performance. Reactor coolant chemistry will not be affected by ,
this equipment dun to the use of stainless steels which are suitable for use inside thn reactor vessel. The subject equipment no longer serves any funct ion. Furthormorn, ovaluations have shown that the structural integrity of the equipment will be maintained for at least another cycle of operation ensuring that no safety related systems or components will be affected.
Thernfore, the actions described will not reduce the margin of safety as defined in the basis for any Technical Specification, i
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Attachment to GNRO-91/00001 SRASN: NPE-90-102 DOC NO: MNCR-90-0093 SYSTEM:
DESCRIPTION OF CilANGE: The dual coil solenoid valves for the main steam isolation valves (MSIV) have been replaced or rebuilt as an interim measure until the valve design can be modified in RF04 Thn only tebuild kits available are equipped with Viton seating material. Viton is an acceptable material for the period in question and is used as the material of choice by scveral utilities. GGNS is not currently using the Viton material in the drywell due to its radiation tolerance performance. An evaluation has been performed to ensure acceptable performance of the Viton material until RF04. Both Viton and the EPDM material have similar thermal aging performance.
REASON FOR CilaNGE: The ASCO PfX- series single and dual coil solenoid valves for thn main ..ean isolation valves (B21) were replaced with ASCO NP- series solenoid valves by DCP 84/3084. The llTX- series were not environmentally qualified valves while thn NP- series are an environmentally qualified valve. Both valves are functionally similar. liNCR 265-89 later reworked the i nboa rd and one outboard MSIV to replace thn internal clastomer. The EPDM elastomer was deteriorating at a faster rate than previously expected. MNCR 0093-90 documents another case where the EPDM elastomer has deteriorated at a faster rate than expected.
SAFETY EVAhUATION: This safety evaluation concluded that t he change did not involve an unrnviewed safety gunstion. The implementation of this MNCR will not increase the probability of occurrence of an accident. The subject valves are funct.iona l ly similar to thosn they replaced. The NP- series valves are environmentally qualified per 10CFR50.49. Viton is thermally equivalent to EPDM (EQDP EQ6.3, Tah ill), however, Viton is more radintion sensit ive. The use of Viton unti1 RF04 has beer evaluated and its radiation threshold is acceptable for greater than one year of service wh!ch wilI not be exceeded prior to replacement in RF04. The clastomer matnrials can bn considered equivalent materials for the perio<1 of time they will be installed.
Because of the functional similarit y of the replacement ASCO solenoid valves to the ones that were replaced, no change in Plant Technicnl Specifications are required. Replacemnnt of the internal elastomer within the solenoid valves wita Viton will not impact the valve function. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specificatfon.
NPE90/SNhlCFI.R - 110
Attachment to GNRO-91/00001 SRASN: NPE-90-103 DOC NO: EER-90-6401 SYSTEM: G36 DESCRIPTION OF C!!ANGE: EER90-6401 evaluated the addition of temporary lead shielding to certain portions of the Reactor Water Cleanup (RWCU) drain lines from the regenerative heat exchangers.
The lead shielding will be installed during operating modes 4 and 5 only and must on removed prior to restart.
REASON FOR CilANGE: To reduce radiation exposure to personnel performing work in these arcos.
SAFETY EVAhUATION: Thin sar c ty evaluation concluded that the change did not involve an unroviewed safety question. These temporary changes do not af fect the structural integrity of the subject piping during cold shutdown. There will be no change to existing designs after the lead shielding is removed. Since cl1 applicable ANSI Ccxle allowable st resses are met , the probability of occurrence of an accident resulting from a seismically initiated pipe break is not increased. No new failure modes arn created.
Structural integrity of the subject piping has been confirmed with temporary lead shielding for Operating Modes 4 and 5.
Installation of lead shielding temporarily does not changn the limiting conditions for operation, applicability, or surveillance requ trements as defined in t he b mis for any Technical Specification. There will be no permanent changes made to existing designs or operational parameters af ter the af fected shielding is removed.
NPE90/SNhlCFhR - 111
Attachment to GNRO-91/00001 1
SRASN: NPE'90-104 DOC Not EER-90-6417 SYSTEM:
DESCRIPTION OF CilANGE: EER-90-6417 request that temporary lead shielding be attached to certain portions of the RWCU system. The 8 lead shiolding will bn installed during Operating Modes 4 and 5 only, and must be removed prior to restart. Reactor pressurn cannot be increased above 280 pounds while shielding is installed.
This evalunt.fon does not cover reactor hydrolyzing.
REASON FOR CilANGE: l' reduco radiation exposure to personnel performing work in this area.
SAFETY EVALUATION: This safety evaluation concluded that the ;
chango did not involve an unroviewed safety question. These temporary changes do not affect the structural integrity of the RWCU piping during cold shutdown. Structural integrity of the RWCU piping has been confirmed with temporary inad shiolding for Operating Modns 4 and 5. Thorn are no permanent changes madn to existing designs after the a f fected shiciding is removed.
Calculations worn performed on the subject piping with the added wofght of the lead shielding. These calculations show that the strut.iural integrity of the subject RWCU piping with the temporary shiniding will be maintained in thn unliknly event of an operating basis carthquake (011E) or a safe shutdown eart.hquako (SSE). All applicable ACHE code stress allowables are mot. Inadvertent pressurization. dun to loss of shutdown cooling (SDC) in Modo 4 was considered, llawever, duo to the nature of the errors of failure required to cause the event, pipo breaks arn not required to be i analyzed.
Structural Integrity of the RWCU piping har heen confirmed with temporary lead shiniding for Operating M, des 4 and 5.
Installation of lead shiciding temporarily does not chan g the limiting condit ions for operation, applicability, or surveillanco requirements as defined in the basis for thn Technical Specifications.
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NPE90/SNhTCFhR - 112
Attachment to GNRO-91/00001 SRASN: NpE-90-105 DOC NO: GN-90-0523 SYSTEM:
DESCRIPTION OF CllANGE: Above normal seat leakage was identified for the Automatic Depressurization System ( ADS) air accumulator stop check valves during leakage testing for valves. (Q1B21F036D, F036F, F03611, F036J, F036p, F036R through F036T, F036U, F039D, F039F, F03911, F039J, F039P, and F039R through F039T) for the . W air accumulators. Additionally, a nonconforming condition fo. .he ADS air supply was identified. The old piping analysis assumed a peak piping temperature of 240'F. Since ADS is required for accident conditions the piping most be analyzed for the peak post accident drywell temperature of 330*F. The evaluated document provides approval and justification for closing one of the accumulator stop check valves and replacement of the remaining accumulator stop check valve with a resilient sent check valve.
In addit ion, this document provides for the necessary piping support modifications to qualify the ADS air supply piping for 330'F.
REASON FOR CilANGE: Closing of one of the inlet stop check valves for each ADS S/RV will not prevent the accumulators from initially charging or prevent S/RV leakage makeup following actuntion since a common two inch discharge line connects both accumulators to the S/KV actuator. Furthermore, there would be negligible pressure drop across the remaining stop check valve in the common one inch supply to the two accumulators considering the leakage makeup requirement of I scfh. Finnily, closure of one of the inlet stop check valves for each ADS S/RV directly prevents accumulator depressurization through that. valve upon depressurization of t.he common distribution header. Replacement of the remaining ADS S/RV accumuintor stop check valves with a resilient seat check valve improves the seating characteristics of the valves. The seating surface of the existing stop check valves is metal-to-metal which results in above normal seat leakage during required surveillance testing for the ADS S/RV accumulators. The resilient sent check valves men'. all applicable code and design requirements, including environmental considerations. The current piping analysis shows that the piping was analyzed for a temperature of 240 F. Since ADS is required for accident conditions the piping must be analyzed for the peak post accident drywel! temperature condition of 330 P.
SAFETY EVALUATION: This safety evaluation concluded that the change did not involve an unreviewed safety quest.fon. The UFSAR cons ide rn various accidents and transients which are postulated to occur in order to determine the capability of the plant to operate within regulatory guidelines without undue risk to the public health and safety. Those accidents and transients whose probability of occurrence may he lucreased due to closing one accumult, tor stop check va lve and replacing the remaining stop check valve with a resilient seat check valve involve only those accidents which are dependent on the ability of the passive air supply system to support the ADS, Low how Set (LLS) and relief NPE90/SN..ICFhR - 113
Attachment to GNRO-91/00001 i NpE-90-105 page 2 functions. For the previously postulated accidents and transients dependent on the ADS, LLS, and relief functions, the required safety functions of the stop check valves will be maintained; therefore, the passive air supply system supports the ADS, LLS and relief functions. In addition, the ADS air supply piping system and pipe supports designs meet ASME Section III requirements for the required accident and transient scenarios and nre qualified as seismic category 1. The piping and pipe supports will function in their intended manner. The proposed changes do not adversely affect any fission product barrier, the ability to mitigate accidents and transic.nts, or the radiological consequences of -
accidents and transients.
The ADS air supply piping system and pipe supports designs meet ASME Section 111 requirements for the required accident and transient scenarios and are qualified as seismic category 1. The ADS, LLS, and relief functions are no more likely to fall when required to function than before.
The ADS and non-ADS air accumulator stop check valves are not explicitly discussed in the bases for TS 3/4.5.1. The bases assume the operability of the passive air supply system to ensure
. that the ADS function to depressurize the reactor vessel so that the low pressure ECCS can inject water into the reactor vessel for core cooling following a small primary system line break if the llPCS system in11s or cannot ceintain reactor water level. The margin of safety associated with the ADS function involves the ability to dercessurize the reactor to prevent exceeding fuel cladding integrity limits. As discussed above, operation with the proposed modifications has been evaluated for its ef fect on the ADS function during postulated accidents. Evaluation results demonstrate that the passive air supply system supports the ADS function with no impact on fuel cladding integrity limits.
The margin of safety associated with the LLS function involves the ability to minimize the induced loading on the containment /
suppression pool bor..idary by ennuring no more than one relief valve opens subsequent to the initial blowdown on an overpressure transient. As previously described, the proposed changes have been evaluated for their effect on the LLS function dur.ing l postulated transients. Review results demonstrate that the passive air supply systnm supports the LLS function with no impact on the ability to prevent more than one relief valve f rom opening subsequent to the initial blowdown on an overpressure transient.
The margin of safety associated with the relief function involves the ability to protect the reactor vessel from overpressure during l upset conditions. As previously dnscribed, the proposed changes i have been evaluated for their ef fect on the relief function during postulated overpressure transients. Review results demonstrate NPE90/SNLICFLR - 114
~_ _ _ _ . ,,,_ -__ _ _ _ . . . _ _ _ _ . _ , . _ _ _ _ . _ _ . . _ . - _,
. _ _ _ _ _ , . _ . __ _ _ ... . - _ _ .. _ ._._ .._ _ ._ _ _ __ . . . _ . _ _ _ _ _ _ - , . ~ . . . -.
Attachment to GNRO-91/00001 [
1 NPE-90-105 Page 3 that the passive air supply system supports the relief function with no impact on reactor coolant pressure boundary safety limits.
Since operation with the proposed changes has been found to bn acceptablo, the passivo air supply system is capable of supporting the ADS. LLS, and roller functions and the margin of safety as defined in the basis ,r the Technical Specifications is not reduced, i
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Attachment to GNKO-91/00001 l
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l SYSTEM:
SKASN: NPF-90-106 DOC NO: CN 90-0537 DESCRIPTION OF CllANGE: CN 90-0537 requested the removal of pressure regulator SP21f438.
KEASON FOR CilANGE:
To reduce the pressure drop in thn Make-up Water Treatment (MWT) system supply to the Circulating Water (CW) pump lube water pumps and enotor coolers.
SAFETY EVAhDATION: This safety evaluation concluded that the change did not involve nn unreviewed safety question. The removal of the regulator will not adverselyControl the function of the affect of lube / cooling water supply system. which system is the supply for the lube / cooling water may he obtained by throttling other valves in appropriato lines or by ad,)ustment of the pressure regulator in the DW supply piping. The change does not compromise any safety rel.ited syst em or prevent a safe shutdown of the plant..
It has no effect on the function or reliability of any equipment important to safety. The design change does not create any interface with equipment important to safety.
No credit is assumed for the DW and CW systems in the Itases of the Technical Specifications. The design change does not affect that part of the MWT system which is addressed in the Technical of containment Specifications, specifically, valves forming a part boundary. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
NPE90/SNhlCFI.R - 116
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Attachment to GNRO 91/00001 !
SRASN: NPE-90 107 DOC NO: ops w/o purge flow To SYS'I EH : l Reactor Recire pump l! DESCRIPTION Or C"ANGE: This Safety Evaluatfon discusses the !
1mpilcations relativo to opea stion with zero sent purgo flow to thn renctor recirculation pump shaft seal assemblies. l REASON FOR CHANGEt Operation with Znro seal purgo flow to reactor recirculat!on pump shaft seal ansemblics will reduce this souren '
of cycIlc thermal st ress responsible for crack init iat ion in the i 4 shaft and heat exchanger.
SAFETY EVAL,UATION: This safety evaluntion concluded that the f chango did not involvo an unreviewed v.ofety question. The UTSAL d
considero various accidents which are postuinted to occur in order
- to determino the capability of the plant to opernto within
. regulatory guidelines without undue risk to the public health and .
i safety. Those accidents whose probability of occurrence may be increased due to operation with zero seal purgo flow involve only those acciden' which are dependent on thn passivo pressure l
4 boundary of the recirculation system. Operation with zero seal 1
purgo supports the passivo pressurn boundary nines cyclic thermni >
stresses will be reduced. Furthermore, thorn are no events postulated in the trSAR directly caused by a reduction in the seal purgo flow and operation with zero seal purgo flow would not i
crnate such an avent. Therefore, since tbo recirculation system l passiva pressur e boundary is not af fected la a manner that could j i lead to an accident or cause an accident previously evaluated to :
shif t to a higher f acquency cat egory, thorn is no increase in the probability of occurrence or in the consequences of an accident or malfunc(fon of equipment important to safety previously evaluated >
in the Safety Analysis Report. Fnrthermore, operat ion with zero *
, seal purgo flow will not provent the recirculation system from performing 4ts design funct lons consistent with the assumptions of the UFSAR accident and transient analyses.
Since operation with zero seal purgo flow supports the passivo pressure boundary as originally designed and since the reactor recirculation sistem is no more likely to fall when required to !
function thnu before, there is nn creation of a possibility for an '
accident or malfunction of a dif f erent type than any evaluated ;
previously in thn Safety Analysis Keport. ,
l Since the seal ;"tge flow is not explicitly discussed in the bases for T/S 3/4.4.1 and since operation with zero seal purge "ow is found to bn acceptab!n for the UFSAR accident and transitat analyses the margin of safety as defirmd in the basis for any ,
Technical Specifications is not reduced. ,
l NPE90/SNLICFLR - 117
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At t achtnent to GNRO-91/00001 hRAsN: NPF.-90-10M IOC NO: 1.I'.R - 9 0 - 6 4 6 6 SY S*1 Dl :
Dl:SCRIPTION OP CilANG/,:
The pump shaft niulingiellerwillbe ieplaced on the r one t or i ec ii colnt inn punip "It ' . A1tbough it js cons ide re<l a "i tk o- f oi-lik e" icplacement, the r ep l a c e rnen t impeller hns minor dimensiotin! <lif f er ences in the itopeller dintnot er from that of the existitg impellnr. llowev e r , the difference is minor niul will not affect the pe r f oi rna nc e of the pu to p . This evnluntion will nihliess the diffetences in the cur r ent ly ins t a l led finpoller atui the replacement i nipe l l e r .
Rf.ASON FUK ChANGF.: The I rnpe l l e t will be t eplaced on t he "li" renctor recirculntion pump due to excer.sive vibratlon.
S AIT.TY f.V Al.U AT I ON : This sa f et y evalunt ton concluded t hat the chnnge did not itivolve nn unreviewed safety questfon. Repl a c e n,e n t of the pump laternals will not result in a change t o the operat ion or performance of the pun p or it s nssocint ed sys t em. The minot difference in itopeller dinmeter will not adveisely nffect the pump cnpacit y. There will be no I tupa c t to any intetincing syntem ns a result of t he r epi n c ernen t .
The replacement has been evnlunt ed ngninst the applienble design eriterin, instalIntton tequirements, and operntfonn) iequirements.
It was determined that all necessa ry s equ i t enient s r.nd cornm i t nient s nre met by the new coniponent and that ..o tiew acc ident p t er u r r. ors ate crented.
The exist Ing syst em and compotient design functions are not affected. Therefore, this chnnge will not reduce t he inntgin of safety as defined in the basis for any Tec hn teni .:peci f icat ion.
Npf.90/ SNI,1 Cl'l.K - 118 l
9
Attachment to GNRO 91/00001 l
SRASN: pl.S-90-011 1)oC NO: UFSAR Appendix 3A SYSTEM:
DESCRIPTION Or CllANGE: Thin chango doloten the referenco in UFSAR Appendix 3A which indicaton that SER1 will comply with Regulatory Guido 8.14 (1976), which addrennes pornonnel neutron donimotorn.
REASON FOR CllANGE: Grand Gulf no longer unen a separato donimetry
, for monitoring neutron exposure, and thereforn this Regulatory Guido dons not apply to our donimetry system. OGNS monts the ANSI N13.11 and 10CrR20,202 requirements for donimetry.
SAFETY EVA1,UAT10N: This safety evaluation concluded that the changn did r.ot involvo an unreviewed safety question. Personnel v
monitoring for radiat ion exposure in unreinted to any accidents
- previously evaluated in the FSAR. Personnni donimetry han no nf fect on or interface with any systemn reinted to plant saf oty.
This change han no ef fect on or interface with equipment important to safety previously evaluated in the UrSAR. It han no effect on the limit ing condition for >poration, applicability, action or 1 l nurvoillance requirements an defined in any Technical )
Specificntion. j i
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l PlS90/SNI,1 cpl.R - 119
Attachment to GNRO-91/00001 SRASN: Pl.S-90-012 DOC Not FSAR C/R 90-0005 SYSTEM:
DESCRIPTION OF CllANGE: This UFSAR change takes except ton to Regulatory Guido 1.137 step C.2.d(3) that requires removing condensato in the Diesel Generator fuel Oil Storngo Tanks one day af ter adding now fuel oil.
REASON FOR CilANGE: Chemistry samples and analyses are performed on new fuel oil prior to discharging to the ruel Oil Storngo Tanks. The sampling requirements are very stringent ($ .05 volumo porcent) thus controlling the amount of water added to the fuel of) tanks. Thorofore the sample required one day af t er adding new fuel is not- necessnry.
SAFETY EVAL.UATION: The chango does not involvo an unteviewed safety question. The Technical Specification sample requiremont for water (.0$ volume percent ) in new fuel precludes putting any s.lgnificant amount of water into the Emergency Diosol Generator Fuel 011 Storngo tanka. In additfon. the fuel Oil Transfor pump suction Ifno(s) nro located 8" above the bot tom of t he ruel Oil Storage Tanks. Water accumulation in the hot tom nf the storage tank would have to be significant (approximately 2000 gnis) hofore the Fuel Oil Transfer pump would pump water into the fuel oil system of the Emergency Diesel Generators. Presently water is ,
removed quarterly and loss than one gallon is routinely removed. i
! Beenuso of the stringent Technical Specification sampling requirements of new fuel (prior to adding to the fuel oil storngo tanks), the probnhility and consequence of equipment malfunction !
due to water intrusion into the fuel oil system of the Emergency F Diesel Generators is not increased.
l Taking exception to the Regulatory Guide 1.137 Step c.2.d(3) does not reduce the margin of satety as defined in the basis for Technical Specifications because the exception doesn't sitor tho ;
surveillance frequenclos or acceptanco critorin for water content in t he fuel oil.
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PLS90/SNh1CFhR - 120 L v+--y- .,w-,--, , ,,m-._ _ .mm-, -,.,w.w m.,_, m,,,.,,,, #,w r .,_,~m.-%... ,w,.~.m.m..-.,-.- i..-----.
Attachment to GNRO 91/00001 SRASN: PLS-90-013 DOC NO: TST1-1017-90-003 0-S SYSTEM: G18 I l
l l>ESCRipTION OF CilAN00: This change nilows the n&tition of sodium !
hypochlorito to a condenante phnso separator tank to stop !
microbiological activit y in t he t ank. j i
REASON FOR CllANGE: Thorn are methann-producing haeteria present i in the tank which causn prensuriant lon of the radioact Ive waste !
liner when the liner in dcwntared. The addition of sodium l hypochlorite to obtain n f rco chlorinn residual of 0.5 ppm for thirty minutes is necessary to provent the gas formation from I I
occurring SAFETY EVALUATION: The change does not involve nn sins eviewed safety question. The performance of this acttvity does not chango ,
the opornt ion of the phann separators, resin t rans for, or l downtering equipment. The chemical to he used will not bc ;
det rimental to the equipment in the concentrations to ho used.
Inndvertent spflingn of hypochlorito into the radwnsto system would result in the early changcout of a demineralized bed, but would not havn any ef fect on the integrity of the piping or ,
Componellt a . L Accidents evaluated in the UFSAR involving the radwnste system nrn i lenks/ tank ruptures in the system (15.7.2 and 15.7.3). System operation is not changed and the chemical is not detrimental to l the equipment. No different failure would be caused by this !
activity, which is bounded by thesn analyses of whole tank l ruptures. ;
i This activity meet s the requirements of the PCp nddressed in !
Technical Specifications and does not af fect the activity of ,
radwnste shipments.
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F pf,890/SNLICFLR - 121
Attachment to GNR0+91/00001
- SRASNt Pl.8-90 015 l>00 NO
- OQAM l'SAR 17.2 SY STI.H : N/A Dr. SCRIPT 10N Ol' CilANGl; This change rennsigun the tempOnnihilition for nudit a ntni evalunt tons of nuppliorn, review of procurement documents nrd roccipt inspection nn delinented in varioun policien
! of the OQAM to the Hannger, Quality Setvicen due to the t ransfer of the current Hanngnr. Quality Systemn to the Hannger, Quality j Servicen ponillon.
Rt.ASON FOR CilANGr.: This transfer of rer.ponalbilit ten will nilow consistonty in the adminint rat ton of thone act tvitles nint facilitate ant icipnted changen in the Qun;i ;' hegrama nren due to .
- connolidntlon.
SAP!:TY f.VAhtlAT10Nt The change doen not involvo an unreviewed nnfety question. These changen are administrative iti nature only and have no affect on nny component or system. Since theno changen do not delet n any t emponnihilit ten there in no t oduct ion in program requirementn. These tinns ferred f unctions are nt 111 being performed and the manngerial chnngen hnvo no ef fect on the nnfety of the pinnt.
