ML19332E318
| ML19332E318 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 05/31/1989 |
| From: | Cottle W SYSTEM ENERGY RESOURCES, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AECM-89-0093, AECM-89-93, NUDOCS 8912070065 | |
| Download: ML19332E318 (221) | |
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. November 30,-1989 iWAxr Orectons '
, U.S. Nuclear Regulatory Commission Mall . Station P1-137
, Washington,.D.C. 20555 Attention: Document Control Desk <
Gentlemen:
SUBJECT:
Grand Gulf Nuclear-Station Unit 1 Docket No- 50-416 License No. NPF-29 Report of 10CFR50.59 Safety Evaluations'- January 3, 1989 through May 31, 1989 AECM-89/0093 In accordance with the requirements of 10CFR50.59(b), System Ener9y Re:.ources, Inc. is reporting those changes,-tests and experiments under the
? y requirements of 10CFR50,59 for the period of January 1,1989 through
, .May 31, 1989. a summary of 'these changes,- tests, and experiments is contained
.in the fIrst M tc hment.
Attachment' 2 contains a- brief description of those evaluations performed
'under 100FR50.59 that support Revision 4 to the Updated Final Safety Analysis Report but which have not yet been included in a 10CFR50.59 summary report.
LThe evaluations listed in Attachment-2 are provided here as required by 10CFR50.71(e) and will be nummarized in the-next-10CFR50.59 report.
'Yours truly, c.4) I d e, i
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7.. AECM-89/0093-Page 2: Tr cc: Mr. D..C. Hintz (w/o) Mr.- T. Hc Cloninger (w/o) Mr. R. B. McGehee (w/o) ; Mr._ N. S. Reynolds (w/o)- Mr. H. L. Thomas-(w/o) Mr. H. O. Christensen (w/a) I Mr. Stewart D.- Ebneter (w/a) Regional Administrator U.S. Nuclear Regulatory _ Commission ; g Region,II. 101 Marietta.St.,.N.W., Suite 2900 Atlanta, Georgia 30323 L. Kintner, Project Manager (w/a) Mr. L.~ Office of Nuclear Reactor Regulation. - U.S. Nuclear Regulatory Commission Mail Stop 14B20 Washington, D.C. 20555 1 l #- l A9111502/SNLICFLR - 2
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-OF-i L10CFR50'.59 SAFETY EVALUATIONS'FOR THE PERIOD' !
4
'JANUI@l.l', 1989 THROUGH MAY 31,_1989 hhy .
2o ' , 'SRASH DOCUMENT NUMBER .PAGE
>e R NPE-89'001-DCP-84-3090-S00-R00 1- ,lj.J NPE-89-002i - CN-89-0037- 2 s . , ,NPE-89-003c PMI-89-00593. 4.
NPE-89-004- CN-89-0049- 6
< ?NPE-89-005- DCP-87-007 5-S00-R00&R01- 8
.NPE-89-006 . DCP-87-0029-S00-R00- 9 --
LNPE-89-007 M-300~.2-S01-R00 10 ; NPE-89-008 SCN-89-0002-to ES-19, Rev. 6 12 1
.NPE-89-009 MCP-88-1019-S00-R00 14 ni .
"NPE-89-010 DCP-85-4005-S00-R00 15 n
.NFE-89-011 EER-89-6066 16 LNPE-89-012J CN-89-0136- 18 NPE-89-013 MCP-88-1050-S00-R00 20 D NPE-89-014 - DCP-87-4007-S00-R00 22' s
h- 'NPE-89-015' - DCP-84-0134-S00-R00 23
-;4PE-89-016 -- MCP-88-1029-S00-R00- 24 2NPE-89-017_ - DCP-83-0568-S00-R00 26
'NPE-89-018 ' Temporary Lead Shielding Request '28
'NPE-89-0194 EERR-89-61031 29
<. i NPE-89-020 DCP-84-3016-S00-R00 '31 M 'NPE-89-021 - DCP-89-1050-S00-R00&R01 32
- NPE-89-022- DCP-89-0203-S00-R00&R01 34 0 NPE-89-023- DCP-84-3051-S00-R00 36' !