The changen do not offect any bnnin in the Technical Speci ficat lonn .
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, Pl,S90/SNhlCFl.R - 122
- - .w. .ew-,...,,.w-w -mw--,-. e,- ,.-.m-e,v.-ow-+ ,,-...-,.-..w..-m+--sece
4 Attachment to GNEO-91/00001 l
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SRASN: PLS-90-016 DOC NO: 04-1-01-N19-1-TCN 25 SYSTEH: !
1 DESCRIPTION Ol' CllANGE: This procedure chnnge allown removal of the water seal from around the liigh Pronsure Condenner rubber i expansion joint while the plant is operating. I REASON FOR CllANGE: The removal of the water moni fr vice in I being performed as an interim meanuto to reduco 1cakng. . to t radwnste f rom the seal. ,
SAFETY EVAI,UA710N: The chango does not involvo an unroviewed .
nafety question. The removal of the water seal from service could result c.nly in increased air in-lenkago into the conienser and !
reduco the ability to detect the lonn or gross degradation of the rubber expannion joint during plant operation. The removal of the sent from service will not cause deterioration of the rubber joint above and beyond normal expected service life. Tho water seal does not directly or indirectly af fect any coniponent a other than the rubber joint. Thorn in no equipment import ant to safet y which could bn af fected by the removal of the water seal f rom service. !
The removal of the seal f rom servico does not reduce the margin of safety an defined in the basis for any of the Technical Specifications, because there are no nafety funct ions or safnty
- limit s which are associated or a f fect.ed by t he wat er seal.
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I PLS90/SNh!CFLR - 123 P
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v m wve-.,, w , r,+mww.. .- . .,..,. ,,--.-,vv ,,,.,,,-,-,-,,-.,3..-c.,~_w---em-,-uy_.-,.---.e,,-.,.,e..--,.--e-,
, -,, y.-,-,.,,-,p..,.--,_,9,--- ,-v.w.,__, m 9 - gr
Attachment to GNRO-91/00001 f 1
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j SRASN: P!,S90-017 DOC NO: WO# 00014194 SYSTEM: P44 i
l DESCRIPTION OF CilANGE: This tempornty change installed a supply :
, and return pipe for the drywell chiller cooling water which I originates from the Plant Service Water (PSW) piping.
I REASON FOR CilANUE: The four wny valve on the normal supply / return l PSW piping to 11. Drywell Chillers was obstructing flow. This j temporary chango bypasses the valve. i I
SArETY EVAbtfAT10N: The chan;;n does not involve an unreviewed '
safety question. This chango does not. af fect the overall flow balance of the PSW nyst em. The potential flows to CCW and Drywell Chillers during normal and I0P conditions have bonn nynlunted and determined necept able. The of fects of the piping nddition have ;
been evaluated and determined to be acceptable, i Standby Service Water (SSW) flow balance will not be adversely affected by this change. The remaining component s are non-sa fety !
related and not required to mitigato thn consequences of nn l nccident. The temporary four way valvo bypass does not ndversely '
affect any system as described in the basis of any Technical '
Specificatfon, u
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PLS90/SNblCFLR - 124
Attarheent to GNRO-91/00001 SRASN: plS90-018 DOC NO: Deleting Operntor SYSTEM:
Actions DESCRIPTION OF CHANGE: This change removes references to specific opernt or act ionn found in Chnpter 15 that nre not t orpilred by t hn safety nnnivsis basis and are not safety actions required to bring the pinut to n st nble condit ion. A stntement is .ulded t o referencn operat or actions t o t he Sit e Speci fic Opnrnt ing procedures and their proginnmntic rontrol. Operator nct ions ar e also removed from the Operating Modes of RC4C System section of Chapter 5. The operator act ion found in Chnpter 7 is deleted and reference to the guidnnce found in the Site Specific Emergency Procedures is added. Specific operator actions at the Remote shutdown panel found in Appendix 90 nre deleted niul reference to the guidnnce found in the Site Spec.ific Emeigency Procedures is added.
KEASON 00R CHANGE: Regulatory Guide 1.70, Rev. 3(15.x.s, 2n) requires that the Eve;nt Evalunt ion sect ion of the FSAR Chnptor 15 include a nequence of event s niul syst ems operations. This listing must include a st ep-by-st ep sequence of events from the event initiat ion to the final stabilized condit ion including all requirest operator actions. The required operator actions must include any operator act (on nssumed in the snfety annlysis and nny actions thnt are not part of the safety annlysis basis, but are safety nct ionr. required t o bring t he plant to a stable condit ion.
FSAR Parnainph 15.0.3.2.1.4 (s8) states "for all anticipated operational transients cited in Chapter 15, no operator corrective actinn is required to provent the plant from extcoding safety design basis limit s." (s9) stat es "In no cnse would t he operator's action or non-action result in an unncceptable ef fect on thn henith and sniet y of the general public The change clarifies that t he operator net lons pr eviously identifiel were not requirnd opernt or act ions assumed in the safety annlysis nor actions required to bring thn plant to a stable condition. The change is consistent with Regulatory Guide 1.70 for Event Evnluntlons.
SAFETY hVAl,UATION: This safety evaluntion concluded that the change did not involve nu unrnviewed safety question. Operator actions arn not taken until after an event has occurred and therefore have no effect on the prohnbility of occurrence. This changa will not ciente the possibility of an accident of a dif ferent type than any evnlunted in the PSAR because it only affects operator actions. The ef fect of single operator error is nirendy analyzed in the USAR and thnrefore bouuds thn scope of tnis change. The Site Specific Emergency Procedures nre nymptomatic in nature and provide guidance to mit ignt n the symptoms and to enintain the plant in a snfo condition regardless of what event occurred to generate the symptom or tegardlesn of what equipment is nynilable to combat the symptom.
PlS90/SNhlCl1,K - 125
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At tachmerit to GNKO-91/00001 1
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! Page 2 !'
l I i This change will not s educe the entgin to unicty an defineil in ihn !
1 basin f or niiy Tochtilcal Specif f ent ion beenuno t he Sit e Specif Ic -
'.j Emntgency Proceditt en are provided to thn operat or for mit ignt flig f
! any nymptom regardlenn of the itilt int ing event ated tilet efore !
actunlly incienne the meitgin to nnfety over the opeintor act lonN j
- prenent ly foutid in t he TSAR. [
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l i Pl.S90/T,NI,1CrlA - 126 I J
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Attachment to GNRO 91/00001 SRASN: Pl.S-90-019 DOC NO: TSTI-lE51-90-002-0-S SYSTEH: E51 DESCRIPTION OF TEST: TSTI-lE51-90-002-0-S places t he RCIC syst em in service in the Test Return Hodo of operation in accoidance with Sol 04-1-01-E51-1 to obtain differential pressurn thrust data on It51-F022 nnd IE51-T059, test return flow path isolation valves.
Onco in survice, the automatic opening funct ion of the RCIC minimum flow valvo, IE51-r019, will be defeated to 9110w determination of peak differentint pressureN nerosH the two valves during performanco of the test and to allow a highnr pressurn dif ferent ial to be devnloped across the valves. Uniture of the minimum f ew valve was assumed in the Maximum Expected Dif ferent ial Prensurn calculations for these valves. The notornat in closure of the minimum flow valvt. will remain effectiva during pnrformance of this test. Minimum flow control valvo operat Jon in the open diecct inn will be cont rolled via the Hnin Control room handswitch. Thrust data will be obtained at. a series of four independent differential pressurn data points.
REASON FOR TEST: The subject dnta in being obtniaed in an attempt to address the issues of GI, 89-10 and Gl. 89-10 Supplement 1.
SAJETY EVAL,UATION: This safety evaluation concluded that. the change did not involve an unteviewed safety questton. With tho exception of minimum finw valvo automatic opening, the RCIC system in operat ed in a normal system configurnt.Jon, t est return mode.
Operation of the RCIC system in this modo is a normal plant activity and does not increase the consequences of an accident.
The system / plant has been evaluntad tor this mode of operation in the originni plant design safety cynlunt. ton. The RCIC system ts not Operable (as defined in the Technical Specificallons) during performance of this test and as such no cindit enn bn taken for RCIC system operation in a capacity to mitignto events. The Technical Specifications provido the necessary finxibility for operation with the RCIC system inopernhin (provided llPCS is operable) for mit igat ion of analyzed events. Oparation of the RCIC system in the test return modn is a previously cynlunted modo of opornt inn for the RCIC system.
Thn RCIC system is declared inopernhin during performance of t his test. The llPCS syst em remains Opernhin during performonen of this tnst. 1he lipCS system providos the uncessary protect ion whnn the RCIC system is inoperable for the RCIC associated event analyses in the SAR. As this is the caso, ado < pint o capability exist s to maintnin event /necident mit igation mntgin for events annlyzed in the SAR. Manunt control will take the place of automatic open control of the RCIC minimum flow valvo and therefore RCIC pump integrity will be maintnined.
Providnd thn llPCS system is Opernble during performance of this TST1, the margin of safety is consistent with that discussed in thn BASES of thn Technical Specifications.
Pl.S90/SNLICFhR - 127
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Attachment to GNRO 91/00001 !
SRASN: pLS90-020 DOC NO: TSAR C/R 90-0008 SYSTEM: N64 f DESCRIPTION OF CilANGES: The following sentenen was deleted from TSAR Section 11.3.2.1.6.2 "During t ransfer of the charcoal into l
. the charcoal adsorbnr vessnis rartial sizing of thn charcoal will i be minimited by pouring the charcoal (by grnvity or pneumntically) over a cono or other instrument to sprend the granules over the ,
surface."
REASON FOR CHANGE: The deleted sentence did not describe hon the !
ndsorber vossols woro actually filled. I SAFETY EVAhUATION: This safety evaluation concluded t W the ["
change did not involve an unreviewed safety question. Tho adsorbor vessols worn filled during construction and havn perfotmed an designed. Through construction experience General ,
Elect ric has determined that the method of filling the adsorber vessnis does not affect adsorber performanco. The charcoal is i intended to last the life of thn plant. During construction the charcon1 was just poured in. This method would be reused if j change out is required in the future. If the adsorber vessnis ever had to ho filled ngnin thn post treatment radiation monitor would confirm thn charconi adsorber performonen. !
The Offgas System outlot vent. val';o ja int erlocked to the of fans post trentment.rndiation monitor that monitors radiation levnis at the outlet of the adsorber vessels and upon receipt of a !
prodotormined high high radiation alarm tho of fgas system is isolated f rom dischntging to tho environment.
The charcoal adsorber vessels are det.ign t o withstand a hydrogen detonation. The composit ion of the charcoal fill does not af fect the process system boundnry therefore t.his chango does not crontn '
thn possibility of an accident of n dif ferent type than any evalunted in the FSAR.
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The only safoty significance of the adsorber ves ois is the pressure boundary and the ability of the vesselt to be isolated by thn post treatment radintion monitors. The charconi fill dons not af fect either thn pressure boundary nor the ability of tho l rndintion monitor to isolato the vnssol:..
1 I This chango will not reduce the margin of sainty as defined in bases for any Tnchnical Specification becauno isotopic analysis l has verified thn ability of the charcoal adsorbers to doiny rolense of fission gases and keep the nf fluent release to utmosphere within prescribed limits. The charconi adsorber fill ;
does not offect the offgns pressure boundary and the offgas ,
process will be isolated from the offgas and radwnste vent upon recnipt. of a high-high radiat icin signal. .
plS90/SNLICFLR - 128
_ _ ~ _ _ _ _ _ . . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ . _ _ _ . _ _ . _ _ . . , . . _ _ . _ .._. _ _ _ _.,_._.._.
I Attachment to GNEO-91/00001 l'
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P1,5-90-021
! Page 2 The design basen for tbn Circ. Water t.yntem as defined in the GGNS l Technical Specification does not contain provisions for nny I specified margin of safety regarding the failurn of a circulating water system component. Therefore, implementfug this work order
]
, does not reduce tha margin of safety ns definnd in the banin for i any Technical Spectficsitlon.
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i PLS90/SNI.1CFI,k' - 130
Attachment to GNRO 91/00001 SRASN: Pl.S90-022 DOC NO: UFSAR 7.7.1.I1.4.2.b SYSTEH: P33 1
DESCRIPTION OF CilANGE: This change allom for chlorides to bc )
analyzed via the Post Accident Sampling i.p. tem (PASS) wit nin 4 '
days (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) instead of the current requirement of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
REASON 00K CilANGE: This change will bring thn UFSAR in complianco with NUREG 0737 Attachment 1, 11.11.3. This change allows the sample to decay for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> whleh reduces personnel exposure.
SATETY EVA1.UATION: This safety evaluation concluded that the chnnge did not involvo an unroviewed safety question. There is no accident evaluation on thn UFSAR for the PASS. PASS in used af ter an accident as a means to estimatn the extent of core damage but has no role in the mitigation of an accident or safe shutdown of the reactor. This UFSAR chant,o does not reflect any change to the PASS system or inter f aced systems.
The only equipment associated with PASS that is important to safety are the containment and drywell isolation valves of the reactor cotlant, suppression pool and atmospheric sample lines, j These valves are not. impaired by PASS sampling and nrn able to i perform their required isolation functions in the event of an ,
accident while a scheduled sampling evolution in it progress. Any '
PASS sampling evolution in progress during the occur rence of an accident would be terminated by the load shedding and sequencing ;
- system and automatic samplo line isolations. A manual reset in ,
required before sampling could resume. This UFSAR change concerns ,
samplo analysis which is portormed on a PASS grab sample and does .
not change the operation of the PASS panel but only clarifica the f analysis requirements which are performed af ter the sample in coll ect ed .
There arn no Technical Specifications bases applicable to PASS. I It is a non-safety related, non-seismic, and non-environmentally qualified system. PASS was constructed by principal construction codo 1431.1. There is no direct or indirect impact to any other margins of safety as definnd in the bases for any Technical SpneifIcations.
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Pl.S90/SNLICPLR - 131
. m. i -r e.w,.v,me,-4., , ..,- <.,., .. .., ,_ e,,_. . - --.-+.--....,e-,..-, .,.,._..-,,,..,-..,_m-r--_m__m.-..,w_,,,,-, e .-y.,w,,,,-.. v--, . n,,,..
Attachment to GNko-91/00001 SRASN: PLS90-023 DOC NO: UFSAR CR 90 010 SYSTLH: E12 DESCRIPTION OF CilANGE: This change added the following statoment to UFSAR 7.7.1.11.4.3: "The Suppression Pool, RilR-A and RilR-B, shall be snmpled through the Post Accident Sample System separately in consecutivo six-month intervals, rotating sampling personnel for training purposes, such that all thren points are sampled on an 18-month interval." This will increase the une of thn PASS system and require occasional operation of thn RilR system pumps for the solo purpose of taking samplos, f
l REASON FOR CilANGE: This was a mandated chango by the NRC and a documented licensing commitment , LCTS ID No. 15799.
SAFETY EVALUATION: This safety evelaation concluded that the chango did not involve an unrovir;ued safety question. Thero is no accident evaluation in the UPSA'. for thn Post Accident Sample 1 System (PASS). Ilowever, samp1!ng of the Supprossion Pool via PASS cnuses a loss of Division 2 Sugpression Pool levnl indication.
This instrument functional loss is temporary, lasting only while sampling is actually occurring. Loss of this instrument function places thn plant in a 7-dny LCO candit ton as por Technical Specification Tahin 3.3.7.5-1 (3.; Action 80. The PASS connection for sampling the Suppression Pool t.ac= of f of thn Division 2 Suppression Pool sensing line. This linn is equipped with a restricting orifico near thn Suppression Pool connectica point.
PASS samples f rom downst ream of this orifico and, while sampling, removes water fastor than make-up can occur through tho
, restricting orifice. This causes the instrument to indicato a falso low low Suppression Pool level. This inputs one cf the two required low-low levn) indications required for Suppression Poel make-up to occur. Therefore, if a single instrument failure along with a LOCA signal were to occur while sampling Suppression Pool via PASS, two Suppression Pool 1.ow-Low 1.evel signals would occur.
This would initiato the Suppression Pool Fake-up (SPMU) system, I
dumping the Upper Containment. Pool into the Suppression Pool. As a safety measure, a step is included in Chemist ry Section Instruction 08-S-04-954, which directs the taking of PASS 11guld samples, that requires Chemistry to have Operations to place thn SPMU Division 2 Modo Snlector handswitch, on Control Room Panel 11113-P-870 Section 10B, in the "0FF" position prior to taking a PASS Supprossion Pool sample. This overrides the SPMU function of the Division 2 Suppression Pool level instrumentation, preventing an inadvertent dump from occurring. Thereforo, the probability of an inadvertent SPMU dump is not increased. The action of placing the Division 2 SPMU Hodo handswitch to "0FF" is acceptable by entry into an LCO condition as por Tech Spec Sections 3.3.8 and 3.6.3.4.
PLS90/SNLICPLR - 132
Atinrbment to GNRO-91/00001 Pl.S- 9 0 - 0 2 3 Page 2 The PASS system is a non-anfety reinted system required by NURI'.G-0737 t o opernt o af t er n design basis accident . Although PASS is used for provhlitig informat ion regntT'ing the ext ent of corn damage following nn nccident , PASS pinys no part in init igat ing t he consequences of nn accident . The only equipment nssocint ed wit h PASS t hat is important to saf ety nic the c ont ainment and diywell isointion valves of the renctor roolnnt, suppiession pool and nirnospheric sample lines. Those isolntion valves required to be opeint ed for per f ormance of those samples (Suppression Pool, RllR-A and RdR-It) nic not irnpn ited by t hese snmpling events and should be able to perform their requited isolnt ton funct tons in the event of an accident shile a scheduled snmpling evolution is in progress. PASS and condensate cooling wat er (CCW) (used for sample coolers in PASS) nre shed in the event of an accident. *lhese systems are net required to init innt e t he consequences of an acc blent and site not required for a sn f o shutdown of the renctor. Any PASS sninpling evolut ion in progicss during the occurrence of an accident would be terminated by the lond shedding and sequencing system and automatic sample line isolations. A rnanun t reset is required before sampling could resume. Therefore, the ioutine snmples described in this UPSAR change will not create the possibility of an nccident of a different type than any already evalunted in the UPSAR (UPSAR references 1.2.? 8.2 and 7.7.1.11.4.2.1.).
PASS itself is a non-safety related system. The additional scheduled samples described in this UPSAR change requires running of the KilR- A and HilR-Il systems and pumps to sample from the respectIvo sample points, it is unlikely that nll of the scheduled PASS samples will coincide with scheduled running of these systems. Therefoic, EllR- A and RilR-11 will need to be started nnd run in olther renetor cooling or suppression pool cooling modes, depending upon pinnt conditions, for the purpose of collecting routine PASS samples. This will add a proportionally rmnll amount of run hours and went on t he systems, licw e v e r ,
assuming the worst ense that nll of the required samples over t he remnining life of GGNS opernting license required the start-up and running of t he Rilk syst ems for t he sole purpose of PASS samples with a conservative estimate of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of run-time pe. sample, would require only a max imum of 4 sarnples (18 monti, t ro piency) and 72 additional hours of ri.n-t ime f or ench of KilR- A and RilR-il pumps. This ndditional r un- t irne is not significant. over the li f et ime of t he RllR pumps and is therefore acceptable.
I Pl.S90/SNLICPI,R - 133
l Attachment to GNRO-91/00001 .
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l'I.S- 9 0- 0 2 3 Page 3 ,
There are no Technical Specificationn banca applicable to PASS.
It in a non-nnfety reinted, non-noimmic, and r;on-environmentally ;
qualified nyntem. PASS was constructed b) principal construction !
codo 1131.1. The operntion of PASS in principally for opernbility i verification and training with tho intent that it be nynllable for annnanment of core conditions following a design bane accident.
PASS in not used t o alt igat e the connequencen of an accident and '
in not required for safe shutoown of the reactor. There in no direct or indirect impact to any other margins of safety an ;
defined in the basen for any Technical SpecificnLions. '
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P1,S90/SNI.1CFI.R - 134
l Attachment to GNRO-91/00001 i I
f SRASN: PLS-90-024 DOC NO: 01-S-06-2 SYSTEH1 f 1 I DESCRIPTION OF CllANGE: This chango adds the Plant Supervisor (SRO) dutien and responsibilition to the conduct i-( Operation !
Administrative Procednro 01-S 06-2. ,
l KCASON FOR CilANGE: The addition of the third SRO to each shif t contributes to the experience and knowledge level to further '
enhance the nafe operation of the unit. l t
4 SAFETY EVAL.UATION: This safety evalunt ton concluded that the chango did not involyn an unreviewed safety question. The added experience and knowledge of the third Sko to each shif t improves overall shift performance and reduces the probability of occurrence of an accident. The added talcut of a third SRO I improves the performance of thn shif t such that if any abnormality occurn, event. evaluation and proper response tend to minimize the '
consequences of the accident. The established control room command structurn remains in effect ensuring continuity during ;
normal and abnormal conditions. Thn pre.4ence of the third SRO .
improves equipment monitoring thereforo detecting symptoms relating to malfunctions earlier. This earlier detect.fon could minimize the consequences of equipment malfunction. ;
The additional knowledge and experience provided by tho third SRO [
can only improvn compliance to Technical Specifications and '
related bancs. The shird SRO provides a valuable resource to l
discuss and evaluate conditions relating to Technical Speci ficat ton .:oncerns.
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i l'LS90/ SNI.lCFl,R - 135
Attachment to GNRO-91/00001 SRASN: Pl,S-90-025 DOC NO: HWP-90-1151 SYSTEMt 1,11 DESCRIPTION Or CllANGE: This safety cynluntion addresses operability of t he flattery Room flydrogen Detector Panel ll22-P535 for all plant modes of operat ion (Modos 1 through 5).
Relocation of the hydrogen detector panol will require the battery room hydrogen det ector circuit s to be inoperable for approximately seven days. This is considered a conservativo number to allow completion of work nnd subsequent re-calibration of the detectos ci rcuit s. During this period, ventilation systems will be verified operable on a daily basis. If ventilation is found to be not opernt ing, vent f lat ion will be restored, or portable hydrogen samples will bo taken dnity and upon overy access into a battery room whern the detector circuit is inopernblo. in addition, a wookly portablo hydrogen detector sample will be taken on all battery rooms where the detector is inoperable until ll22-PS35 is restored.
REASON FOR CllANGE: Replacement of UPS Inverters lY87, 1Y89, lY95 and lY96 will require relocat ion of Ilydrogen Det ector Panel ll22-P535 to facilitato maintenance on the now inverters.
This relocation will result in the liydrogen Detector Panel being inoperativo during the disconnection, relocation and reconnection.
SAFETY LVA1,UAT10N: This safety ovaluation concluded that the chango did not .involvo an unroviewed safety question. The battery hydrogen detector panel serves no safety function, nor is it required to be operable as part of the fire protection system.
None of tho accident s previously evaluated in t he FSAR are a f fected by the bat t ery room hydrogen detnctor panol. The battery room hydrogen detectors play no rolo in mitignting the consequences of any accidents described in the FSAR. The hydrogen detector panel performs an informat.f on function only. The hydrogen detector panel does not af f ect malf unction of any equipment important to safety. I No margins of safety as defined in the basis for any Technical Specifications are associnted with thn ilydrogen Detector Panel 1122-P05, t here fore there is no roduction in the margin of sa fety.
PlS90/SNLlrFI.R - 136 il
= -- =- . . . - - .,,== amer..-wv-w--e--r-r-2.-- ww-r, ...ew-rm-+----nr+-.+s-ews+weww.w---w-e*w--s- --e.=ever==.a-w w -~+=-ww a==--a-wee---rw-=ww-s----w*v,-eew-
Attnehment to GNRO-91/00001 SKASN: pbS-90-026 DOC No: DCp-88-0051 SYSTEM: h62 DESCRIPTION OF CilANGr.' Thin safety cynlunt ion addresses the opeinbilit y concer ns associat ed wit h supplying t empor ar y power in pinen of the normal invn:ters lY87, 1Y88, lY95 and lY96.
RF.ASON FOR CilANCE: The subject Invertets are to be teplaced.
SAFETY EVA1,UAT10N: This safety evnluntion concluded that the change did not involve nn unreviewed safety questlon. A temporary inverter will be supplied by a non operable (not the declared operable, but energized), clnss IE bus, thus failure of the tempornry invertet could not affett or degrade a class 10 power source. The output of each temporary inverter will be isointed from its respect ivo dist ribut ion panel via the pnnel circuit brenker nnd fused disconnects for each panel branch circuit. The inverter it self is considered non-essent ial since none of the circuits that power is being supplied to require power in order to pe r f o r rn their safety function. The ternporary power supplied is n knewn capacit y and qualit y , i.e., regulated to maintain voltage at 118 Voc plus 3 1/ 2* t o 2 1/ 2*. wl ;h less t han 5', ba rrnon ic d ist ort lon. Therefore, there is no degradation in quality of power by using the temporary supply. The inverter itself is not essent inl, therefore failure does not increase the consequences of any accident, further, the clnss lE circuits are isointed from the temporary invert er output vin n circuit breaker and fused d isconnect .
The only failure postulat ed is a f ailure of the tempornry invert er it self or the temporary cable supplied to t he UpS dist ribut ion panel. The power supply, i.e., the temporary inverter is non-ensentin), and failure does not prevent any safety functfon.
1ho circuits and inst run ent s that are being supplied power to, tio not require that power t o per form t heir sa fet y funct ion. Sensors, sensor channels and trip logics of the renctor protect ion syst em are not used directly for nutomat ic cont rol of process system.
Therefore, failure in the controls and instrumentation of process syst en s cannot induce failure in any port ion of the protection nystem.
Since the power supply is not rolled on 1o porform any snfet y functlon, the margin of saf ety as defined in t he basis for any Technical Specifications will not be reduced.
IhS90/SNI.1CFhk - 137
Attnchment to GNKO-91/00001 SRASN: phS-90-027 DOC NO: CR-90 011 SYSTi.M : N11 Dl: SCRIPT 10N Or CllANGl: This change m9kes the inspect icn int ervn1 of t he Turbine Stop and Control Valve at least once in 40 rnonths rather thari the pt vious once chch year.
Rr.ASON FOR CilANGr.: This change makes the UFSAR ngrea with the Technient Specifications f or the insp"ction interval.
SAPITY IVAl.UATION: This sn fety cynlunt ion concluded thnt the charige did not involve nn urireviewed sainty quest lon. Turbine Stop nnd Cont:ol VnIvo opernbi1ity t est ing for overspeed protectioe is nddressed in both the UFSAR nnd the GGNS Technical Specificottons. These Stop nnd Control Vnives ns well as the two overspeed devices are t est ed for opernhilit y every 14 dnys. This requirement remains n o c h n t.ged . The UFSAR is only being changed to make the valve f rispee i len cycle consistent with the r e<pii rernen t s of t he GGNS Tochn f ra1 Speci ficnt ion 4. 3.9.2c.
The design funct ion of t he overspeed pr otect ion syst em is not of fect ed and will perforrn in its intended manner. Overspeed prot ect ion for the Turbine / Generator is not compromised and will function in it8 int ended manner t o prot ect sn fet y related components, equipment, and st ruct ur es from damnge induced by Turbine generated missiles. ATT testing of the overspeed syrtem provides periodic t est ing t o ensure opernhilit y and integrit) of the Turbine Stop and Control Valves. Requi rernent s of t he UFS AR Section 10.2.3.6 for demonstrating the integrity of t he overs >eed protect ion systern by exercising of the ill , hp, and Dypass St op and Cont r ol Valves through ATT program t est ing rerna lns the same. The overspeed trip system will perform it s design f unct lon.
The mnrgin of safety as defined in the basis for nny Technient Spec i ficnt ion renniins unchanged. No GGNS Technical SpeciflentIon is affected. Only the requirements for vnive inspection intervals as listed in the UFSAR Section 10.2.3.6 is being changed to make it consistent with Technical Specifications, pl.S90/GNhlCl I.R - 1%
Attachment to GNRO-91/00001 SRASN: "I.F -9 0- 0 2 8 DOC NO: T.S. 3.0.4, ACTION C SYSTEM:
Dr.SCRIPTICM CF CilANGE: This safety evaluntton addresses the une of Technical Specificat ion 3.0.4 to enter Operational Condit ion 5 (iiigh Water 1,evel) ftom Operational Condition 5 (l.ow Wnter 1.evel) while complying with Action Statement c for Technical Specification 3.5.3.
kr.ASON FOR CilANGE- This nn fety evalunt ion document s the annlysis of ent ry int o Opernt ionni Condition 5 (lligh Wat er I.evel) from Operat ionni Condit ion 5 (I.ow Wat er 1.evel) when one suppression pool levnt instrumentation is inopernble due to removing r.CCS jockey pump from service.
SArr.TY EVALUATION: This snfety evnluntion concluded that the change did not involve nu unreviewed safety questlon. The Suppression Pool is required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of wat er is available to the llPCS ,
1.pCS nini 1.PCI systems in the event of a BOCA. This limit on suppression pool minimum volume ensures that sufficient water is nynilable to permit recirculation cooling f low to the core. The 01 ERAB11d TY oI t he suppresslon pool in OPERATIONAb CONDITIONS 4 and 5 is not required by Specificntion 3.6.3.1 for presr.ure suppression, in OPERATIONAI CONDITIONS 4 and 5 t he suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200 P. Since pressure suppres.ston is not required below 212'r, the minimum required water volume is based on NPSil, recirculat ton volume and vort ex prevention plus n 1 foot 2 inches safety margin for conservatism.
The UFSAR evaluntos several accidents (events) which are considered to be applicable during OPERATIONAb CONDITIONS 4 and 5.
The majority of these are unrelat ed t o the proposed applient inn of TS 3.0.4 for TS 3.5.3 and thus the probability of occurrence of these events does not inciense. The react or drain down event is not specifically addressed in the liFSAR during OpERAT10 nab CONDITIONS 4 atul 5; however, events that result in reactor vessel invent ory loss (i .e. , reactor drain down) are most directly affected by the ove of TS 3.0.4 while in ACTION c of TS 3.5.3. In accordance w i t h TS 3. 5. 2 and 3. 5. 3. f.CCS niul t he sr ession pool are not required tc be OPERAHl.E provided that the renctor vessel hend is removed, the envity is flooded or being flooded from the suppr ess ion pool, t he reactor cavit y and t ransfer canal gates in the upper containment pool are removed when t he cavit y is flooded and the water lesnl is mnintained within the limits of TS 3.9.8 and 3.9.9 Therefore, provided TS 3.5.3, ACTION c is complied with (verify suppression pool level at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by ,
nn niternnte indientor) the flexibility provided by the provisions of TS 3.0.4 does not incrense probability of an accident previously evaluated in t he UFSAR.
l PLS90/SNhlCFI,R - 139
Attachment to GNRO-91/00001 PhS-90-02ft Page 2 The accidents considered by the UPSAR during shutdown condit tons are not changed by the use of TS 3.0.4. This appilcation of TS 3.0.4 neither adds or removes systems or components, nor does it chango pienent system design fontures or plant operating procnduros. No new mechanism for draining the reactor vessel is created.
Thn bases for TS 3.5.3 discusses the need for suppression pool volumn during OPERATIGNAL CONDITIONS 4 and 5 is to prevent NPSil l concerns, provide recirculation cooling volume and vortex prnvention. Complyhig with TS 3.5.3 ACTION c will ensure that those concerns and the margin proventing t heno concerns are adnqunt ely addressed during flooding of thn reactor cavity.
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At t nthrnent to GNkO 91/0000)
SRASN: Pl.S-90-029 I)DC NO: CR-90 006 SYSTEM: N11 DESCRIPTION OF CHANGE' 1he purpose of this saf ety evnlunt ion is to evalunto tha impact on plant safety of delet ing sent ence 8 of the UTSAR Section 10.2.2.4.
REASON FOR CHANGE: UFSAR 10.2.2.4 sentence 8 states "All motor-operat ed valves will be bench t ested or in place t sst ed",
referring to motor-operated valves associnted wit h the Tut blue Generator system. The sentence does not specify the type of testing or the bnais for snth testing. No other requirements exist which reference this sentence or explain the type or basis f o .- the testing. Due to these facts, and hnsed on a review of the plant progrnms currently in pince t o monit or the condition of the motor opernted valves, it has been determined that Sentence 8 of UFSAR Sect ion 10.2.2.4 serves no usef ul purpose and sh >uld be de1eted f rom t he UFSAR.
SAFETY EVAL.UATION: This sn f ety evalunt ion concluded that the change did not involve nn unroviewed safety question. If any of the motor-operated volves malfunction, such that steam flow could not be stopped the steam would be exhausted into the coedenser, and only result in n lons of plant megawntt output. Also, the stenm supply to these valves will be isointed in an accident. The motor operated valves do not directly or indirectly affect any components necessary for safety. There is no equipment important to safety which could be a f f ected by delet ing t he bench or in-pince t esting of all motor-operat ed valves in the turbine generator system.
Deleting in-pince or bench testing of nll motor-operated valves in the turbine generator system does not reduce the margin of safety as defined in the basis for any of the Technical Specifications, because there arn not safety functions or safety limits which are associated or affected by the test ing of these valves.
Pl.S9 0/ SNI.1 Crl.R - 141
Att achmonit to GNRO-91/00001 i
SRASN: PLS-90-030 DOC NO: TEMP AhT 90-0004 SYSTLH1 P47 l DESCRIPTION Ol' CllANGE: This change abandons the prelubo system on ]
l radial wolin 1, 3 and 5. ;
I 2 REASON FOR CllANGE: The subject prolube systemn worn unnecennary for proper operation of the radial wolin and worn high maintenance j i items. '
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SAFETY EVAhUATION. This safety evaluation concluded that thn i chnngo did not. involve an unroviewed safety quent lon. Thin !'
temporary alteration (TA) does not, af fect the operation or
- rollability of any safety related system, No accident. cvaluated l in t ho UFSAR in aifect ed by t his TA. Thia TA dons not affcct tho ,
operat ion or rollability of the radini well system as described in l the l'S AR . !
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Attachment to GNRO-91/00001 l !
SRASN: Pl.S90-031 DOC NO: W.O. 27751 SYSTEM: D17 l 1 l DESCRIPTION OF CllANGE: This change provides temporary !!OP power for the Auxilinty liullding fuel Handling Aron Ventilation Exhnust .
thn Auxi1in1y Ihitlding runi lintulling Aren Pool Sweep Exhaunt, the Containment and Drywell Ventilation Exhnunt niid the Control Roo9
- Vent fint ton Rndint ion Honitoring Syntemn.
REASON FOR CllANGE: The tennon for the need for temporary 110P q pownr wnn due to n Itun 15 and a llun 16 outogo.
SAFETY EVAht!ATION: This saloty evaluation concluded that. the I chango did not involve an unreviewed safety question. The loss of the rad iat ioti monitoring nyntems' tempornt y 110P pownr or degrndation of that power will enuno tho initintton of the intended anfety function. Loss of power to the radintion monitoring nynteen will actunto t he npproprint o annuncint or in the main control room. Degrndat ion of the power will enuno initiation hecnuno decro.ining voltage will enuro a indintion monitor high
, volt age (downnenlo) inop trip. An increnso in the temporary power's voltage will enuno an increased rndfation indicatton. A l
constant voltage t ransformer will bn unod to condition the tiOP temporary power to maintnin rolinhility of the radint ion monitor 1 power supplien. Thn constant voltage transformer's output will be hold to 120 VAC with it s input voltago varying f rom 95 to 130 VAC.
With tempoinry power applied to the radiat ion monitoring system an i isolation will occur upon a high-high radiation or an inop signnt.
A high rndintton signal will cnunn an alaim in the main control room, l.onn of power will enuse an inop nignal and thun an inolation will occur. Per thn GGNS TSAR thn safety functions of theno radiat ion monitorn in to isolate ventilnt ion nyntemn and or start the approprinto filtratton system and to provide indiention l and alarm in the main cont rol room. i This chnngo does not reduco the margin of nafety an doncribed in !
l the hanin for any Tnchnical Specification beenunn the margin of uninty in maintained by the initintion of thn intended safety !
i functfon. Any f ailuin of the temporary power supply will result.
In the initiation required to ensure safety.
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Attachment to GNRO-91/00001 SRASN: plS 0"l2 DOC NO: HWO 26063 SYSTEM: R21 i l
DESCRlpTION OF CllANGE: HWO 26063 provides t emporary power f rom ,
ESF Bus 16AD and B0p Buses 11!!D and 13AD to londs normally !
4 supplied by Bus 15AA. The additional power requiroitents being pinced on Busen 11HD and 13AD are negligible and no londing eniculations wnrn required. No additional lond is being placed on l Bus 16AB. No components being supplied temporary power will bo l considorod operabic, in all cases temporary power was being i supplied as a matter of convenienen and not plant safoty. '
Required 1,00s woro entered when normal power was removed. All work was done while in Reactor Hode 5. Temporary power was '
supplied as shown in Table 1.
TABLE 1 ~
Tempo ra ry_ f.onsis 1,onds Normal power Supply Temporary _ power Supply Battery Charger IK4 52-15104 52-16106 linttery Charger 1D4 52-15102 52-132249 .
highting XFMR 1X113 52-154224 52-111217 T Refuni platform 52-154223 52-111217 SI.C Operat ing llenter $2-154221 52-111219 REASON FOR CHANGE: To allow requirnd maintenanco and cican2ng of the 15AA EST Bus.
SAFETY EVAbCATION: This safety evaluation concluded that the chango did not involvo an unreviewed safety question. Temporary i power will be supplied in a similar manner as the normal power supply. Cnbin sizing and breaker solection will be such that adequatn circuit protection is enintained. The lands bning ,
supplied tempornry pwer will not be relied upon to perform a safety function. Review of the load shedding tables in the FSAR [
shows that all loads being supplied temporary power are '
non-essential. In any ac.cident situation in which lond shedding H worn to occur all loads 1isted in Table 1 would be shed, by either thnir normal or temporary supply. The only possiblo failure of ;
thn circuits supplying tonporary power is thnfr loss'of power. !
Regardless of 1ow that, loss occurs, the end result is failure of component to function, boss of power to all components Ifsted in i Table 1 has already been considered. Using Buses 11110, 13AD and t 16All does not diminish the quality of power to the temporary '
londs, nor does it decrease the rollability of the power nyn11abin ,
to the loads normally supplied by Dusca 1111D , 13AD and 16AB. t Breaker S2-16106 which normally rapplies power to bat tery charger 11,4 will be disconnected and reconnected to battery charger 1K4.
This does not constituto a viointion of divisional separation and it does not increnso thn lond on ESF Bus 16AB sinco battery chargers 1K4 and IL4 are identical. None of the loads being ,
supplied power arn required to perform safety functions, and in ,
pLs90/SNLICrLR - 144 t
Attachrnent to GNRO-91/00001 I
PLS-90-032 Page 2 i
the caso of any load shr'.iding accident approprinto lond shedding of londs in Table 1 will be accomplished.
Sinco Technical Specif tention requirements will hn mot. with ,
Division 11 and/or Division 111 operability, and nonn of the londs !
1.nink supplied t emporary power will be required to perform any roft.ty function, the margin of safety an defitied in the basis for any Toc!nicrl Specificat ion will not be reduced.
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Attachment to GNKO-91/00001 i SRASN: PLS "0-f2P DOC NO: MWO 26064 SYSTEM: R21 DESCRIPTION OF CilANGE: MWO 26064 provides temporary power from .
ESP Bus 15AA und 30P Buses 1211E and 2111D to loads normally .
supplied by hus 16AB. The additional power requirements being !
placed on Buses 12ilE and 211lD are negligibic and no loading calculat.fons were required. No additional load is being placed on Bus 15AA. No components being supplied temporary power will bo ;
considered operable. In all cases temporary power is being supplied as a matter of convenience and not plant safety. -
Required LCOs were entered when normal power was removed. All -
work was done while in Reactor Mode 5. Temporary power was supplied as shown in Table 1.
1 TABLE 1 Temporary _ Loads Loads Normal Power Supply Tempo _ra ry_ Power Supply Battery Charget IE4 52-16102 52-124118 Battery Charger IL4 52-16106 52-15104 Lighting XFMR IX114 52-164211 52-125125
< Drywell Floor Drain 52-1P66111 Control Room Sump Recorder Wall Socket 4 Unit 1 Inst. Air Dryer 52-lP64218 52-213104 REASON FOR CllANGE: To allow required maintenance and c1 caning of the 16AB ESP Bus.
SAFETY EVALUATION: This safety evaluation concluded that the change did not involve an unreviewed safety question. Temporary power will be supplied in a similar manner as the normal power supply. Cable sizing and breaker selection will be such that adequate circuit protection is maintained. The loads being srpplied temporary 1.ower will not be relied upon to perform a safety function. Review of the loud shedding tables in the FSAR shows that all loads being supplied temporary power are non-essential. In any accident situation in which lead shedding were to occur all loads listed in Table 1 would be shed, by either their normal or temporary supply. The only possible failure of the circuits supplying temporary power is their loss of power.
Regardless of how that loss occurs, the end result is failure of l component to function. Loss of power to all components listed in l Table 1 has already been considered. Using Buses 12ilE, 15AA and 21HD does not diminish the quality of power to the temporary i loads. Nor does it decrease the reliability of t he power availabic to the loads normally supplied by Buses 1211E, 15AA and 2111D. Breaker 52-1510^ which normally supplies power to battery charger 1K4 will be disconnected and seconnected to battery charger IL4. This dnes not constitute a violation of divisional separation and it does not incrensa the load on ESF Bus 15AA since l battery chargers 1K4 and IL4 are identical.
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l Att.achment to GNRO-91/00001 Pl.S-90-016 Page 2 None of the loads being supplied power are required to perform r,afety functions, and in the case of any load shedding accidnnt appropriato load shedding of loads in Table 1 will be accomplished.
Since Technical Specification requirements will be meet with Div 1 and/or Div 111 operability, and none of the loads being supplied ten.porary power will be required to perform any safety function, the margin of safety as definad in the basis for any Technical Specifications will not be reduced.
PLS90/ENI.ICFI.R - 147
Attachment to GNRO-91/00001 SRASN: pl.S-90-043 DOC NO: TST1-lG17-90-004-0-S SYSTEM: G17 1
DESCRlpTION OF CilANGE: This activity will add Dearborn 702 biocido to a condensato phase separator tank to stop microbiological activity in the tank and allow burial.
REASON FOR CilANGE: There arn methano-producing bacteria present in the tank which cause pressurization of the radioactive waste linnr when the liner is dowatered. The addition of 600 ppm Dearborn 702 is necessary to prevent the gas formation from occurring. The following steps were taken to develop this process:
- Samples of the tank worn taken to determinn the dosagn necessary to kill the bacteria.
- The radwaste liner supplier was contacted to ensure that the use of Dearborn 702 would not. advnrsely affect the liner.
- The ChemNuclear burial facility .in 11arnwell, South Carolina was contacted to ensure the t.reatment process was in accordance with the burini regulations at the site.
- The resin manufacturnr was contacted to ensurn the treatment process would not adversely a f fect the wasto resin and filter media.
- The solidificat.Jon process vendor was contacted to ensure thn addition of this chemical would not adversnly ef fect the solidification process.
This safety evaluation was applicabin for the treatment of both l thn cor.densate phase separat ors and the Reactor Water Cleanup phase separators with Dearburn 702.
SAFETY EVALUATION: This safety evaluation concluded that the chango did not involvo an unreviewed safety quent lon. As stated in l'FSAR 11.4.1.1, the radwaste system ". . . is designed so that failurn or maintenance of any f requently used component shall not impair system or plant. operation." The performanco of this activity dons not changn the operation of the phase separators, resin t.rans for, or dewatering equipment.. The chemical to be used will not be detriment.a1 to the equipment in the concentrations to be used. Inadvertent spillagn of Dearborn 702 into the radwaste system could result in the early changnout of the domineralizer bed, but would not have any affect on the integrity of the piping or components.
The radwaste system is not necessary for the safe operation or shutdown of thn plant. A failurn in the radwaste system would havn no applicable effect on the core or NSSS performanco (15.7.2 l and 15.7.3). Acc id en t.s involving the radwaste system are bounded l by the UFSAR wholn tank rupturn analyses.
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Attachment to GNRO-91/00001 PLS-90-043 l Page 2 1 Accidents evaluated in the UFSAR involving the radwasto system are leaks / tank ruptures in the system. System operation is not changed and the chemical is not detrimental to the equipment.. No dif ferent failuro would be caused by this activity, which is bounded by these analysos of whole tank ruptures. This activity !
meetn the requirements of the PCP addressed in Technical Specifications. This activity will not affect the activity of radwasto shipments.