-f' ' , , i .NPE-89-024' - DCP-86-4500-S00-R00 37
! NPE-89-025 DCP-82-0636-S00-R00 38' NPE-89-026- DCP-84-0011-S01-R00 40 NPE-89'027 DCP-84-0064-S03-R00 42' j NPE-89-028- DCP-84-0235-S00-R00 46 ylt- ;NPE-89-029 MNCR-88-0129 47
< NPE-89-030 _ DCP-83-0202-S00-R00 48 iNPE-89-031 DCP-83-0034-S01-R00 49 -i DCP-84-0236-S00-R00 50 NPE-89-032
.NPE-89-033 MCP-88-1038-S00-R01 52 NPE-89-034 DCP-88-4501-S00-R00 53
.NPE-89-035 DCP-84-3000-S00-R00 56
*NPE-89-036 DCP-84-4077-S00-R00 57 NPE-89-037 MCP-89-1055-S00-R00 59 NPE-89-038 DCP-84-0020-S00-R00 60 4! NPE-89-039 DCP-84-4032-S00-R00 61 !
LNPE-89-040 DCP-85-0066-S00&S01-R00 62 - NPE-89-041 DCP-83-4528-S00-R01 63 NPE-89-042 DCP-88-0010-S00-R00 64 NPE-89-043 MCP-89-1052-S00-R00 67 j NPE-89-044 DCP-82-5073-S00-R00 69 NPE-89-045 DCP-86-0037-S00-R00 70
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TABLE OF_ CONTENTS l(Continued)- ;f <
* "' PAGE-SRASNI DOCUMENT NUMBER NPE-89-046 CN-A30-032- 71' Bi M NPEE 89-047: DCP-87-0053-S00-R00- 72
.NPE-89-048- DCP-89-0231-S00-R00 74 -l NPE-89-049 DCP-85-0163-S00-R00 76 5 iNPE-89-050- -DCP-86-0016-S00-R00 78-9; : NPE-891051 DCP-86-0055-S00-R00 80 ;
NPE-89-052: DCP-86-0116-S00-R00' 81 i < , NPE-89-053= DCP-86-0125-S00-R00. 82-um~
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NPE-89-054 DCP-86-0126-S00-R00- 86-
.NPE-89-055' 7DCP-87-0055-S00-R00- 96 ,
NPE-89-056 -DCP-87-0090-S00-R00- -98 (NPE-89-057 DCP-87-4001-S00-R00 100
'NPE-89-058' DCP-87-4018-S00-R01&R02 102 ;
NPE-89-059 DCP-87-0002-S00-R00- 105
=NPE-89-060. DCP-88-0008-S01-R00 108- -;
'NPE-89-061J DCP-88-0008-SO2-R00= 109 'i
- NPE-89-062 .DCP-88-0008-S03-R00 111 NPE-89-063 DCP-88-0008-SO4-R00' 112 NPE-89-064 DCP-88-0008-S05-R00 114 1'
NPE-892 065 DCP-88-0008-S06-R00 116 ' NPE-89-066- DCP-88-0008-S07-R00 118 NPE-89-067 DCP-88-0009-S00-R00 122 NPE-89-068 DCP-88-0012-S00-R00 125' , NPE-89-069- -DCP-88-0036-S00-R00 -127 i rNPE-89-070- MCP-88-1027-S00-R00 129 NPE-89-071 DCP-88-4500-S00-R00 130 NPE-89-072: MCP-88-1052-S00-R00 132- !