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Attachment to GNRO-91/00001 SRASN: Pl.S-90-044 DOC NO: WO #29996 SYSTEM:
DESCRIPTION OF CllANGE: WO #29996 connects a mobile demin water trailer to temporarily supply domineralized water to the Demin Wator Storage Tank. This supply of water will bo made through the manua1 "modiffed" valyn NSP21F077. The vaIvo was modifled to facilitate a connection for the temporary mobile demin water trafier through the valve bonnet via a hose connection from the trailor. Temporary valve connections controlled by Temp Directivo 04-S-01-P21-1-TEMP 17 Rev. O, will allow chemistry sampling and analysis of the water supply prior to connection t.hrough the modified valve to the Demin Water Storago Tank.
REASON FOR CilANGE: .To provide a temporary domin water source to the Demin Water Storago Tank.
SAFETY EVALUATION: Since no FSAR accident is postulated on any demin water failure, this safety evaluation concluded that the change did not involve an unrnviewed safety question. The subject activity represents a temporary changn to P21 domin water system as described in the FSAR only becauso valve NSP21F077 will be in ef fect removed (will not serve as a valyn) during the durat ion of the activity. Final supply water quality will be well within the conductivity parameters given in FSAR 9.2.3.1.2.
P21 domin water has no safety-related function. P21 serves no system or component in a way vital to reactor shutdown. Thoro is no reduction in the margin of safety as defined in the basis for any Technical Specification, because no Tech Spec govnrns the filling of P21 domin water or the manipulation of valva NSP21F077.
PIS90/SNI.ICFLR - 150
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Attachment to GNRO-91/00001 SRASN: NI.S 002 DOC NO: SERI Operations Manual to SYSTEM:
Operations Hgmt Manual DESCRIPTION OF CilANGE: This changn changes thn title of tlin SERI Operating Manual to the Entergy Operations Management Manual.
REASON FOR CilANGE: This manual title chango reflects thn new namn of the company.
SAFETY EVALUATION: With this title chango all the responsibilities and commitments bning performed in the existing operating manual will continuo to be performed. The title change will havn no of fcct. on plant design or operations; therefore, there will be no increaso in probability of occurrence or consequences of accidents proviously evaluated in the UFSAR; nor will thorn bn any lucrease in probability of occurrence or consegunnces of a malfunction of equipment important to safety previously ovaluated in tho UFSAR; nor will thorn be created thn possibil!ty of an accident or malfunct.fon of equipment important to safety different than any previously evaluated in the UFSAR.
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Attnchment to GNRO-91/00001 SRASN: N!.S-90-003 DOC NO: GGNS Emergency Plan SYSTEM:
l Section 6.6.86 DESCRIPTION OF CllANGE: Tha word " drinking" is doloted from the sentenco in the Emergoney Plan Section 6.6.86 that stated: The requirements of 10CFR20, Appendix II, are met for air and drinking water.
REASON FOR CilANGE: 10CFR20, Appendix B is not applicahin to drinking water.
SAFETY FNA!.UATION: Th'is sofoty evaluation concluded that the chango did not involvo an unreviewed safety question. With this chnugn, thn Radioactivo 1.fquid and Gaseous Wasto Sampling and Analysis Program will continuo to be performed as required in 3/4.11 of the GGNS Unit Onn Technical Specificat ions. The change will havn no offect on plant design or opernt.lons; thornforn, thorn will bn no increase in probability of occurrence or consoquenens of accidents previously evalunted in the FSAR; nor will thoro be any increnso in prohnbility of occurrence or consnquence of a malfunction of equipment important to safety previously evaluated in the FSAR; nor will thorn be created the possibility of an accident or malfunct.lon of equipment important to safnty different. than any prnviously evaluated in thn FSAR.
The 11ASES section of the Technient Specifications provide general rnquirnments applienhlo to each of the Limiting Conditions for Opernt.fons and Surveillanco Requirements within Section 3/4, and the justiff ention for Safety System Settings. The bases for thn Radioactivo Effluents i.COs will not be altered as a result of this changn; thus the margin of safnty as dnfined in the bases for any Technten1 SpecificntJon is not. reduced.
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Attachment to GNRO-91/00001 l
SRAbN: NLS-90-004 DOC N0t TS 3.7.2, ACTION b.1 SYSTEH:
DESCRIPTION OF CllANGE: The safety evaluntion addresses the use of TS 3.0.4 for entry into Operational Condition 4, 5, or
- when ono control room emergency filtrat lon (CREP) subsystem is inoperable.
The specified condition
- is defined as "when irradinted fuel is being handled in the primary or secondary containment."
REASON FOR CilANGE: During refueling outages, situntions may arino due to maintenanco, implementation of modifications, or survoillances such that it is necessary to entor into one of the subject OPERATIONAL CONDITIONS or specified condition with a CREF subsystem inoperablo. This nynluation nasumes onn CREF subsystem remains OPERA 11LE and ja operating in the isolat ion modo of operation.
SAFETY EVALUATION: This safety nvaluntion concluded that the change did not involvo an unreviewed safety question. Th, CREFS is not a component procursor to any of the accidents evaluated in the UFSAR. Additionally, the operation of onn subsystem of the CREFS in the isolation modo as required by TS 3.7.2, Action b.1, is in accordance with snfoty design basis defined in the UFSAR.
Operation of one of the CREFS subsystems in thn isolation modo prior to or following OPERATIONAL CONDITION changes or specified condition changes does not af fect ita opernt17 ns or the operation of equipment that. could be procursors to accident s. The Safety design basis of the CREFS, in conjunction with other control room design provisions, is to ensure that the control room will remain habitable for operations personnel during and following all design basis conditions, and that the radiation exposure to the personnel will bn 5 rom or less whole body in accordanen with GDC 19 of Appendix A, 10 CFR Part 50. Operat ion of onn subsystem of the CREFS in the isolation mode is in accordance with its safety design bases as defined in the UFSAR, and is valid for all OPERATIONAL CONDITIONS including 4, 5, and when handling irradiated fuel. The system is also designed to allow for isolation modo f resh air makeup to allow for dilution of carbon dioxide (CO2) buildup. The system design allows for manual initiation of the fresh air makeup 10 minutes following initiation of thn isolation modo; however, fresh air makeup is not required until approximately 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following isolation based upon CO2 buildup f rom respiration of 12 persons as nascribed in UFSAR Section 6.4. Due to the potential bulldup of CO2 with the CREFS operating in t.he isolnLion modo for extended periods of time during OPERATIONAL CONDITIONS 4, 5, and handling of irradinted fuel in accordanco with TS 3.7.2, Action b.1, and TS 3.0.4, CO2 l levels could potentially buildup to higher than normal levels. In l accordance with Station Operating instruction No. 04-3-01-Z51-1, l llenlth Physics will sample the control atmosphere every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to l ensure that the oxygen levels remain above 20% and the CO2 levels remain bn' low 1%. The fresh nir makeup would then be utiiIzed as NLS90/SNLICFLR - 153
Attnehment to GNRO-91/00001 NhS-90-004 Pagn 2 required and is available for use within ten minutes af ter the isolation signcl. The only accidents which are evalunted in the UFSAR that rely upon the CREFS for consequenen mitigntton are LOCA
- inside containment (UFSAR 15.6.5, NSDA Event 37, Figure 15A.6-37) nu,i steam line break outside containment (UFSAR 15.6.4, NSOA Events 38, 39, 6 40, Figure 15A 6-38). The LOCA and steam line break accidents nre not applicable to the OPERATIONAh CONDITIONS 4, 5, and when handling irrndiated funi being evaluated for thin technical specification action statement. The fuel handlit.g nccident (UFSAR 15.7.4, NSDA Event 36, Figure 15A.6-36) atui the liquid radwaste system failures (UFSAR 15.7, NSOA Events 43 5 44, Figures 15A.6-41, 42) were also reviewed for potential consequence i m pa c t. . Again, operation of the CHEFS as evaluated during OPERATIONAL CONDITIONS 4, 5, and when handling irradinted fuel, or when changing between these, does not impact these accidents. The operation of the CREFS as described in TS 3.7.2, A: tion b.1, is .o accordance with the snfety design basis of the CREFS, and does not i mpa c t. the cont.rol room IIVAC f ailure analysis and Control Room Atmospheric Control and Isolation System (CRACIS) failure analysis presented in UFSAR Table 9.4-2 and habitability of the control room and does not negatively impact the mild environment equipment qualification requirements for equipment in the control room. Operating the CREFS in the isolation modo during OPERATIONAh CONDITIONS 4, 5, and when handling irradiated fuel, or when entering into one of these conditions, does not have any control interface to other equipment important to safety-The change in OPERATIONAh CONDITIONS or specified ccndition with the CREFS operating in accordance with TS 3.7.2, Action B.1, is in accordance with and does not reduce the margin of safet.y described in Technical Specification Bases for TS 3.0.4. In addition, the TS Bases for TS 3.7.2 requires operation of the CREFS so that the control room will remain habitable for operations personnel during and following all design basis accident conditions, and limit the radintion exposure to 5 rem or less whole body. This requirement is met by one subsystem of the CREFS being placed in the isolntion mode of operations and does not reduce the sa fety margin of the CREFS. This is applicable whether operating in one of the applienble conditions or entering into one of the conditions.
Finnily, the operation of the CREFS does not adversely impact or interact with other pinnt systems ns described in the Technical Specificntton and their bases.
l 1
N'.S90/ SNLICF1.R - l 'i4
- ~ . ~ . - - . - - . - - - _ - ----- . - . - _ - - - ---
T Attachment to GNRO-91/00001 SRASN: NhS-90-005 DOC NO: TS 3.6.4, Actions b & c SYSTEM: ,
DESCRIPTION OF CllANGE: This evaluation addresson the safety implications of commencing core alterations and/or handling of irradiat.ed iunt in the primary or secondarv containment with containment and/or drywell penetrat.Jons already isolated by an acceptable method as allowed by TS 3.6.4 Action b or c as compared to taking these actions af ter beginning coro alterations or the handling of irradiated fuel, REASON FOR CilANGE: During refueling outages, various isolation valves must be made inoperable to por:orm maintenance, conduct surveillance t ests and inspections, or implement design changes.
TS 3.0.4 allows the plant to begin core alterations or the handling of irradiated funt without having all required isolation valves OPERAliLE provided that the requirements of thn applicable Action Statements are met.
SAFETY EVAhUATION: This safety evaluation concluded that the change did not involve an unreviewed safety questine. The function of the containment and drywell isolation Talvon is to ensure that drywell and containment penetrations are isolated in the event of a radioactivo release inside the containment. This assures that an environmental relonso of radioactivo material is controlled to within the design leakage rate of the containment.
systems, thereby preventing offsito doses from exceeding thoso determined by plant safety analyses. During core alterations or the handling of irradiated fuel in OPERATIONAL CONDITION 5, certain containment, and drywell isolation valves (Groups 5, 6A, 6B, 7,.8, 10) are required to ho OPERABLE as specified in TS 3.3.2 to mitigato radioactive releases which might occur. The UFSAR considers events which may potentially result in a radioactivo ,
releaan during refueling whilo performing core alterations or the handling of irradiated fuel. These include inadvertent criticality, failures of various plant systems and components, loss of of fsite power, and fuel handling accidents. Of those, only the fuel handling accident insido containment generated a radiological release which results in the need for automatic isointion of containment and drywell penetrations. Should isointion valvos become inoperable while performing core alterations or the handling of irradiated fuel, Action b or c may be entered to indefinitely provide an equivalent levol of protection by isolating the affected penetrations. Under TS 3.0.4, Action b or c will be taken prior to beginning corn alterations or the handling of irradiated fuel for those penetrations with inoperable isolation valven. Footnote
- is also present to assure that isolation valves remain closed to provide a level of safety equivalent to the LCO when beginning core alterations or the handling of irradiated fuel. This flexibility has no af fect on the methods or equipment used for fuol handling or the monitoring and control of refuel.ing activities. The flexibility of TS 3.0.4 as applied in TS 3.6.4 also does not NLS90/SNLICFLR - 155
Attachment to GNRO-91/00001 l
[
N!.S-90 005 Pago 2 changn or af fect the number of activition defined as coro alterations. Thern are no changes in refueling interlocks, so the probabilit.y of an inadvertent criticality is not increased.
Isolating any penetrations having inoperable inolation valves before beginning corn alterations or the handling of irradiated fuel completely fulfills the safety functton of the valves. The radiological consequencos of a fuel bandling accident will thus be no worso than analyzed. Also, none or the analyzed accident sequences are changed by isolat ing the af fected penetrations prior to beginning core alterations or the handitn3 of irradiated fuel rather th:.n at some lato. time. Fun! handling techniques and equipment are not altered, monitoring and control methods are not modified, nor are the types of activities defined as coro alterations changed. Refueling interlocks remain unchanged. No radioactivo material release mechanism or path is created where none previously existed. Exercising the provisions of TS 3.0.4 in this caso maintains the plant. In an acceptably safe condition relative to the radiological consequences of potential accidents during coro alterat.lons or the handling of irradiated fuel.
This application of TS 3.0.4 may dirnetly affect equipment important to safety in two ways. Firritly, the isolation valves and pennt. rations themselves will be affected dun to the requirement to close and/or deactivate valves or affix blind ,
flanges in order to isolate panotrations. Secondly, systems and equipment served by the penetrat ions may also be af fected duo t o thn blocking of various finw paths. Refueling equipment and other-plant components are not impacted by this usn of TS 3.0.4.
The containment /drywn11 penetretions will bo inolated under the provisions of TS 3.6.4, Action b or e in the event that their isolation valves are madn inoperablo for outagn activities prior to or whiln core niteration or irradiated fuel handling activitics worn underway. Thoro is no additional effect on the valves and penetrations themselves as a result of pntforming thn isolation prior t.o beginning core alteratices or thu bandling of irradiated fuel. Thn mnthod accomplishing thn required isolation is identical in nither case, and the maintenance or testing of the valves or penetrations will also be unchanged. Similarly, systems whose flow paths arn altered as a result of isolated penetrations will bo impacted in the namn manner regardless of the timing of the Actions.
l l
NI.S90/SNI.ICFl.R - 156
_---- _ _ , . ~ _ __ - ._ _ _ _ _ _ ._ . - , _ - _ _ _ _ _ _ _ _ _ -_ -.
-_ . . - ~ . . . ._
Attachment to GNRO-91/00001 NhS-90-005 pago 3 Ilaving performed the required isolatlons prior to beginning corn alterations or the handling of irradiated tuni may actually reduco the probability of an equipment malfunction by rnducing thn amount of system and component manipulation otherwise required. Without the rollof of TS 3.0.4, each timo coro alterations or the handling of irradiated fuel wnro to commence, any isolation valvos undergoing maintenanco or testing would havo to first be mado OPERABhE. Then coro alterations or irradiated fuel handling could begin and the valves subsequently declared inoperablo and TS 3.6.4 Actions taken.
Tho equipment impo r t.an t to safety which may be directly affected by this application of TS 3.0.4 includes the isolation valves and ponntrat. inns required to ho OPERABhE by Specification 3.6.4 during core alterations or the handling of irradiated fuel as well as the equipment and components in systems served by thoso penetrations.
The radiological consequences of a malfunct lon of such equipment ir, not increased by taking the required actions to isolato penetrat.Jons prior to beginning core alterations or the handling of irradiated fuel rather than af ter coro alterations or tho handling of irradiated fun! havn begun. The degree of isolation and the maintenance and tnsting to be dono on the valvos is the same in either case. Thorn are also no changes to refueling procedures or monitoring capabilities. For other plant equipment important to safety not directly af fected by this use of TS 3.0.4 but which may malfunction due to unrnlated events, the '
radiological consequences of any such. malfunction would be no more severn under TS 3.0.4. Should a release of radioactivn material take place insido the containment during core alterations or the handling of irradiated fuel while under the requirement.s of TS 3.6.4 Action b or c, those requirements provide the necessary isolation for penetrationn with inoperable isolation valves. The safety function of the valves has alrnady been fulfilled by isolating the penetrat. ions. This is trun whether those Actions were taken before or af ter beginning core alterations or thn handling of irradiated fuel.
The Bases for Technical Specification 3.6.4 discusses the i
necessity for the OPERADILITY of thn cont.afnment and drywell faulation valvos to provnnt the relaaso of radioactive matorial to the outsido environment under postulated accident scenarios.
During corn alterat. ions or the handling of irradiated fuel, the accident of concern for this Specification is a fuel handling accident insido containment and the margin of safety of interest as addressed in the UFSAR analysis, is thn margin of 257, of 10CFR100 limits. Inndvertent criticality and other accidents considered during core alterations are cit. hor not possible or have no radiological consequences. Also of concern in the Bases are the closure times of the isolation valves to ensure that any release is terminated in a time f rame cons istent with safety analysis assumptions.
NhS90/SNh1CFLR - 157
Attachment to GNRO-91/00001 N1.5-90-005 pngn 4 tlnder the flex 1hilit y of TS 3.0.4, any penet r at ions with rerpiired ,
but inoperabin, isolntton valves may be isolated in accordance with the requirements of Action b or e prior to beginning core alterntions or the handling of irradinted funl. Taking those actions at. t ha t. time as com tred to taking them nftar core alterations or thn handling of irradinted fuel hav<, begun does not impnet any of the above considerations regarding the margin of safety. Since the isolation is alrendy accomplished the safety funct.!on of the isointion valves is f ul f illed and the penet rnt ion will hn no morn susceptible to nilowing n radionctivn relennn thnn if rnlying on automatic isoint(on. The isointion times nre no longer of concern nnd penet r at ion lenk rates orn unn f f ect eri.
Under this application of TS 3.0.4, the nffec6ed penetrations remain as capable of preventing a relenso as if the necessary isolnLions were taken a f t er beginnleg core alterat.lons or t hn handling of irradinted fuel. Thorn are no chnnges to procedures, controls, or interlocks associated with core nitnrntions or the hnndling of irrndlated fuel which could result in a larger relonne of material insidn thn containment than previously calcuinted.
Thus, the margins of safety described abcyn are not reduced by the flexibilit y of TS 3.0.4 as applied ' a TS 3.6.4, Act. ions b anti c.
NI.S90/SNh1 CPI.R - 158
_ _ _ _ _ _ . . - . _ _ ~ . _ _ _ _ _ _._
Attachment to GNRO-91/00001 i
SRASN: NLS-90-006 DOC NO: TS 3.6.6.2, Actions b 6 c SYSTEM:
i DESCRIPTION OF CllANGE: This evaluation addresses tho safety implication of commencing core alterations and/or the handling of irradiated fuel in the primary or secondary containment with i secondary containment penntrations already isolated by an acceptable method as allowed 1.y TS 3.6.6.2, Actions b or c an i compared to taking thesn actions af t er beginning core alterations or the handling of irradiated fuel. This relief has been previously approved by thn NRC for a limited-timo exception.
REASON FOR CilANGE: During refueling outages, various isolation valves must be modo inoperable to perforn maintenanco, conduct survoillance tests and inspections, or implement design changes.
TS 3.0.4 allows the plant to begin core alterations or the handling of irradinted fuel without having all required isolation valves OPERABLE provided that the requirements of thn applicablo Action Statement are met.
SAFETY EVALUATION: This safety ovalunt.fon concluded that thn chnngo did not involvo an unroviewed safety question. The function of the secononry containment isolat. ion valves and dampers is to inolatn secondary containment penetrations when necessary.
This function, along with t. hat of thn Standby Gas Treatment System (SGTS), ensures that secondary containment integrity assures that j onvironmental releases of radioactivo material are minimized.
- thereby preventing offaite dosos form exceeding t. hose determined by plant. safety analyses. During coro alterations or the handling of irradiated fuel in OPERATIONAL CONDITION 5, all secondary containment isolation valves and dampers are required to be OPERABLE to mitigato radioactive releases which might occur. The UFSAR considers events wnich may potentially result in a radioactive rnleases during refunling while performing core alterations or the handling of irradit.ted fuel. These include inadvertant criticality, failures of various plant. systems and components, loss of of fsite power, and funi handling accidents.
Of these, only the funi handling accident insido primary or secondary containment generates a radiological reinase which results in the nood for isolation of secondary containment penetrations. Should isolation valves or dampnrs become inoperable while performing core alterations or the handling of irradinted funl, Action b or c may be entered to indefinitely provido equivalent level of protection by isolating the affected secondary containment penetrations. Under TS 3.0.4, Act.fon b or c will be taken prior to beginning coro alterations or the handling of irradiated fuel for those penetrations with inoperable isolation valves or dampers. This flexibility har, no affect on t.he methods or equipment used for funl handling or the monitoring and control of refueling activities. The flexibility as applied in TS 3.6.6.2 also does not change or af fect. the number of activition defined as core 'iltnrations. Thnre are no changes to refueling interlocks, so tha probability of an inadvei tent criticality is not increased.
NLS90/S\l.ICTLR - 159
l At t nr limelit to GNuo-91/00001 i
l N t .S - 9 0 - 0 0 f' l'n go 2 For otlici Is i n tit e<lti l l'm " D l I*I" "
g gi i may a l ti t .t ici i diie t o af rect ed I)y this use of , - -
, (s te r ad i o l og i c a l Co"S"'l""""""
'I unreinted events. tiie onwil" > severe under TS 3.0 4-of any such malf unct ions won The margitt of snfety is not rehEi hedthe flexibility of t o Teclinlen) 1ca,nicni si>,..a nceuee ,.0 ^ a on' 3;,ec i f t en t. ion 3. 6. (> . 2.
\
N1.S90/ SNI.lCI't K - 160
Attachment to GNRO-91/00001 SRASN: N!.S-90-007 DOC NO: TS 3.4.9.2, Action a SYSTEM:
DESCRIPTION Of CilANGE: The Safety Evaluation documents the use of TS 3.0.4 for entry into OPERATIONAL CONDITION 4 from OPERATIONAL.
CONDITION 5 whei one loop of HilR (either A or B) shutdown cooling is inoperable while in Action a of GGNS Technical Specificat ion 3.4.9.2.
REASON FOR CIIANGE. During planned refueling outage activities, situations may arise where one shutdown cooling loop ( A or B) is inoperable in order t o per form maintennnce act ivit ies ,
surveillance tests, or design change implementation.
SAFETY EVAhUATION: This snfety evnluntion concluded that the change did not involve nn unreviewed safety question. The UFSAR evaluates several nccidents (events) which are considered t o be rpplicable during GPERATIONA1. CONDIT!ONS 4 and 5. The majority of these events are unrelated to the proposed appliention of TS 3.0.4 for Technical Specification 3.4.9.2 in that their probability of occurrence is unnf fect ed by the shutdown cooling system status or method by which shutdown cooling is provided.
If the alternate method is in service while tensioning the reactor vessel head closure bolts, the prohnbility of a complete loss of shutdown cooling is not increased since an OPERABhE RilR shutdown cooling loop remnins available in standby just an it would under full LCO compliance. I f the nit.crnate method worn to in11, the standby loop could still be placed in operation as described in the UFSAR. This would be the caso during a loss of of fsite power as well, since the OPERABhE shutdown csoling loop is associnted with an OPERABhE diesel generator.
Provided that TS 3.4.9.2 Action n is complied wit h (opernbility of one shutdown cooling loop of RilR with associated diesel generator and nn alternate cooling met hod provided) the flexibility provided by the provisions of TS 3.0.4 does not incronse the probability of the occident previously nvalunted in the UFSAR. This npplication of TS 3.0.4 neither adds or removes systems or components, nor dem it change present syst em des t;;n features or plant opernting stocedures. No new mechanism for draining the reactor vessel is cented. No new or d i f ferent procedures or nquipment nre used for tenstoning the reactor vesseI hend closure bolts.
NhS90/SNhlCFhR - 161
c Attachment. to GNRO-91/00001 e
NLS-90-007 Pagn 2 .
There arn minimal radiological consequences to this event. provided thn appropriatn mitigating actions arn taken. Thesn include ti-establirhing shutdown cooling with alternato means or, if necessary, using Emergency Coro Cooling modos available under TS 3.5.2 to maintain reactor water levnl. Thnsn actions may still bo taken while under the requirements of Action a whether Actton a was entered prior to tensioning the reactor sessel head closurn bolts or af ter having donn so. Furthnr, tensioning thn reactor vossol head closurn bolts has no af fect on the degree of decay ,
heat generation by thn reactor core. Should a complotn loss of !
shutdown cooling occur whiln tensioning the bolts nnder Action a requirement.8, thn consequences would thereforn be no morn severn sincn the amount of decay heat to be removed is unchanged.
Thus, tensionit.g thn rnactor vnssel head closurn bolts while i
alrnady under the provisions of Action a as allowed by TS 3.0.4 has no af fect on thn consequences of accidents previously analyzed.
The Bases fce Specification 3.4.9.2 do not. specifically discuss margins of safnty associated with thn LCO. Discussions of thn ability of only onn shutdown cooling train to provido adnquate decay heat removal capability imply that if a completo loss of shutdown cooling is prevented, there is no negat.tvo impact on plant safety. UFSAR 15.2.9 does discuss the failure of both redundant.kilR shutdown cooling trains resulting in a complete loss j of shutdown cooling event. Ava11abin mitigating actians such as j injection from ECCS provido adnquato cooling to provent. any temperaturn or pressurn transients .in excess of thn critnria for which thn funl, pressurn vnssel, or containment arn designed.
Those actions may be taknn just as rnadily under the application of TS 3.0.4 described in this evaluation. The releasn of
! radianctivity to the environment is thnre fore not. incrnased and l
remains boundnd by more .4evern accidents such as a complete tiSIV closure. This applicat..lon of TS 3.0.4 dans not changn tho l protection against a complete loss of shutdown cooling capability provided by the requirements of Action a of TS 3.4.9.2. Should such an event occur, thn necessary mitigating actions may st ill be taken to provent damaging conditions. Thus, thn margin of sa fety dnfined in the TS Bases is not reduced.
{
NhS90/SNh1CFhR - 162
Attachment to GNRO-91/00001 SRASN: NLS-90-008 DOC NO: TS 3.7.1.1, Action b SYSTEH:
DESCRIPTION OF CllANGE: This Safety Evaluation addresses the application of TS 3.0.4 when either SSW subsystem A or B is inoperabic while entering OPCON 4 from 5.
REASON FOR CilANGE: During refueling outages, situations may arise where one service water subsystem (A or D) is made inoperable in order to perform maintenance or implement design changes. It may also be necessary to change plant. OPERATIONAL CCNDITIONS while in this situation to facilitate other planned outage activities. l Provided that the above individual system LCOs are fully satisfied l via the cooling capability of the remaining OPERABhE service water train, those LCOs do not impact any such changes. If reliance on any Action Statements of these Specifications is necessary, however, further consideration is required with regard to the flexibility allowed by TS 3.0.4 for the OPERATIONAL CONDITION or specified condition change contemplated. The provisions of TS 3.0.4 allow entry into an OPERATIONAL CONDITION or specified condition while complying with t he requirements of an Act.lon ]
Statement only if those requiremants allow continued operation in that situation for an unlinited period of time. Specifically, the case examined in this evaluation is entering OPERATIONAL CONDITION 4 from OPERATIONAL CONDITION 5 by tensioning the reactor vessel head closure bolts with either SSW A or B inoperahin. Each Specification impacted by TS 3.7.1.1 Action b requirements must then be considered with respect t.o this change in OPERATIONAL CONDITION. There is no creation of a possibility for an accident y L
or malfunction of a dif ferent type than any evaluated previously in the Safety Analysis Report. This application of TS 3.0.4 to TS 3.7.1.1, Action b does not change the protection against a complete loss of shutdown cooling capability provided by the requi rements of TS 3.4.9.2, Act ion a. Should such an event occur, the necessary mitigat.ing actions may still be taken to prevent damaging conditions. Thus, the margin of safety defined in the Technical Specification Bases is not reduced.
SAFETY EVALUATION: This safety evaluation concluded that the change did r ot involve an unreviewed safety question. This evaluation only addresses the entry into OPERATIONAL CONDITION 4 while under the requirements of Act ion a of TS 3.4.9.2 as directed by TS 3.7.1.1, Action b. Since all other LCOs directly or indirectly related to service water OPERABILITY in OPERATIONAL CONDITION 4 are sat.isfied, no other Specifications require consideration of TS 3.0.4 provisions reintive to an increase in accident probability. Thus, the question of an increase in probability of occurrence of previously analyzed accidents must only be addressnd relative to TS 3.4.9.2, Action n. The UFSAR evaluates several accidents (events) which are considered to be
! applicable during OPERATIONAL CONDITION 4. Thn majority of these events are unrelated to the proposed application of TS 3.0.4 for Technical Specification 3.4.9.2 in that their probabilit y of occurrence is unaffected by the shutdown cooling system status or i
NLS90/SNhlCFLR - 163
Attachment in GNRO-91/00001 NLS-90-008 Pago 2 method by which shutdown cooling is provided. Corn.nquently the probability of occurrence of these events does not increase.
These are:
- a. Losses or failurns of various plant systems er componenti, (other than f. below)
- b. Inadvertent operation of various plant systems or components
- c. Loss of AC power
- d. Inadvertent crit icality events The UFSAR niso considers two specif,1c events relat.ed to shut.down cooling while shutdown:
- o. Inadvertent. increase in rhutdown cooling. This accident in I only significant near unit criticality and thus does not.
apply for this caso (OPERATIONAL CONDITION 5 to 4),
- f. Loss of shutdown cooling. TS 3.4.9.2 exists to ensurn long term cooling capability while thn reactor is shutdown.
OPERATIONAL CONDITION 4 is entnrod from OPERATIONAL CONDITION 5 by tensioning the reactor vessel head closure bolts. Under >
the ficxibility of TS 3.0.4, TS 3.4.9.2, ActJon a requirements will first. be mot by demonstrating an alternato method of decay heat removal prior to tensioning tho head closure bolts. Tensioning the head closure bolts has no affect on decay heat generation or the alternato method provided. Since the alternate method provides for adequate decay heat removal capability, a completo loss of shutdown cooling is no morn 1ikely under thosn conditions than if the head closure bolts wnre tensioned with both RilR loops OPERABLE and Action a subsequently entered.
Also, if the alternate method is in service while tensioning the reactor vessel head closure bolts, the probability of a compinto loss of shutdown cooling is not. Increased sinco an OPERAI1LE RilR shutdown cooling loop remains available in standby just. as it would under full LCO complianco. If the alternate method worn to
, fall, the st.andby loop could still be placed in operation as l
described in the UFSAR. This would be the caso during a loss of offsito power as well, since the OPERABLE shutdown cooling loop is associated with an OPERABLE diesel generator. This appl (cat.fon of i TS 3.0.4 to TS 3. 7.1.1, Action b dans not. change the protection against a complete loss of shutdown cooling capability provided by the requirement.s of TS 3.4.9.2, Actton a. Should such an event I occur, the nocessary mitigating actions may still be t.aken to prevent damaging cond it.ious. Thus, the margin of safety defined in thn Technical Specification Basns is not reduced.
I i NLS90/SNLICFLR - 164 i
Attachment to GNRO-91/00001 SRASN: N1.S-90-009 DOC NO: TS 3.4.9.2, Action b SYSTEM:
i DESCRIPTION OF CllANGE: This safety Evalunt.lon addresses the applicat ion of TS 3.0.4 to enter OPERATIONAh CONDITION 4 f rom OPERATIONAL. CONDITION 5 while relying upon an n1 ternate reactor coolant circulation method as allowed by Action b of TS 3.4.9.2.
REASON POR CilANCE: Ducing planned refueling outage act ivit ies ,
situnt ions may arise where no RilR shutdown cooling loop or recirculation pump is in operation due to maintenance or surveillance activities, and/or due to use of the RHR system in other designed modes. TS 3.4.9.2, Action b allows an alternate coolont circulation mnthod to be established, niong with nppropriate monit oring to verify proper mixing. 'I t may niso be necessary to change OPERATIONAL. CONDITIONS from 5 to 4 in this situntion to incilitate preparation for pinnt startup. This will result in entering the jurisdiction of TS 3.4.9.2 under the requirements of Action b once the reactor vessel head closure bolts are tensioned.
SAFETY EVAhUATION: This sa fety evaluntion concluded t hat t.he change did not involve an unreviewed safety question. The UFSAR evaluates several accidents (events) which are considered to be applienble during OPERATIONAh CONDITIONS 4 and S. The majority of these events are unrelated to the proposed applicat ion of TS 3.0.4 for Technican Specification 3.4.9.2 in that their prohnbility of occurrence is unaffected by the status or method of reactor coolant. circulation. There are also no changes to any system's design configuration, or operating procedures that would affect these events. Consequently, the probability of occurrence of these events does not increase. These are:
- n. hosses or frtilures of various plant systems or components (other than those discussed separately hnlow)
- b. Inndvertent opernt ion of various plant systems or components
- c. I,oss of AC power
- d. Inndvertent crit italit y events The UFSAR also cons iders some events related to reactor coolant circulation while shutdown:
- e. I ne<!ve rt ent incronse in shutdown cooling. This event is only significant nonr unit criticality and thus does not apply for this case.
NhS90/SNhlCFhR - 165
Attachment to GNRO-91/00001 NLS-90-009 Page 2
- f. Loss of shutdown cooling. OPERATIONAL CONDITION 4 is entered f rom OPERATIONAL CONDITION 5 by t ensioning the reactor vessel hond closure bolts. Adequate shutdown cooling methods are provided by ndherence to TS 3.9.11.2 when lowering reactor cavity level to prepare for head placement and tensioning, and by ndherence to TS 3.4.9.2 a f ter tensioning. The alternate react.or coolant circulation method establishad for Action b of TS 3.4.9.2 may also serve as an npproved decay heat removal method if one RilR shutdown cooling loop is inoperable (TS 3.9.11.2, Action a). llowever, adequate decay heat removal capability is still provided whether it is put into place prior to or af ter head bolt tensioning.
Tensioning the hond bolts has no offeet on decay heat generat ion nor does it result in different methods being used for decay heat removal. A complete loss of shutdown cooling is no more likely while under reliance upon Act. ion b than if OPERATIONA1 CONDITION 4 were entered and Action b subsequently taken. This would also be the caso during a loss of AC power, since an OPERAIRE RilR shutdown cooling loop nssociated with an OPERAIRE diceel generator remains available. (Note that a separate evnluation is required for entry into OPERATIONAL CONDITION 4 under TS 3.4.9.2, Action a requirements.)
- g. Recirculation Loop Pump Trips. Loss of recirculation pump flew results in loss of forced reactor coolant circulation within the vessel. This safety evnluntion already assumes that no recirculation pump is in operation so that the requirements of Action b apply. This event is therefore not applienble to this evaluation. Further, for shutdown conditions, the UFSAR considers n loss of recirculation flow to be within the bounds of norma l operation.
The UFSAR does not specifically consider the loss of reactor coolant circulation while shutdown as an accident or event requiring mitigating nctions to prevent potentla1 radiological consequences. Ilowev e r , the occurrence of such events in the industry has led t o establishing operat ional procedures and testrictions to prevent and mitignte the consequences of a loss of teactor coolant circulation while shutdown. Under Action b requirement.s, an alternate method is established to provide the required circulation, and reactor temperature and piessure monitoring are implemented to verify the effectiveness of the niternate method. The methods available have been shown to provide the necessary mixing and monitoring capabilitv. Whether alternate reactor coolant circulation methods and system monitoring are put into pince prior to t r nsioning the reactor vessel head closu e bolts under Action b, or af t er doing so has no af f ect on the likelihood of n loss of coolant circulation. Tbc alternate method put into place is the snme in eit her ense, NbS90/SNblCFbR - 166
i Attachment to GNRO-91/00001 1
NhS-90-009 Page 3 and once the vessel head is positioned, the tensioning procedure has no affect on reactor coolant circulation or the ability to monitor temperature and pressure.
The application of TS 3.0.4 also does not increase thn probability of a vessel draindown, although this event is not specifically considered in the UFSAR while shutdown. Thn alternato reactor coolant circulation method will be the same regardless of the timing of its implementation, and the t ensioning of the reactor vessel head closure bolts has no nf fect on draindown potent ial.
The consequences of other events related to coolant circulation will also be no more severn under this proposed application of TS 3.0.4. The same mitigating actions may st ill be taken as described in thn UFSAR for a loss of shutdown cooling. These includn re-establishing shutdown cooling with alternate means or, if necessary, using Emergency Core Cooling modes available under TS 3.5.2.
Tensioning the reactor vessel head closure bolts to enter OPERATIONAh CONDITION 4 while under TS 3.4.9.2, Action b has no affect on the availability of these methods or the amount of decay heat generated. Should a complete loss of shutdown cooling occur, the consequences would therefore be no more sovern.
The equipment important to safety under consideration for this evaluation involves systems or components associated with alternate reactor coolant circulation methods, or equipment which enay hn affected by failure t.o provido adequate coolant circulation. Since Action b requirements of TS 3.4.9.2 provide for sufficient circulation and monitoring of vessel condition, the probability of a malfunction of such equipment remains essentially the same whether Action b is entered after tension.ing of the reactor vessel head closure bolts or prior to tensioning. The raet hod and equipment used for tensioning arn also no different in either case, so the t.ensioning procedure would not nffect equipment important to unfety in a different way.
The llFSAR also discusses the possible harmful effects of thermal stratification on the vessel during periods of low coolant circulation. himits arn placed on recirculation pump restart during such conditions to prevent thermally induced stresses in excess of vnssel design. These limitations are unchanged by this application of TS 3.0.4, therefore the poss ibility of vessel dnmage due to thermal stratification !
from a loss of renctor coolant circulation is also not l increased.
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NhS90/SNh!CFhR - 167
- .-..--- .- . - . - . . - - . . - - - - - - . - . . ~ - - . . - . - - - . . ~ . - - - - .-
l Attachment to GNRO-91/00001 ;
NLS-90-009 Pago 4 l
Tha Bases for Specificat.fon 3.4.9.2 do not specifically discuss margins of. safety associated with the LCO.
Discussions of the nbility of only one shutdown cooling train j to provido adequato coolant miving imply that if a completo '
loss of coolant circulat.fon zor a time period sufficient to induce thermal stratification is prevented, there is no negativo impact on plant safety. Other TS Bases discuss the need to provent thormal stratification in order to assure accurate temperaturn indication and allow for proper mixing of neutron poison solution should it be needed, as well as to provent unduo thermal stresses on the vossol when mixing is reestablished.
Establishing an attornato coolant circulation method un<ler TS 3.4.9.2, Action b requirements prior to tensioning the l reactor vessel head closure bolts as opposed to af ter doing so does not negatively impact the ability to prevent thermal stratification. The time to reestablish coolant-circulation is not increased. Availabic coolant circulation methods are no different and monitoring instruments are unaffected.
, Tensioning the head closuro bolts to enter OPERATIONAh CONDITION 4 does not affect coolant circulation or the actions to bo taken in the event circulation is lost. Also, the 1. imitations imposed to provent thermal stresses when restarting a recirculation pump are unchanged should this be the method selected to reestablish coolant circulation. This application of TS 3.0.4 thus does not reduce the margin of safety as defined in the Technical Specification Bases.
1 NhS90/SNh1CFhR - 168 l
Attachment to GNRO-91/00001 SRASN: NI.S-90-010 DOC NO: TS 3.6.4, Actions b I, c SYSTEM:
DESCRIPTION OF CilANGi;; This evaluation addrnsses the safety implication of commencing opnrat ions with a lut ent in t for draining the t eactor vessel (OPDRVs) during OPl; RATIONAL, CONDITION 4 or 5 with contninment and/or drywell penet rat ions nit endy isolated by an accept able method as allowed by TS 3.6.4 Act ion b or c as compared to t aking t hnse actions o f ter beginning OPDRVs. rhis relief has been previously approved by thn NRC for a limited-time exceptlon.
REASON FOR CllANGE: During refueling outages, various isolntion valvns must be made inoperable to perform maintenance, conduct surveillance tests nnd inspections, or implement design changes.
TS 3.0.4 nilows the plant. to begin OPDRVs without having all required isolnt ton valves Ol'ERAll!.E provided thnt thn requirements of thn applicabin Act ion Statements are mot .
SAFETY EVAI.UATION: This safety evaluntion concluded that the 1
chango d id not. involve nn unreviewed safety question. The function of the containment and drywnll isolation valves is to ensurn that drywnll and containment penetrations are isolated in the event of a rndionctive rnlense insidn the containment. This assures that nn environmental relense of radianctive materini is cont rolled to within the design leakage rate of thn containment systems, thereby preventing offsite doses from exceeding those determined by pinut safety analyses.
During OPDRVs in OPERATIONAI. CONDITIONS 4 or 5, cert ain containment and drywell isolntion valves (Groups 5, 6A. 6B , 7, 8,
- 10) nre tequired to be OPERAll!.E as speci f ied in TS 3.3.2 to mitigate radioactivn releases which might occur. The UFSAR considers events which may potentially result in a radionctIvo release during shutdown and refueling. These include inndvertent crit (cality, failures of various plant systems nnd components, loss of offsito power, and fuel handling accidents. Of thran, only the fuel handling accident insido containment gennratos a rndlological release which results in thn need for automatin isolation of containment and drywell penetrations. The UFSAR does not specifically consider a vessel draindown while shu' lown .
Should isolat ion valvns becomo inopornble while performing OPDRVs, Act ion b or c may be entered to inde f initely provide an equivalent levn1 of protection by isointing thn affocted penetrations. Under TS 3.0.4, Action b or c will be taken prior to beginning OPDRVs for those penetrations vith inoperable isolntion valves. Thn flexibilit y of TS 3.0.4 as npplied in TS 3.6.4 also does n o t.
chnngn or nfrect thn number of act ivit ies dnf ined as OPDRVs.
There are no chnngns in refueling interlocks, so the prohnbility of an inndvertent crit icality is not increased.
NT.S90/ SNI lCFl.R - 169
1 Attnchment to GNRO-91/00001 NhS-90-010 pagn 2 1solating any penetrations having inoperable isolation valvns before beginning OPDRVs completely fulfills the safety f unct ion of the valves. The radiological consequences of n fuel handling accident will thus be no worsn than nnnlyzed. Also, none of the analyzed accident sequences orn changed by isolat ing thn a f fected penet rat ions prior to beginning OPDRVs rather t hnn at somn later time. Puol handling techniques and equipment nrn not a lt ered, monit oring and cont rol methods are not modi fied , nor n ro t he types of activities defined ns OPDRVs changed, Refueling interlocks remain unchanged. No radioact ive mat er ial r elease mechanism or path is created where nono provinusly existed. Exercising the provis ions of TS 3.0.4 (n this caso maintnins the plant in nu necept.nbly sa fn cond it ion reintive to thn radiological consequences of potentint accidents during OPDRVs.
This application of TS 3.0.4 may directly affect equipment importnnt to safety in two ways. Firstly, the isolation valves and penetratJons themseIves wi11 be affected dun to the acquirement to close and/or donct ivat e volves or af fix blind fInngns in order t o isolat e penetrat(ons. Secondly, systems and equipment served by t he penet rat ions may also be nffected dun to thn blocking of various flow paths. Knfunling equipment and other pinnt components are not impacted by this use o f TS 3.0.4.
The contn inment/drywnll penetrat ions will hn isolat ed under the provisions of TS 3.6.4, Action b or c in thn event that their isolation valves are made tuopornble for outngn activities prior to or while OPDRVs were underway. There is no additional offect on thn valves and penet rat ions thnmselves as a result of performing thn isolat.f on prior to beginning OPDRVn. Thn mothmt l
accomplishing the required isolation is ident ica l in either caso, _
' nad the maintenance or testing of thn valves or penetrattons will anno be unchnnged. Similarly, systems whose flow paths nie alt ered as a result of isolated penet rnt lons wi11 be impncted in the same manner regntdless of thn t iming of t he Actions.
Mni f unct ions of equipmeni important to sa fet y have been cons idered in the UFSAR for plant- conditions associnted with shutdown and rnfueling.
The equipment which couhl be a f fected by this applicnt ion of TS 3.0.4 includes thn isolntlon vnIves and penet rat ions ndd ressed in Speci f icat ion 3.6.4, as wel l as the systems and components snrved by these penetrations. The possibility of failurn of isolnLion vnives is considered in that redundant isolation capability is provided. While under the requiremnnts of Action b or c, this prot ect ion is preserved beenuse nny penetrations having inoperable isolation valves will already be isolat ed prior to the initiation of OPDRVs. Performing -
this nct ion prior t o beginning OPDRVs rather thnu a f ter os allowed under TS 3.0.4 dons not cronie the opport unit y for a new or diffnrent type of mal f unct ion of t he isolation vnives.
NbS90/SNLICFl.k - 170
I Attachment to GNRO-91/00001 N1.S-90-010 Pago 3 l The Danes for Technical Specification 3.6.4 discusses the necesnity for the OPERAD11.lTY of tho containment and drywell f r olnLion valves to provent the rolcase of radioactivo materini to the outsido environment ender postulated accident scenarios.
During OPDRVs, the accident of concern for t.his Specification is a j funt handling accident insido containment. and the margin of safety of Interest, as addressed in the UFSAR analysis, is the margin to 25% of 10CFR100 limits. Inndvertout criticality and other accidents considered during shutdown and refueling are either not possiblo or havo no radiological consequences. Thn UFSAR dons not specifically address vossel draindown ovents while shutdown. Also of concern in thn Bases are the closure times of the isointion valves to ensurn that any release is terminated in a t.ima frama consistent with safety analysis assumpt.fons.
i Under the finxibility of TS 3.0.4, any penet.rntions with required, but inoperable, isolntion valvns may be isointed in accordanco with the requirements of Act.fon b or c prior to beginning OPDRVs.
Taking these actions at. that Limo na compared to taking them after OPDRVs have begun does not impact any of the abovo considerations regarding thn margin of safety.
I i
l l
NT,890/SNI,ICFI,R - 171
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Attachment to GNRO-91/00001 l
l SWASN: NI.S-90-011 DOC NO: TS 3.6.6.2, Actions b & c SYSTEM:
DESCRIPTION OF CllANGE: This evaluation addresses the safety implication of commencing operations with a potentini for draining the renctor vessel (OPDRVs) during OPERATIONAL CONDITION 4 or 5 wit h containment and/or drywell penetrations already isolated by an acceptable method as nilowed by TS 3.6.4 Action b or c as compared to taking these act ions a f t er beginning OPDRVs. This relief has been previously approved by the NRC for a limited-time exception.
REASON FOR CllANGE: Technical Specification 3.6.6.2 identifies operability requirements for secondary containment automat ic isolation valves and dampers in OPERATIONAI, CONDITIONS 1, 2, and 3 and at other times including during operations wl.th a potential for draining the react or vessel (OPDRVs). During refueling outages, va r.l ous isolation valves must be made inoperable to perform maint.cnonce, conduct surveillance tests and inspections, or implement design changes. TS 3.0.4 allows the pinnt. to begin OPDRVs without having all required isolation valves OPERAI11.E provided that the requirements of the npplicable Action Statements are met.
SAFETY EVAI.UATION: This safety evaluation concluded that the change did not involve an unreviewed safety question. The function of the secondary containment isolation valves and dampers is to isolate secondary containment penetrations when necessary.
This function, along with that of the Standby Gas Treat. ment System (SGTS), ensures that. secondary containment integrity is maintained when required. Secondary cont 11nment integrity assures that environmental releases of radioactive material are minimized, thereby preventing offsite doses from exceeding those determined by plant safety analyses.
During OPDRVs in OPERATIONAI, CONDITIONS 4 or 5, nll secondary containment isolation valves and dampers are required to bn OPERA 111.E t o mitigate radioactive releases which might occur. The UFSAR considern events which may potentially result in a radioactive release during shutdown and refueling. Thesn include inadvertent criticality, failures of various plant systems and components, loss of offsite power, and fuel handling accidents.
Of these, only the fuel hanlling accident inside primary or secondary containment generates n radiological release which results in the need for isolation of secondary containment penetrations. The UFSAR does not specificn1ly consider a vo w l draindown while shutdown.
Should isolation valves or dampers become inoperable while performing OPDRVs, Action b or c may bn entered to indefinitely prov.ide an equival nt. level of protection by isolating the affected secondary containment penetrations. Under TS 3.0.4, Action b or c will be taken prior to beginning OPDRVs for those penet rat ions wit h inoperable isolation valves or dampers. This N1.S90/ SNLICFI.R - 172
Attachment to GNRO-91/00001 NhS-90-011 Page 2 flexibility has no affect on the methods or equipment used for fuel handling or the monitoring and control of refueling activities. The flexibility as applied in TS 3.6.6.2 also does not change or affect the number of activities defined as OPDRVs.
There are no changes to refueling interlocks, so the probability of an inadvertent criticality is not incrnased. The SGTS is anaffected and remains able to provide its mitigating function.
Exercising the provisions of TS 3.0.4 in this case maintains the plant in an acceptable safe condition relative to the radiological consequences of potantial accidents during OPDRVs.
This application of TS 3.0.4 may a f fect. equipment important to sa fety in two ways. Pirstly, the isolation valves, dampers and penet rations themselves will be directly a f fect ed dun to the requirement. to close and/or donctivate valves / dampers or af fix blind flanges in order to isolate penetrations. Secondly, systems and equipment served by the penetrations may also be affected due to the blocking of various flow paths. Refueling equipment and other plant components are not directly impacted by this use of TS 3.0.4.
The Bases for Technical Specification Section 3.6.6 discusses the necessity for the OPERAB1hlTY of the secondary containment isolation valves and dampers to prevent the release of radioactive material to the outside environment under postulated accident scenarios. During OPDRVs, the accident of concern for this Specification is a fuel handling accident inside primary or secondary containment and the margin of safety of interest, as addressed in the UFSAR analysis, is the margin to 25% of 10CFR100 limits. This is derived f rom the interpretation of the NRC Standard Review Plan. Inadvertent criticality and other accidents considered during shutdown and refueling are either not possible or have no radiological consequences. The UFSAR does not speci fically address vessel draindown events while shutdown. Also of concern in the Bases are the closure times of the isolation valves and dampers to ensure that any release is terminated in a time frame consistent with safety analycis assumptions.
Under this apnlication of TS 3.0.4, the af fected penetrations remain capable of preventing a release as if the necessary isolations were taken after beginning OPDRVs. There are also no changes to procedores, controls, or interlocks associated with OPDRVn that could result in a larger release of material inside the primary or secondary containment than previously calculated.
Thus, the margins of safety 6 scribed above are not reduced by the flexibility of TS 3.0.4 as applied to TS 3.6.6.2.
i I
NhS90/SNhlCFhR - 173
t j Attachment to GNRO 41/00001 i f
SRASN: N1.S-90-012 DOC NO: CR NI.-90-009 SYSTEM:
t DP. SCRIPT 10N OF CllANGE: This l'FSAR change rovined the outngo data for the of f atto 500 kV t ransmission linen. A decrene. (0.90 to -
0.94 outages / year /100 milesi in the overall pertorm- -
of the 500 1 kV nystem was renifred. Also tho dat.1 was changed for the 115 kV transminnion line between Natchez SES and linator Wilson SES and to i GGNS. The ll5kV transm hsim linn han experienced an overall {
- outngo rate of 1.79 outagen/ year /100 miles compared to an overall l 0 ratn of 1.65 currantly in the UTSAR. ;
P REASON FOR CilANGE: % 1 inn outage data in revised annutisey 1o updnto the UFSAR. No physical changen to the transmission linen !
were made under this evaluation. i SAFETV LVAl.UATION: Thin fn #ety ovaluat ion concluded t hat the chango did not. Involve an unreviewed safet.y quent lon. The changes
- to the UFSAR consist of revintons to HP&l, transminnion linn outago ,
data. Information on outages for the period f rom June 1. 1989 to Hay 31. 19c0 was added to Chapter 8. Statistics on the ,
transmission line outage rate are routinely updated based on the new data. In addition. data for previous years in corrected based [
on informntion received from HP61,. The chnnges to the UrSAR j reflect the actual performanco of the HPSI. transmission system.
~
No physical change io GGNS, the opetation of GGNS, or the tbree offaita power sources han occusaed.
The outage rain han . increased slightly over the values currently in the UFSAR. .UFSAR Section 15.2.2 identiffns grid disturbnnces that cauno closure of the turbino control valves an events of I moderate frequency (1 to 0.05 events per year). The slight increase in outage rate does not change the clansification as an i ovent of moderate frequency. Thus the Chapter 15. analysis in not affected. The probshility or consequencer. of an accident. or ,
malfunction in the Chapter 15 analysis are not changed. t A loss of all grid connections has been annlyzed in the UTSAR. No change to the plant design or operation are being made so no possibility of an nccident or malfunction dif ferent than prnviously evaluated is created.
There is no reduction in the margin of snfety es defined in the i' basis for any Technical Specificatico. The action requirements specified in the Technical Specifications assumo a loss of offsitn power and nrn intended to provide assurance that a lons of offsite 4
power will not. result in a complete loss of safety function of critical syste:ns.
NI,S90/SNLICFI,R - 174
At tachment to GNRO 91/0000) i 1
SRASN NLS 90-013 DOC NO: TS 3.9.11.2. Action a SYSTEM: j DESCRIPTION OF CllANGE: This safety evaluntion addrennen the uno or Th 1.0.4 to ent er tha APPLICA!!!LITY of LCO 3.9.11.2 by i l r w..nioning the reactor prosaurn vessel head closurn bolta and e nt ering OPERATIONAL CONDITION $ f rom 4 while complying wit h ;
Act(on Statement n. I REATON FOR CllANGE: Should one or more of the required shutdown {
cooling systems become inoperable, ACTION n allown tho plant to i remain in this condition indefinitely provided that an OPERAltl.E alternate method of decay heat removal is madn availchio for the j
- inoperablo system. During planned outage activities, situations i may arian whero one or morn shutdown cooling systems are inoperabin in order to parform maintenance activition,
, surveillanco test s, or design chango impicmentation. This
, condition is allowed by TS 3.9.11.2; however, such situationn requirn compliance with Action a of TS 3.9.11.2 to ensure adequate plant protection while the reactor cavity water level is less than j 22 fnet 8 inches in OPERATIONAL CONDITION 5. '
SAFETY EVALUATION: This safety cynluation concluded that the chango did not. involyn an unreviewed safety question. The UFSAR evaluates severni accidents (events) which are considered to bc applica.ilo during OPERATIONAL CONDITION 5. The majority of theso events are unrelated to the proposed application of TS 3.0.4 for i Specification 3.9.11.2 in that their probability of occurrenco in i
. unaf fect ed by the status of shutdown coaling or the mechanism by y which shutdown coolfog in bning provided. Thus, the probability !
of occurrence of thenn events is not increased. Those are: ,
- n. Lone of plant instrument air l
- b. F..nl loading and handling orrors ;
i
- c. Radwnste system malfunctions
- d. Rod withdrawal errors
- c. Inndvert ent pump st art s (llPCS, Rec i ce. . )
- f. Tendwater cont roller failurn
- g. Loss of onsite or o f m it o AC pow er l Thn UFSAR also considnts two specific events related to shutdcwn cooling while shutdown:
- h. Inndvertent increnso in Shaldown Cooling - moderator :
temperature decronno i.
L NLS90/SNLICFLR - 172
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4 Attachment te GNRO-91/00001 NLS-90 013 Page 2 !
r As stated it the UFSAN (Reference $) this accident in only of I concern during startup or cooldown near critical which is not !
I the case in this situntfon.
+
1
] 1. boss of Shutdown Cooling (References 6 and 9)
Even though ovent (i) is not considered a Design Basis Accident, !
the Technient Speciffention LCOs (including 3.9.11.2) are provided 1
to maintain the probability and conscrpiences of such previously ,
evnlunted events conNINtent With analyses by ensuring that equipment and systemw assumed in the analyses remain operable. [
When 78 3.9.11.2 4; not met due to one Rilk shutdown cooling mode i train and ADilRR neing inoperable, Action a allows continued operation for e.n unlimited period of time by providing for an ,
alt ernat e method capabic iJ decay heat- r emoval . This alternate ;
method providen protection in the event the remaining RIIR trnin also becomes inoperable.
t Under the provisions of TS 3.0.4, detensioning the reactor vessel f head closure bolts is allowed provided an n1 ternate method of '
, doeny hent. removal has been demonstrated. The method of detensioning the clnsure bolts does not change whether r detensioning is being performed whilet
- n. n1 ready under reliance of an alternate method of decay heat !
removal in accordance with Action n; or,
- b. in full compliance with the 1,00 and subsequently entering ;
Action a once the reactor vessel head closure bolts are i detensioned.
Also, if the n1 ternate method is in service while detensioning the !
- closure bolt s, the probability of a complete loss of shutdown cooling is not increased since nn OPERABLE RilR shutdown cooling train remains nynlinbic in standby as it would under full LCO ;
complinnee. If the alternate method were to in11, the standby RIIR ~
train could still be pinced in operation as described in the UFSAR. This would be the caso during n loss of offsite power as well, since the OPERABLE RilR shutdown cooling train is associnted with an OPERABl.E diesel generator.
The likelihood of a complete loss of shutdown cooling may in fact be decreased by taking steps to demonstrate an n1 ternate met hod j prior to beginning closure bolt detensioning. Without the j flexibility of TS 3.0.4, outage activities would have to be interrupted to make the affected shutdowr. cooling system operable ,
prior to bolt detensioning. After a closure bolt is detensioned.
the RilR loop and/or ADliRS would again be made inoperable and Action n entered. Should the remaining OPERAHLE RilR shutdown cooling loop fall before the alternate method has been adequately '
demonstrated, the time to provide niternate cooling is less since no method of shutdown cooling rer as OPERADLE. E i
NLS90/SNLICFLR - 176
-m-.,, .i%e~-emer-%-,---#-,-e.e-.-,- ,,--r----r- ----+-e-------,-,-4-.,-,,y--.,w ey,,-,-vw-y--vv,,,,v-,--m,-,yw+-,-ycw,,-,v-r-r-m- , ya < y v -s*e
Attachment to GNRO-91/00001 i
l NLS-90-013 pege 3 Thin applicat ion of TS 3.'t.4 also doen not affect the potential !
for draining the onctor vennel sinco no procedures are changed or !
equipreent. modj f f ed, although vessel draindown in not specifically 1 addressed in t he UFSAR for refueling. ;
4 l An stated nbove, the only UPSAR annlyzed accident requirint i
considerat ton for thin application of TS 3.0.4 in a loss of !
, shutdown cooling during refueling. There are no indiological i consequenenn to this event provided the appropriate mit ignt ing !
actions are t air e n . Theno include rn establishing shutdown cooling ;
I with alt ernate mennn or, if necessary, using Emergency Coro !
Cooling modes nynilable under TS 3.5.2 t o maint ain reactor water )
, level. Theno actions may st ill bn taken while under the i requirement s of TS 3.9.11.2, Act ion a whet her Action a was entored prior to closure bolt dntensioning or af ter. Further, closure f bolt detensioning has no af fect on the degree of decay heat gennration by the reactor core. Should a completo lons of ;
shutdown cooling occur whiln under Action a requirements, thn !
consequences would therefore be no more severn since the amount of l decay heat to be removed is unchanged. '
Tho equipment important to safety under consideration for this evaluntion involven systems or components associated with EllR or the niternate methods of decay heat removal, or equiprnent which
, may be af fected by failure to provido adequatn shutdown cooling while refueling. Since the Action a requirements of TS 3.9.11.2 provido nufficient shutdown cooling capability, the probability of a malfunction of such equipment remains essentially the samo l whethnr Action a is entered af ter closure bolt detensioning at prior t o closure bolt det ensioning. Thn method of bolt detension.ing in the same in either cano, so this process would also not affect equipment important to safnty in a different way.
Technical Specificatio. 3.9.11.1 and 3.9.11.2 Bases discuss the icquirement for the RPR shutdown cooling system and ADilRS to provide suf ficion; cooling capability t o remove decay heat to maintain the nyerage reactor coolant temperature below 140*F. The TS 3.9.11.2 requirement to have RilR nhutdown cooling trains and/or -
ADllRS OPERABLE when reactor cavity water level .in less than 22 -
feet 8 inchen ensures that a comp 1nto loss of shutdown cooling ;
, capability will not occur. By demountrat ing thn OPERAlilLITY of an npproved niternato decay heat removal method should onn of the ,
, required RilR shutdown cooling trains or ADilRS becomo inoperabic per Action a, cdequatn shutdown cooling capability is maintained. '
r l
NLS90/SN1.lCFLR - 177
Attnchnent t o GNKO '11/00001 N 1.S - 9 0 - 013 l'nge 4 linito r TS 3. 0. 4, t he Ol'r.K Allli,l TY o f the a t t ei nnt e mnt ho<l i s dernonst rnt eil prior t o det ensionitig t he closure bolts, as opposed to detennioning the hond closure bolt s and subnequent ly demonst rat ing it s OPlikAltll.lTY with on RitR loop niut ADilRS inoperable. Both sit unt ions provide essentinlly equivalent denny hent t ernovn l su f ficient t o rnnint n in t he nyernge reactor water t r+perat ure bnlow 140*P. In fact, under reliance on TS 3.0.4, having de monst rated the nynileihility of a bnckup decay heat removnl method before bolt detensioning may reduce the amount of t ime necessary to pince such a syst em in set vice.
Thus, t he proposed nppliont ton of TS 3.0. '. does not decrease the margin of safety as discessed in the Technic nl Specificnt ton Bases.
Nl.S40/SNLICFLR - 178
l Attachment to GNRO-91/00001 l
$RASN: NhS-90 015 DOC NO: CR-Nh-90-006 SYSTE!it I
DESCRIPTION OF CilANGE: These changes in the arens of organir.ntion, communientions and related fulds wnre mnde over tho l past year. These changes affected senior manngerrent down to plant !
staff management.
REASON FOR CilANCE: These changes updated the UFSAR. {
SAFETY EVAL.UAT10N: This safety cynluntion concluded that the chango did not involyn an unroviewed safety question. The i administrat ivo changes are int ended to reflect the current !
structure of both onsite and offsito organizations, which are not ,
being handled /nddressed by other Change Requests. Each I substantive chango ik designed to consolidnto and strengthen i manngement and administrativo functions, and provido a more l cffective manngement chnin of-command. Individuals nasigned t o any newly created positions arn t equired to moet the qualifications specified in the UFSAR.
The chang 9s to tbn communication system reflect recent trans' ors in ownership and equipment upgrades, which are der.igned to enhanco system coverngo and relinhility. Thoro are no general design !
criterin or regu!ntory guides that directly apply to the safety-related performance requirements for thn design and use of ,
the communication system during normal plant operations and !
translent conditions. i The organizational changes havo no impact on the mntgin of snfoty ;
due to their administrative nature. Qualification requirements ,
for newly created posit ions meet 'ny ne:.ilenhle standards i represented in UFSAR Section 13.1.3. The changes to the cornmuntention system reflect equipment upgrades, which aro designed to enhanco system coverngo and relinhility. Thoro are no 4 general design crit erin or regulatory guides that directly apply
, to thn snfety-related performance requirements for the design and use of the communication system during normal plant operatione. and trnunfont conditions.
P NhS90/SNhlCFl.R - 179 l
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Attachment to GNRO-91/00001 l I
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SRASN: N!,S-90-016 DOC NO: TS POS STMT 128, Rev. O SYSTEM: !
! l DESCRIPTION Of CilANGE: This safety evaluation documents the l 1 ovalunt ion of the ef fect of the Division 11 ESP Switchgont room :
coolnrs being out of snrvice during a refueling outage with !
temporary ventilation provided to maintain temperature below tho l Technical Specificatton (TS) limit of 104 degrees P.
- r REASON FOR CilANGF.: To allow for flexibility in outngo work that i requires the Division II ESP Switchgear room coolors to bn removed j from snrvice. ;
i SAFETY EVA!.UATION: This snfoty evaluntion concituted t hat the !
chango did not involvo an unroviewed sn fot y quest ion. The altnrnato motbod of room cooling provides air flow rates to maintain room temporntures below 104'r. This evaluntion considort.
that the Div 11 ECCS pumps and EDG 12 will not. be required during this timo and that offsito pcwor is nynilable. Since Div 1 or 111 [
ECCS will be nyallable to meet Technical Specification ECCS L requirements, the relevant concern is maintaining power dintribution nynilable for opernt ton of Div 11 Primnry Containment ;
Isolat.fon Valves. An additfonal conenrn is to ensurn that a loss [
of r.hutdown cooling does not occur due to a failure of equipment in an ESP Switchgent room causing an inndvertent isolation of tho ,
Div 11 SDC isolation valve. ;
The failurn of thn altoinnte cooling equipment does not incronso the likolfhood of the occurrence of loss of SDC. The maximum temperature cniculated to occur with no room cooling available is less than the calculated temperaturo at which the limit ing equipment failure k.ll occur.
UTSAP sect ion 15. A.6.3.3 assumes t hat hPC1, LPCS, or lipCS can bn used with the reactor vossol head off in the event of n loss of SDC. If the reactor vessel hond .is on and the system can be pressurized, the ADS or manual operation of relief valves in :
conjunct ion with any of the ECCS and the RilR suppression pool ,
, cooling modo can be used to maintain water level and removn decay s hent. ECCS Div I and 111 will not b, affected by Div 11 ESP swit chgenr room tempornture. Thornfore, the consequences of a postulated loss of FDC nro not increnned. !
The nit ernato room cooling is provided to maintnin Div 11 primary contninment isointion valves opernble while thn redundant Div 1 isolntion valves arn also operable. Tho alternate coolinpt !
equipment will be connected to BOP power supply and therefore will ;
havn no effect on power supplies important to safety. '
- i NhS90/SNb1CI'IR - 180 -
l Attnchment.to CNRO 91/00001 i
N1,S-90-016 pagn 2 l I
The proposed activity does not crente the possibility of a !
l malfunction of a dif ferent typo ninen only a ningle t ra tti fatturo !
would occur an a renuit of exccanivo temperaturen in the Dfy 11 F.SF Switchgent room. The flow rato requirernents for the n1 ternate room cooling equipment are sufficient to maint ain room temperaturn -
below the environmental qualification TS limit of 104*F. Thn l possibility of a malfunct ion due to excessive ternperaturen is not likely due to the maximum cniculated temperatures rise with no ;
toom cooling hoing lenn than that at which thn limiting equipment '
fn11ure would occur.
r Thern is no reduction in the margin of safety an defined in the basis for any Technical Specification. The Technical Specification 1%it of 104'T is establishen by TS 3.7.8 to ensure
{ that the temperaturo remainn below the environmental qualification temperatures of the equipment located in the r.SP switchgear rooms.
The n1 ternate method of room cooling establishen sufficient air flow to maintain room temperatures bnlow this limit during this period of time when Div II heat loads are minimal.
i The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> survellinnen requirement of Technical Specification E 4.7.8 will be maintained to confirm t.hnt the temperature in the T.SF Switi.hacar rooms remains lenn than the 104*r TS limit, i i !
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, NhS90/0Nh1CPhR - 181 l
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Atinchmt+t to GNRO-91/00001 l l
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SRASN: N!.S90-017 DOC Not TSPS 128 R00 SYSTEM:
i f
DESCRIPTION OF CilANGE; This safety evaluntion documents t he ;
evaluation of the ef fect of the Div 1 ESP switchgonr room coolers being out of service nuring a refueling outage (Operational Conditionn 4 or 5) with temporary ventilation ptovided to mnintain temperaturn hnlow the Technient Specification (TS) limit of 104*r.
Previous tests and calculations have shewn that equipment in those .
rooms will remnin functfonally capable of performing the safety functions at temperatures well in excess of 104*F l
REASON FOR CllANGE: fo allow the Div. 1 ESP switchgent room !
coolnts to be out of servico dur ing a refueling outage with tempornry ventilation provided.
SAFETY EVA!,tlATION: Thin safety evaluntIon concluded thPt tho ;
chango did not involvo an unreviewed safety question. Thn '
alternate method of room cooling provides air flow intna to r maintain room tempornturns below 104*P. This cynluation takes ,
into cons.ideration that the Div 1 ECCS pumps and EDO 11 will not '
be required during this timo and that offsito power is nynilable.
Sinco Div 11 or 111 ECCS will he nyntinble to meet Tech Spec ECCS requirements, the concern is mnintnining power dist ribution
- avnlinble' for opernt loa of Div 1 pr imary containment isolation valves. An addit.lonal concern is to ensuro that a loss of ,
shutdown cooling does not occur dun to a fnilure of equipment in l
- an ESF switchgear room causing an inndvertent incintfon of the Div !
1 SDC isointion valve. ,
l Although the n1 ternate method of room cooling in not designed to ,
the namn rnquirements of the ESP switchgear room cooling i equipment, the (niluro of thn alternato cooling equipment does not increase the 11ko11 hood of the occurrence of loss of SDC. The maxituum temperaturo calculated to occur with no room cooling nynilable is less (Lin thn cniculnted tempernture nt which the ;
limit ing equipment. failure will occur. <
UFSAR Sectton 15.A.6.3.3 assumes that 1.PCI, LPCS, or HPCS can bc used with t he reactor vessel head of f lo the event of a loss of
! SDC. If the reactor vessel hond is on and the system can ho pressurized, the ADS or manual operation of rnlief valves in conjunction wit h any of the ECCS and the RHR suppression pool cooling modo can bn used to maintain water level and remove decar hnnt. ECCS Div 11 and 111 will not be affected by Div i ESP switchgent room temperaturo. Thereforo, the consequences of a postulated loss of SDC nrn not increased. primnry contninment isolation system is designed to ho single failurn proof and the redundant Div 11 equ.ipmont will ba operable when the alternato room cooling is provided tc thn Div 1 ESF switchgear room.
NLS90/SNhlCFLR '.82 n'--%w,r y - . ,.-- ---,,..y-- -,-,..,-.3 w,m.,,. _,.y , ,, e, e-.,- - ..----- w. %-,e-.--%,.,.,.,,-.---m..... e----,---._w- ~ -,w-,. -r,, .. - , -
i Attachment to GNRO-91/00001 i
i NhS-90-017 ;
Page 2 '
The alternatn room cooling is provided to maintnin Div I primary !
containment isolation valves operable while the redundant Div II
, isointion valves are alno opornble. Thn niternato cooling ,
equipment will bn co inocted to it0P power supply and therefore will l have no affect on pownr supplies important to snfoty and does not creato a possibility for an accident or malfunction of a dif f erent .
typn than any evalunted previously in the Snfnty Analysis Rsport.
There is no reduct ion in the margin of safety as defined in the
. basis for any lechnical Specification. T'ie alternato method of room cooling for ihn Div 1 PSP switchgent room is intended to :
maintain the room temperaturn below the limit of 104*r established h) Technical Specification 3.7.8. Thn Technien! Specificntions do not address opernbility requirements of the Div I ESP switchgenr .
room coolere nor specify the method of maintaining room temperaturn below the 104'r TS limit. finintaining thn temperaturo limit halow 104*P casures that the environmental qualification limits are not. exceeded. Technical Specifications requirn the temperaturn in Div 1 EST switchgear rooms to be determined to be i within the 104'F limit a t. least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the ,
equipment in the rm ms are required to bo operable. This !
surveillance will cont inne with the room coolers out of service.
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i l NhS90/SNI,1CFLR - 18'i 1 ,
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. Attachment to GNRO-91/00001 SRASN: NLS-90-018 DOC NO: UTSAR CR-Nir90-014 SYSTEM: l DESCRIPTION pF CllANGE: The USAR change adds e descriptfon of present administrat ivo cont rols which rest rict the hnndling of i loads to ensure t hat in the onlikoly nvent of a lond drop into !
spent fuel, the radiological results are well within the guidelines of 10CTR100. l l
REASON FOR CilANGE: To provide consistency and clarification in !
the discussions of cont rols for hanlding loads over spent fool.
SAFETY EVAh0AT10N: This snicty evaluntion concluded that thn 1
changa did not involve an 'inreviewed sninty questfon. The dr acribed controls on lond handling have no adverso af fect on the
. Integrity of the handling system. Thereforo, the probability of a lond drop onto spent fuel is not lucreased by tl'n existing administrative controls set forth by Technical Specificat ion Position Statement 126 and Plant. Administ rative
! Procedurn 07-S-05-300. The proponed UFSAR change simply summarizes these controls. The GGNS SER sets the acceptanco criteria for the consequences of a fuel handling accident to bo "woll within the guidelinen of 10CTR100" (less thnn 29 percent of 10CFR100 limits). The described ndministrativo controls limit the ,
potential impact energies (by weight and height restrictions) such i that the rndfological consequences of a postulated drop onto spent '
fuel assemblics is within the acceptanco criterin (less than 25 ,
percent of 10CFR100 limit s). '
The lond handling systems are not being subjected to a different applicatinn than previously used. Thn administ rative controls do ,
not involvn any handling equipment not previously considered in UFSAR for thn hand;!~g of lands ovos the core or the spent fun! l storage areas. Tho described controls do not subject thn i equipment to different appliention than previously used and thernforn the margin of safety as definod in the basis for any ,
Technical Specification were not reduced since the described !
controls place additional conservativo restrictions when handling loads over spent fuel assemblies.
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N1.S90/SNLICFhR - 184
Attachment to GNRO 91/00001 i
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SRASN: N!.S-90-019 DOC N01 OpCon 4 Entry While in SYSTEMt TS 3.5.3 DESCRIPTION OF CilANGE: This safety cynluation addresses the uac j of Technical Specificat ion 3.0.4 to enter Operationn! Coiidition 4 !
f rom Operat ional Cendition 5 while complying with Action Stat ement ;
e or d of Technical Specification 3.5.3. When one or more l suppression pool level instrumentation divisions are inoperable, j The evalunt ion lucludes t he following conalderations: i
- n. Either or both supprin.nion pool level instrumentat f on divisions (A or B) may be inopernble.
- b. An alternate indic9 tor of suppression pool water level in !
used at Innst once per 12 hourt to verify suppression pool ,
water level is greate. than or equal to 12 feet 8 inches. '
- c. There are no operat f ons that have a potentint for draining the reactor vessel in progress. L l d. With no suppression pool level instrumentation OpERAlti.E.
there are no evolutions wit h the possibility of depleting suppression pool inventory (e.g., suppression pool cleanup) 1n progress.
REASON FOR CilANGE: During planned refueling outage activities, situations may arise where suppression pool Invel instrumentation ;
is inoperable in order to perform maintenance act. <ities, surveillance tests, or design change implementations. !
SAFETY EVALUATION: This safety cynlunt.fon concluded that the change did not involve an unreviewed safety question. TS 3.5.3 requires a suppression pool water level of at least 12 feet 8 inches in Operational Conditions 4 and 5. The suppression pool provides a primary source of water for the ECCS in the event of an 1 nce lent to provide cooling water for irradinted fuel. The required pool level is auf ficient to provido the required heat sink capability and water supply to the CCCS. The Ol'ERAllli.lTY of the suppression pool in Operational Conditions 4 and 5 is not required by TS 3.6.3.1 for pressure suppression.
in Operational Conditions 4 and 5 the suppression pool minimum required water volume is reduced because the reactor coolant is innintained at or below 200"F, thn minimum required water volume is ba .d on NpSil, recirculation volume and vortex prevention.
t N1.S90/SNhlCFI.R - 185
Attachmtet to GNKO-91/00001 N1.S-90 019 pnge 2 l 1h> suppression pool water level inst rutnent n t loti provides control room visual confIrmntlon of pool level. TS 3.5.3 Acttons c ntni d require the suppression pool level t o he ver i f ied by an alt ernate irdicator n t. lenst once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> f ri t he event of inoperable suppress ion pool wat er level inst ruinentnt lon.
The UFSAR evaluntos several accidents (esents) which nro conr hiered to be npplicable dus ing Opernt lonni Cotulit ions 4 niel 5.
The en jorit y of t hese are unreinted to the proposed appliention of TS 3.0.4 for TS 3.5.3 i ti that their probability of occurrence is unnffected by suppression pool level instrumentatton status.
Consequent ly, the probabilit y of occurt once of those events does not incrense.
These are:
- n. 1.osses or f ailures of various plant systems or components (other than g. below).
- b. Inndvertent operat don of sarious plant systems or components,
- c. Loss of AC power.
- d. Inndvertent crit.icality events.
- c. Fuel handling nccident,
- f. Inndvertent incrense in shutdown cooling, j - g. l.oss of shutdown cooling.
A renctor vessel drain down event or n loss of suppression pool inventory event are not specifically addressed it the UFSAR during OPERATIONAL CONDITIONS 4 and 5. Ilow eve r , both events are n concein during the use of TS 3.0.4 while in Action c or d of TS 3.5.3 during the period of time changing from Operation 91 Condition 5 to 4.
Tensioning of the renctor vessel head closure bolts has no nffeet on eit her the tr actor vessel or suppression pool water inventories. An alternate indicat ion rnet hod will be used every twelve hours. In addition, no pinnned operations with the potentini to drain either the reactor or suppression pool (when no suppression pool level instrumentation is OPERABhE) will be in progress. Therefore, the intent of the TS basis is met even with suppression pool water level instrumentation inoperable.
NI.S90/SNLICF1.R - 186
Attachennt to GNRO-91/00001 l l
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, Pago 3 !
The use of TS 3.0.4 does not alter the f unction or operation of [
cither F.CCS or suppression pool. Complying with Action c or d of !
- TS 3.5.3 will ensure that suppression pool levol is maintained such thnt the minimum water invol based on NPSil, recirculat ion i volumn and vortex prevention is maintained.
The bases for TS 3.5.3 discunnes the need for suppression pool f'
! volume during Operational Condittons 4 and 5 and is banod on NPSil recirculation volumo and vortex provention. Complying with {
TS 3.5.3 Action c or d will ensure that these concerns and the '
margin proventing these concerns are adequatoly addressed during i nsioning of the reactor vessol hnnd closure bolts. Therefore, the use of TS 3.0.4 will not result. in a decreano of any safety margin as defined in the bases of nny Technical Spncificat.fon.
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Attnehment to GNRO-91/00001 4
1 SRASN: NLS-90 020 DOC N01 OpCon 4 Entry While in SYSTEM:
TS 3.3.6 and !!nsen DESCRIPTION OF CilANGE: Technical SpecificatJon 3.3.6 governs the OPERAlllLITY of the cont rol rod block inst rumentation. Two l
chantiels of the instrumentat ton nanociated with the Reactor Hodo i Swit ch shutdown position rod block are required OPERAllLE to prevent withdrawal of a cont rol rcxl in OPCON 4. If one. or more of the required channels are inoperable, a rod block must he initiated in accordance with ACTION 63. This safety evalunt ion documents the analysis of the use of IS 3.0.4 for entry into OPCON 4 from 5 khen one or more channnis of the Reactor Hodo Switch shutdown position trip function are inoperabin.
This evaluation considers the following
- a. One or more t hannels of the Renctor M(xie Switch shutdown position trip funct.lon are inoperable. ,
- b. A rod block is initinted in accordance with ACTION 63.
REASON FOR CilANGE: During pinnned refueling outage activitfus, altuntions may arise where one channel of the Ronctor Modo Switch shutdown position t rip funct ion is inopernb]n in order to perform i maintennnen or surveillance activtties and/or due to divisional bus outages affecting the normal power supply to the associated instrumentatfon.
SAFETY EVALUATION: This safety evaluation concluded that the change did not involve an unroviewed safety question. The control ,
rod block instrumentation supplies input to the Rod Control and t Information System. The function of tLese inputs is to inhibit control rod movement or selection to i ovent vnnctivity changes in OPCONs "I and 4. Two channels p1av W input to this trip functlon.
Technical Specification 3.3.6, ..CTION b, establishes requirements through Tabin 3.3.6-1 for the minimum numbnr of operable channels.
If thn minimum number of operable channnis cannot be met , Action Statement. b refers to Tablo 3.3.6-1, whl n specifies required ACTION 63 to be t aken. ACTION 63 requires a rod block to bn initiated thereby positively fulfil 11og the safety function.
l The UTSAR evaluntos several accidents (events) which are considered to be applicabin during OPERATIONAb CONDITION.*. 4 and S.
, Tha majority of these nrn unreinted to the proposed applimtion of l
TS 3.0.4 for TS 3.3.6 and thus the prohnbility of occurroico of these oventn does not increase. The reactivity insertion event is not specifically analyzed in the UFSAR for OPCONs 3 and 4 because the core is assumed to be subcritical. With the control rod block initiated in accordance with ACTION 63, a control rod cannot be
- inadvertently withdrawn. The coro remains subcritical and the assumptions of the accident analyses are preserved.
j NLS90/SNLICFhR - 188 l
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NI.S-90-020 Page 2 l
Technical Specification 3.0.4 presently allows entry into an OPCON '
or speciflod condition when I,COs arn not met if the plant is in conformanco with the 1,C0 Act fun requirernents and thoso requirements permit continned operation of the facility for an '
, unlimited period of timo. This is in accordance with the NRG's stated position and has been accepted for GGNS. Although not specifically addressed in the finses for TS 3.3.6, compliance with the requirements of ACTION 63 while changing from OPCON S to 4 vill provjde the samo invol of safety as cornpliance with the ICO.
Thnrnforo, the use of TS 3.0.4 will not result in a decrenso of any snfoty margin as defined in the bases of nny Technical i Specificatfon.
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NT,S90/SN1,1CFl.R - 189
At.tachment to GNRO-91/00001 SRASN: NJ.S- 90-021 DOC NO: OpCon 4 Entry While in SYSTEM:
7.5 DESCRIPTION
OT CilANGE: This safoty ovaluat ton documents the analysis of the use of TS 3.0.4 for entry into Operational Condition 4 from Operat ional Condition 5 while complying with Actfon 81 of Tablo 3.3.7.5-1 for the following instruments of Table 3.3.7.5-1:
l n. Containment /Drywell Area Radiation Monitors (ltem 13)
- b. C >nt ainment Ventilation Exhaust Radiation Monitor (Item 14)
- c. Offgas and Radwaste llullding Ventilation Exhaust Radiation Monitor (ltem 15)
- d. Fuel llandling Aron Ventilation Exhaust Radiation Honitor (ltem 16)
The evaluation takes into consideration that thn preplanned j altr4rnate method of monitoring the appropriate parameter (s) is initiated within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
REASON POR CilANGE: During planned refueling outago activition, situations may arise where instruments are inoperable in order to perform mr.intenanco activities, surynillance tests, or design chango impicmentation.
SAFETY EVALUATION: This safety evaluntion concluded that the chango did not involve an unroviewed safety question. The Accident Monitoring Instruments covered in this ovaluation do not perform any automatic functions to mitigate a UDA or transient.
These instruments ensure that suf ficient information is available during a DBA or transient. Act.f on 81 requirements providn an acceptably safe alternative means of meeting the LCO. First, the Action requires that an alternatn prop anned method of monitoring the appropriato parameter be initiated. This ensures that. In the event of a DBA or transient, an alternative method of monitoring Sho parameter is availabic which will allow ar.nessment of important variables following an accident. Secondly, the Action requires that a special report be pregiated and submitted to the NRC outlining the cause of the inoperability and the plans and schndule for restoring operability. This ensures timely attention and resolution of the inoperability as well as an additional review by the NRC of the specific conditions involvt. I in the inoporabi1ity.
The requirements of this Specification are applicable in various Operational Conditions dependent upon the instrument.. Since the action requirements set up conditions equivalent to those required by the LCO, none of the evolutions involved in changing to Opnrational Co~11 tion 4 f rom 5 result in any change to the level of safet y.
NLS90/SNLICFLR - 190
Attachment to GNRO-91/00001 SRASN: NhS-90-022 DOC NO: UFSAR Apnendix 13A SYSTEM:
DESCRIPTION OF CllANGE: The revision to Appendix 13A primarily reflects chinges t o updat e the Resumes of key personnel associated with the operation of GGNS. Other changes include: addition of a few key positions, deletion of the lechnical Assistant to the Operations Superliitendent, and removal of the resumes of those posit ions i cport ing to t he Rndfation Control Superintendent.
These changns are explained as follows:
- The positions ndded nre: Director, Fuels; Manager, Nuclear fuels Supply; Mnnager, Nuclear Fuels planning.
- The functions of the Teclinica l Assist ntit to tlic Operatloris Superintendent have been nssumed by the Operations Assistants per MTO-90/0386
- The posit ions currently reporting to the Rndintion Control Superintendent include two Kndiation Control Supervisors and a Technical Assistant. The Nuclear Plant Sn f et y Coord innt or reports to the Technical Assistant. With the exception of certniti positions eport ing t o t he Opet at ions Super intendent ,
Appendix 13A gentrally does not capture " Supervisors" unless they happen to be on t he " Superintendent" reporting level.
REASON FOR CllANGE: Tha revision of t esurnes contained in UISAR Appandix 13A is an administrative change which provides (1) the most current 1ist Ing of posit ions support lug the opernt ion of GGNS, (2) names of personnel f1i1ing t hose posit ions , nint ( 3) the most current status of experience for these individuals.
SAFETY EVAh0AT10N: This sa fety evalunt .lon concluded t hat the change did not involve an unroviewed safety question. Individunts assigned to new positions are required to meet the qualificatton requirements specified in the UFSAR. Analyses in the UFSAR which resume operator error would remain unchanged based on these individuals meet ing t he UFSAR sequirements. No syst em functions or designs are being changed.
Individuals assigned to new positions are required t o meet t he qualificatfon requirements specified in the Teci.n i ca l Specifications; therefore, there is no reduction in the margin of safety as defined in the hnsis for any Technical Specification.
NhS90/SNh!ClI,R - 191
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Attachment to GNRO-91/00001 f
t l SRASN: NSP-90-003 DOC NO: Change of Executive Director, SYSTEM:
Operatione Support to i Vico President, Operations Support l t
DESCRIPTION OF CilANGE: This evaluation changes the tillo Executive Director of Operations Support (EDOS) to Vice President, Operations Support (VPOS). j REASON FOR CilANGE: This title change reflects the correct level of management. who reports to the Executiva Vice President and Chief Operating Officer. ;
i SAFETY EVAL,UATION: With this title change all the duties, !
responsibilities and commitments being performad in thn existing i organfr.ational structure will continuo to be performed. The title ,
chango will have no offect n plant design or operations; }
thereforn, there will be no increase in probability of occurrence !
or consequences of accidents previously evaluated in the UfSARt not will thorn be any increase in probability of occurrence or >
consequnncca of a malfunction of equipment important. to safety [
previously nyaluated in the UFSAR; nor will there be created the ;
possibility of an accident or malfunction of equipment important !
to safety dif ferent than any previously evaluated in thn UFSAR. !
l b
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s l
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L NSP90/SNhlCFLR - 192 f I
Attachment to GNRO 91/00001 SRASN: NSP-90-004 !)DC NO: Onsito Storage of New SYSTEM:
Fuel for GGNS Cycle 3 I)ESCRIPTION OF CllANGE: This nyaluntion is for those act ivities concerning the new funi bnndles produced for Cycle 5 by Advanced Nuclohr Fuels Corporationt
- a. The movement of new f uel to either the new fuel vault or the spent Nel rack,
- b. The storage of ANF-1.4 fresh reload fuel in the new fuel vault,
- c. The storage of ANF-1.4 fresh reload fuel in the spent funi pool.
REASON FOR CHANGE: The introduction of the new fuel design at GGNS is to improvn thn fuel cycle economics and inctnase the operational ficxibility of the reactor care.
SAFETY EVALVAT ON: Confirmatory analyses havn been performed to show that the ANF-1.4 reload fuel bundles have weights and geometries similar to those of thn GE fuel bundles on which the analyses described in the SAR are based. No new act.fvities are required for the movement of ANT-1.4 fuel bundles to the new fuel vault or the spent fuel pool. Precursors to any accident previously evaluated will not be affected.
Confirmatory analyses have been performed to show that the ANF-1.4 reload fuel is compatiblo with, and similar to, the roload fuel stored in the new fuel vault during previous roload activities.
The NRC has approved a revision to thn licensing basis for storage of the ANF-1.4 rolond fuel in the spent fuel storage racks.
Because of the similarity of the ANF-1.4 reload fuel to the reload funi stored in the spent fuel pool during previous roloads, the storagn of the new fuel types in the spent. fuel storage racks will not. af fect, the precursors to any accident previously ovaluated.
Thereforo, performing the activities in connection with onsito storage of new fuel for Cyclo 5 will not increase the probability of occurrence of an accident previously nyaluated in the FSAR.
l The fuel handling accident is nyaluated in the FSAR. Thn i
radiological consequences of dropping an unirradiated fuel bundin on thn spent fuel racks was evaluated and found to meet tho applicable acceptanco criteria. This analysis includus fc-1 parameters applicable to ANF-1.4 The radiological consequences of dropping an unirradiated fuel bundle are determined by the performance of the irradiated fuel in the spent fuel rack.
The re fore , the consequences of dropping an ANT-1.4 fuel bundin on the spent fuel racke are unchanged.
NSP90/SNhlCFLR - 193
l Attachment to GNRO-91/00001 NSp-90-004 page 2
- Confirmatory analyson have shown that the reactivity of the !
ANr-1.4 reload fuel in the new fuel vault is within the acceptance i criteria established for previous rnloads for new fuel. l Thorofore, as for provfous reloads, the occurrence of inadvertent ;
criticality is precluded for ANF-1.4 reload fun 1. !
The analyses perforand in support of the revised basis show that the maximum reactivity of the racks when loaded with ANT-1.4 roload fuel is within the acceptance crit eria for the spent fuel !
l pool criticality analysis. The ANf-1.4 reload fuel has a similar !
static and dynamic responso and therefore, the consequences of a l solamic event remain unchanged. Therefore performing the !
activities in conunction with onsit e storage of new fuel for !
Cycle 5 will not increano thn consequences of an accident ;
previously evaluated in tho FSAR. ;
The equipment required to be used for the onsito storago and ;
handling of the new fuel bundles is similar to that required to bo ;
used for previous roloads; no addit ional loads will be imposed on '
any equipment.; no increase in frequency of operation of the equipment will result. The precursors to any malfunction of '
equipment important to safety will not be affected, Thornfore. I performing t he activitier. in connection with onsito st orage of new .
fuel for Cycle 5 will not increano the probability of a malfunction of equipment important to safety previously evaluated I in the FSAR. l The fuel handling and storaan equipment will not bo subjected to operational conditions different f rom t hoso during provf ous ;
reloads; chnngen to the equipment protection features will not bn ,
required. Thereforo, performing the activities in connection with -
onsite storage of new fuel for Cycle 5 will not increase tho consequences of a malfunction of equipment important to safety ,
previously evaluated in the FSAR. !
The activities associated with the onsite storage and handling of ANT-1.4 ralcad fuel are unchnnged from those associated with the i onsitn storage and handling of new fuel for previous reloads; no i new operational modes will be required; no plant modifications will be requi ted. Therefore, performing the activities in connection witMthe onnito storago and handling of new fuel for .
Cycle 5 will not. create the possibility of an accident, of a !
different type than any already evaluated in the FSAR. l l
- NSP90/SNI.lCFLW - 194 l \
l
_ _ _ . _ . - . ~ _ _ _ _ _ _ _ _ _ -._
Attachment to GNRO-91/00001
} NSP-90-004 Page 3
, Based on the operatinnal requirements for the fuel handling i equipment, no new equipment is required for the storagn or handling of the ANF-1.4 roload fuel. No new fuel handling activities are required in connection with the onsite storago of ANF-1.3 reload fuel; no modifications to the existing equipment nrn required; no changes in operational setpoints are required.
Therefore, performing the activities in connection with the onsito storagn and handling of new fuel for Cyclo 5 will not create the possibility of malfunct ion of equipment important to safety different than previously evaluated in the FSAR.
The fuel handling accident. has been evaluated. An analysis of the radiological consequences of dropping unirradiated fuel on the spent fuel racks, with and without secondary containment, was performed. This evaluation established height / weight restrictions on the movement of objects above the spnnt fuel racks which were implemented. These restrictions assure that the radiological consequences of a fuel handling accident for unirradiated fuel monts the acceptance criteria of 25% of 10CFR100 dose rato limits.
This ovaluation included the fuel design paramet era applicable to '
the ANF-1.4 fuel design. Therefore the margin of safety remains unchanged.
Analyses have been performed to determinn the reactivity for the ANF-1.4 roload fuel. The acceptance critorion stated in the F9AR for K-of fective in the new fuel vault is 0.95. The corresponding licensing basis value for maximum in-core reactivity is 1.31 (K-lufinity), as determined for previous reloads. The analyses described in the FSAR and which form the bases for the Technical Specification are based on this value of K-infinity. The maximum !
in-corn reactivity was calculated to bc ).1847 (K-infinity). This volun is bolcw the acceptance criterion of 1.31 established for previous raloads. p The NRC has approved a revision to the licensing basis for the storage of the ANF-1.4 reload fuel in the spent fuel pool storago racks. Thn maximum react ivity for the storago of ANT-1.4 fuel as
. stated in the NRC safety evaluation is 0.9452 (K-offective). This 1 is below thn acceptance criterion of 0.95 (K-of fective). The ;
acceptance criterion remains unchanged from that for previous reloads.
The static and dynamic responso of ANF-1.4 rnload fuel is similar
, to that for the funi used in provfous cycles. The margin of
! sciety for scismic events is thernfore unaffected by the use of l ANP-1.4 reload fuel. Thoroforn, performing the activities in connect. ion with onsite storage of new fuel for Cycle 5 will not result in a reduction in the margin of safety an defined in the hasis for any Technical Specificatfon. ,
f NSp90/SNI.lCFI.R - 195 ,
,,-.y~r----,----,--e ..w.y .-e,-
-.-*+=-e.- - - . . . . m.---.as -----..--...-,--.---.---,.w., m-----.mm.e --w...--w.,-e,y-e-w-y--vy.%%~-.w.,--,m,,+--3,
Attachment to GNRO-91/00001 SRASN: NSP-90-005 DOC NO: RF04 fuel Manngement SYSTEM:
DLSCRIFT10N OF CllANGE- This evaluntion is for the movement of fuel buintles and thn shuffling of funt assemblics in thn reactor core.
RFASGN POR CnANGE: This fuel movement was to support Refueling Out ago Number Four (RF04).
SAFETY EVALUATION: Tho safety evaluntton concluded that the change dbl not involve an unreviewed safety questlon.
Confirmatory evaluations have shown that the accident analysns described in the UFSAR, which were applicable to previous relonds, continue to remnin npplicable and bounding for RF04. The precursors to any accident previously evalunted will not be affected.
A confirmatory evaluntion has been perfotrned to show that the ANF fuel assemblies have weights, geometries, and seismic responso characteristics simlinr to those of the GE fuel assernblies, on which thn nnnlyses described in the UFSAR are based. Because the masses and drop heights are essentinIly the same, the momentum and k inet ic energy ef f ect s of dropping an ANF fuel assembly are similar to those for previous reload fuel types. A bounding evaluntion has shown that t hn dose rates resulting from the drop of an ANP fuel assembly are within thn dose rates acceptance crit erion stat ed in t hn GGNS-1 Saf ety Evaluation Report . Using the same annlysis assumptions for the GE and ANF fuel types, it has been shown that the radiological consequences resulting from the drop of an ANP fuel assembly are bounded by the consequences that would result f rom the drop of a GE tuel assembly. The procursors to any accident previously evnlunted will not. be affected. ..
Cniculations have been performed to show t hat adequat n shutdown margin exists during fuel shuffling. Restrictions applicable to fuci shuf fin act ivit ies have been provided to GGNS-Reactor Fugineering for inclusion in the approprinto procedures in a manner simlinr to previous relonds. The precursors to any accident previously evaluated will not be affected.
The equipment required to be used during R104 is similar to that used for prnvious relonds; no additional londs will be imposed on ,
any equipment as a result of hnndling thn ANF fuel assemblies; no -
incrense in frequency of operation of the equipment will result; no new operat iona l mados a rn required; no plant modificat ions are required; no changes in operatJonni setpoints are required. The precutsors to any malfunction of equipment important to snfety will not be affected. The consequences of a malfunction of equipment important to safety are bounded by the consequences evalunted in the UFSAR.
NSP90/SNhlCFI,R - 196
At t nchmetit to GNRO-91/00001 NSP-90-005 Vnge 2 Accident annlyses applicnble to previous selonds during opnintionni modes 4, 5, and
- cit her cont inue t o remain applicnole for Rr04, or nec cycle-specific. The accept atico cr it oria applienbic to the accidents nnnlyzed for previous relonda cont inue to be satisfied. Accident nonlyses/evnluntions have been performed to show that cycle-specific events meet the acceptance criterin. A bounding evalunt ion, using conservat ive assumpt ions, hns r,hown that the ind f ological conserpiences of dr opping an ANT f uel assernbly cont inue t o sat is f y the Sa fety F.valuat ion Report neceptance criterion (25'. of 10CIR100 limits); tha dropping of n Gr. fuel assembly, on shich the UFSAR annlyses ,sre based, continues to I:- the limiting event f or det er mining the rndiological conserpiences of t he fuel linndling Accident . The Shutdown Margin determined by the annlyses is within the accept ance critorion (19.)
established in the UFSAR. Theiefore, performing the act ivit ies in connect ion with reincling act ivit ies during Rr04 for Cycle 5 fuel will not result lii a reduction in the margin of safety as defined in thn bnsin for any Technica l Speci ficat 100.
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i NSP90/ SNI,1CFl.R - 197
Attachment t o GNRO-91/00001 4
! SRASN: NSP-90-006 DOC NO: Refueling Operations with SYSTEH:
Revised Core bonding Plan DESCRIPTION OF CilANGE: The refueling operations 1.1 Modes 4, 5 and
- worn previously evaluated assuming fuel bundan XNH-487 would remain in the corn for Cyclo 5 operation. 'inir. safety evaluation ,
addresses refueling operations with funt bundle XNH-529 rep 1ncing l XNB-487 in it s beginning of cycle (HOC) locat ion (21,58) for tan ;
following proposed activities:
- 1. The movement of funi bundles, and
- 2. The shuffling of fuel assemblion in the reactor core, !
REASON TOR CilANGE: During the course of loading the GGNS-1 Cycin !
5 core, funi bundin XNB-487 was dropped from slightly above the core into its designated location (21,58). This bundin was i
initially inscr* ~1 in the corn during Cycle 3 and reinserted in Cycin 4. It has similar reactivity characteristics to fuel 3 pinnned for dischargo during HF04. Therefore, replacing thn ,
l bundin with a bundin planned for discharge was considered morn !
practical than requalifying the dropped bundle for nn additional cycle of operation. Fuel bundio XND-529 was identified as the appropriate repincoment bundle. This bundin has similar reactivity performanen to XNB-487. Both fuel bundles have the ;
namo nucient design but the repincement bundin (XND-529) has a slightly higher burnup and therefore slightly lower rnactivity.
SAFETY EVAbOATION: The safety evaluation concluded that the chango did not involve an unroviewed safety question. A confirmatory nvaluation has been performed to show that thn ANP fuel assemblies havn weights, geomotries, and seismic response characterist ics similar to those of the GE funi nssemblies, on which the analyses described in the UFSAR arn based. Because the masses and drop heights are essentin11y the same, the momentum and kinetic energy of fects of dropping an ANP funi assembly are similar to thoso for previous reload fuel typns. A bounding nynluation has shown that the dose rates result ing from the drop of nn ANF funi nssembly are within the dose rates acceptance criterInn stated in the GGNS-1 Safety Evaluntion Report. Using the samn nnnlysis assumptions for the GE and ANF funi types, it i
has been shosn that thn radiological consequences resulting from ,
the drop of an ANF fuel assembly are bounded by the consequences that would result from the drop of a GE funi nannmbly. Thn changn in the Cycle S core londing plan only *cplaces one ANF fuel bundin with a similar bundle of the same design. The precursors to any nccident previously nyaluated will not be affected.
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NSp90/SNh1CFLR - 198
A t t a chnien t to GNKO-91/00001 NSP-40-006 Page 2 Calculntions havn been per f ormed to show that adequate shutdown innt gin exist s dur ing f uel shuf fling. These calculat ions bound the rcvised core configurntton. Restrictions appltenble to fuol shuffle not tvit les hnvc been provided to GGNS-Reactor 1:ng i nee r i ng for incloston in the appioprinte procedures in n inannei aimilar to previous relonds. The precursors to nny accident previously evnlunted will not be nifected.
The fuel that will bn hnndled during kr04 is similar ,o, and compnt ible wit h, t he fuel that was barnlled for previous relonds.
The equipment required to be used during kr04 is similar to that used for previous relonds; no additlonn1 loads will be imposed on any equipment ns a rer. ult of hnndling the ANT f uel nascrublies; no increase in f requency of opernt ton of t he equipment will tesult.
The refueling act ivit ies associnted with Cycle 5 fuel will not subject the equipment to opnrntional conditions different from those during previous relonds; chatiges t o t he equipment protection fontures will not be required.
A bounding evnlunt f on, using conservat ive assumptions, has shown thnt the radiological consequences of dropping an ANT f uel assettbly centinue to satisfy the safety F.vnluntion Report acceptnnco cr it erion ( 25'. of 10C1R100 1imits).
The Shutdown Margin (SDM) determined by the nuntyses is within the acceptance criterion (11) established in the UFSAR. The revised core configurat ion has the snme or slight ly less SDM niul therefore the previous annlysis results remains applicable. 'i h e r e f o r e ,
performing the act ivit les in connect ion wit h ref uelitig act tvit les during Kr04 for Cycle 5 fuel will not result in a reduction in t he mntgin of safety as defirud in the basis for nny Technical Specifications.
NSP90/SNhlCFI.R - 199 1 1
At t nchttent to CNRO-91/00001 SRASN: NSP '30-00 7 DOC NO: t!rSAR 15.5.1 SYSlui:
DESCRI PTION Ol' CilANGF,' A tevialoti of t he (TSAR wns matie to ndilr os s the inndvertent statt op of the Iligh Pressut e Cot e Spray (llPCS ) systom. The revision <ieserIbes two alt ernallye event n c< pie n c e s that could result from itindvet t ent llPCS startup.
REASON FOR CilANGT,' To describe in t he lifSAR nn evet,t serpiente that was observed during the inndvertetit. IIPCS nct't.st ion that occurred on October 10, l 'J 8 8. The reactor level control syt, tem was utiable to compensnte for t he level incrense resulting from lil'CS Inject ion. In spit e of opet ator act lot.s to mit ignt e the level increase, the teactor vessel level increased, resulting in a trip sinnni in one :.f t wo i nn,t or prot ect ion syst em inst t umet:t at ion channels.
SAFLTY EVAL.UATION: The saf et y evalunt lon concluded t hat the
- hnnge ditt not itivolve nn unreviewed snicty question. Of the two nit einnt ive event serpiences i nsult inn fi nm the inndve rt ent IIPCS nct unt ton ( Ill A) , the sequence that result s in n new equilibrium power level has been annlyzed pt eviously ntut is described in the ISAR. The sequence lending to the high level t rip is boututed by t hn reedwat er Cont iolle r l'ailure - Maximum Demniul tiansient. Thn two alteinntIve event sequences do not requite the use of any new equipment or t he use of exist liig equipment in any new functlocal capacity. No changen to pinnt operntional modes are t e<pil t ed . No plant modifirntions nre required.
The delt a-Crit ical l'ower Rat to (CPR) for the event sequence tesulting in a new equilibrium power level has been nonlyzed previously. The dnit n-CPR for the event sequence resulting in the high level t r ip is boutuled by the deltn-CPR for t he Feedwn t e r Cont roller Failure - Mnximum Demand t ransient ; this transient has beeti annlyzed on n cycle-c.pecific basis as one of the l i m i t. i ng transients thnt causes increase i ti reactor vessel invetit ory niid decrease ist reactor coolnnt temperature. Consequent ly, t het e is no t educt ion in t he mar gin of sa f et y As defitied iti the basis for any Tochtilen i Spectflentlon.
I 1
NSP90/SNI.lCFl.R - 200 l
r Attachment t o GNRO-91/00001 SRASN: NS.'-90-008 I)DC NO: Cycle 5 OPS With Revised SY S TF.'i t Core Configuration Dr.SCRil'T10N OF CilANGr.: 'lhis safety evaluntion was written to demonst rate the acceptability of Cycle 5 opeint ion with a revised core configurntlon.
RF.ASON l'OR CilANGr.: The new cotifsguration was necessitated by the replacement of an Adynneed Nuclear fuels (ANP) 8x8 fuel assembly (XNil-487) wit h n s imilar , less renctive ANT M8 fuel assembly
( X Nil- 529 ) in core locr.t ion ( 21,58 ) . The old assembly was diopped during refnnling.
- ' Al f.TY FN Al.U ATI ON : The saf ety evolunt ion concluded that the chnnge did not involve nn unreviewert snfety quer.tton. The r e pl acernen t f uel nssernbly is of a design similar to the assembly thnt wns to bn present in the NRC-approved Cycle 5 core configurntlon. The supporting annlyses for thn NRC-npproved Cycle 5 core configuintton continne to temnin applicnbin for the revised configurntlon.
The repincement f uel assembly is less renct ive nnel hns been pinced in n non-limiting core location. The postuinted accidents for the revised Cycle 5 core configuration have been shown to be no inore Fevere tilan the post ulated accidernt s for the NRC-njiptoved Cycle 5 core configuration. Ilocause the repincement f uel assernbly is similar to, and compatible with, the fuel nssembly it hns replaced, no new equipment will bn required; no new nctivities are required; no modifications to the existing equipment are required; no changes in opernt ion setpoints are required.
An evaluntion of thn impact of the revised core configurat ion on the fuel mechanical design limits, plant transients, and postul.ted accidents has shown that the supporting annlyses that were per formed for the NRC-approved Cycle 5 core conf igurat ion remnin applicabin for the revined Cycle 5 corn configurntion. The analytically determined limits npplienbin t o the NRC-approved Cycle 5 core configurntion continun to be applicable to the revised Cycin 5 corn configuration; the availnble margins to their respactive acceptnnce i f in i t s are unaffeeted.
NSf90/SNhlCfl.R - 201
~.
V =l 1
Attachment to GNRO-91/00001 SRASN: NSP-90-009 DOC NO: Cyclo 5 OPS With SYSTEti:
9X9.5 Reload i
DESCRIPTION OF CilANGE: This safety evalunt.fon addresses those issues associated with CycIn 5 operation with ANF 9x9-5 fuel assemblies that have not already been ovaluated under othnr 50.59 safety evaluations or in thn Cycin 5 reland PC0h. Items evaluated included: ,
- 1) A confarmatory analysis to verity .. m the baseline at41yses continun to remain applicable to the ANF 8x8 corn from thn standpnint of energy releases to thn con t.n inmen t.
- 2) An analysis comparing the energy release from a ANP 8x8 fuel
- assembly with that of and ANP 0x9-5 fuel assembly.
- 3) An analysis to confirm adequate recombiner capacity for cycle 5.
- 4) A Firo Scenario Evalunt a for 9x9-5 Reload Funi.
- 5) An analysis to ensure compliance with the Anticipated Transients Without Caam (ATWS) rulo. The baseline analysis, which assume a GE 8x8 fueled core, were reevaluated for i applicability to the ANF funi typos.
- 6) The Emergency Procedures were reviewed to ensure no changes to thn fuel related inputs to the supporting analyses for the Emergency Procedures worn necessary.
REASON FCR CIIANGE: To assess a11 other fue1 dependent issues for Cycle 5 operation not previously addressed.
SAFETY EVALUATION: There is no incronsn in thn probrbility of occurrence or in the consequances of an accident or malfunction of equipment important to sn r oty previously evaluated in the Safety
, Analysis Report, because:
a) Tho events that could result in a design basis LOCA (DBLOCA) are bnand on certain a prior assumptions. They arn independent of fuel stored energins.
I b) The events that could result in a DBLOCA are based on certain
! a prior assumptions. They are independent of active clad
( volumn.
c) The events leading to a major fire that could affect safe shutdown' capability are a function of pinnt operational conditions. They cre independnnt of the funi typus resident in the core.
I d) The events leading to an ATWS are determinnd by the responso l of thn r< 1etor shutdown systems to abnormal plant conditions.
They are independent of the fuel types resident in the corn.
N8p90/SNLICFLR - 201 l
Attachment to GNRO-41/00001 NSp-90-009 pagn 2 e) The stored energies in the funl assemblics, which are tho only significant fuel
- dependent parameters used in determining containment. response to a DBh0CA, have been compared for the GE and ANF funi types. The comparison has shown that the maximum stored nuergy in the ABF 9x9-5 fuel assembly is bounded by that in the ANF 8x8 fuel assembly; the differnnce in the maximum stored energy between the ANF and GE 8x8 fuel assembly is insignificant, Furthermore, tho fuel stored .cnergy is a small part of the total energy released to thn containment. The parame ors used to dotormino containment response during the DIlh0CA nro unchar.ged, f) The active clnd volumn that. was used in sizing t.he hydrogen recombiners bounds tho active clad volume that will bo present in the Cycle 5 core, g) The peak clad temperatures (pCTs) during a major firn have been shown to be well below the temperaturn of incipient clad deformation for all ANP fuel types that will be present in the Cycle 5 corn.
h) The ANF fuel designs are compat ible with the GE fuel dos.ign, on which the FSAR analyces arn based. The core-wide response to an ATWS ovent resulting from the insertion of Cycin 5 fuel has been determined to be no morr snvern than that for previous cycles. *
-The postulated accident s for Cycle 5 havn been shown to be no more severo than the postulated accidents for previous cycles. There la na creation of a possibility for an accident or malfunction of 3 a different type than any evaluated previously in the Safety Analysis Report.
Thn Cyrin 5 fuel is similar to and compatiblo with the fuel inserted into the corn during previous reloads. The design of the Cycin 5 fuel does not requlr2 any activities different from thosn associated with previous cycles; no new operational modes arn requirnd; no plant modifications arn required. Additionally no new equipment will be required; no now activities are required; no mtAlfications to the existing equipment are required; no changes in operational setpoints are required.
a) The fuel atored energy constitutos a small part (6.8%) of the total energy released to the containment during a DBh0CA.
The impaut of changes in the stored energy (0.38% higher for ANF 8x8 funl, compared to GE 8x8 fuel and 12.2% lower for ANP 9x9-5 fuel, compared to ANF 8x8 fuel) results in a decrease in stornd energy for thn Cycla 5, as compared to the GE 8x8 core.
NSP90/SNLICFLR - 203 s
- - - - - _ ~ . ~ . - - - . ~ . - - . . . . .
At t achment to GNRO-91/00001 l
'I NSP-90-009 I Pagn 3 l
b) The active clad volumn for the Cycle 5 corn (2693 cubic inches) is less than that used to sizo thn hydrogen )
i recombiners (2696) cubic inches). The design basis criter. i for sizing the r trogen recombinors continues to bn satisfied for Cycle 5.
c) The PCT during the major fire event. for GE fuel (700 degrees F) provides for a margin of 1190 - 700 = 490 degrees F to ,
incipient cindding deformation. The corresponding margins for ANF 8x8 and 9x9-5 funis are 1500 - 870 = 630 degrees F and 1500 - 801 = 699 degrees F. respectively. The availablo margin for ANP 9x9-5 fuel is greater than that. for ANF 8x8 fuelt both ANF fuel t.ypes havn incronsed margin to incipient clad deformat lon than GE fuel.
d) The coro averngo responsn and vossol pressurizat.lon effects for the Cyclo 5 core during an ATWS havn been dotormined to ho no morn severo thnn those for previous cyclos bnenunn the ANF and UE fuel dt <,f gns are similar. The nctions required to mitignt.c the ef fects of t.bn limiting ATWS ovent for Cycle 5 nro unchanged; the ability to maintain critical plant parameters within thn limits nstablished prnviously is unchanged.
The acceptanco criterin applicable to previous cycles continue t o
. be adequately sat isflod for Lli o issues described.
Thnrefore, by implementing or performing the acti.ons described, a reduction in the margin of safety as defined in the basis for any technient specifications will not. result.
I NSP90/SNLICFLR - 204 L 2