,NPE-89-073- .MCP-89-1010-S00-R00 133 ,
NPE-89-074 MCP-89-1045-S00-R00 135
-NPE-89-075 MCP-89-1001-S00-R00 136 NPE-89-076 DCP-84-4013-S00-R01 137 NPE-89-077 DCP-88-0018-S00-R00 138 ;
PLS-89-001 06-ME-1M10-R-0003 140 PLS-89-002 06-RE-1C51-0-0001 141 ' PLS-89-003 FCN-89-003 143 PLS-89-004 05-1-02-III-3, Rev. 18 145 PLS-89-005: Aux. Steam System 147 PLS-89-006 MWO-90989 149 PLS-89-007 MSTI-1F15-89-001-0-S 151
' PLS-89-008 CN-89-097 153 PLS-89-009 FCR-89-005 154
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PLS-89-010 FCR-89-001 155 3 , PLS-89-011 060P-1C71-V-0002-Rev. 24 156 PLS-89-013 06-OP-1F11-V-0001-Rev. 24 160
.PLS-89-014 S.O.I. 04.1-01-E12-1 164 PLS-89-015 CR-89-008 166 PLS-89-016 CR-89-009 177 PLS-89-017 ANSI-N13.5-1972 170 PLS-89-018 ANSI-N42.3 171 PLS-89-019 Temp Alt 89-0003 173 M9082206/SNLICFLR - 2
s--- T 4 k i - TABLE.0F CONTENTS (Continued). SRASH DOCUMENT NUMBER PAGE PLS-89-020 MWO-F86672 176 PLS-89-021 MWO-F84433 17 8 PLS-89-022 MWO-F91-305- 181 PLS-89-023- FCR-89-007 183
- PLS-89-024- MWO-91577 185-PLS-89-025- 07-S-14-339 187
.PLS-89-026 Temp Alt 189
, PLS-89-027 - MWO-E91704 190 PLS-89-028. S.P. 06-ME-1M10-0-0002, Rev. 22 191 PLS-89-029 . Temp Alt 89-0002- 198
- PLS-89-031 MSTI-1P64-88-001-0-S 200 NSP-89-001 ANF-1.3 Reload Fuel- 202' NSP-89-002 Cycle 3 Operation .
207 NSP-89-003- 28 ANF-1.1 8x8 Fuel Assemblies 209 NLS-89-001 SOM Directive No. 1.101, 212 Page 4 of 5; Attachment 1, Page 3 and 4 of 5 Y M9082206/SNLICFLR - 3
3 , m 6 + Attachment to AECM-89/0093 e r {' I SRASNi NPE-89-001 DOC NO: DCP-84-3090-500-R00 SYSTEM: E51-1 0ESCRIPTION OF CHANGE: . This change replaces the meter scale on pane 1Lmeter 1E51-R604 with a meter of the same range but different 1 subdivisions. This meter on panel H13-P601-21B indicates the RCIC- ' pump suction pressure.- The meter is a G.E. type.180 with a scale reading 30 (inches of Hg vacuum) 85 (psi). Scale ~ divisions on the high side (0-85 psi) are 71/2 psi each with. numbered indications at 15, 30, 45, 60, 75, and 80. The new scale has the same-range and numbers on the high side but with increment markers
.of 5 psi each, REASON,FOR CHANGE:' To make the reading of the meter easier.
SAFETY EVALVATION: The scales on the new meter are more readable, resulting in more accurate-readings. Th n iore there is no increase in the probability or ecL6equences of an accident or of a malfunction of equipment important to safety. Since there is no change'to the' meter circuit parameters and no new failure modes are introduced,.there is no possibility of an accident or N malfunction of equipment different from any pre"iously evaluated ; in the FSAR. Sine:e no change to the meter circuit parameters is ! made and no new failure modes are. introduced, there is no reduction 'in the margin of safety as defined in the basis for any Technical Specification.
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, Attachment'to AECM-89/00931
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h ' SRASN: 'NPE-89-002
. DOC NO: CN-89-0037 SYSTEM: C83 j '
p s DESCRIPTION OF CHANGE: - CN 89/0037- installs three new guard towers P and relocates one existing tower at~ strategic points adjacent to
.the security fence. The. sole purpose of these towers is guard posting when directed by Security. Three of the towers (Towers 1, j 2,' and 4) will be: located along the ea' stern perimeter of the fence while Tower 3 will be situated along.the fence'in a location adjacent to the SSW Basin B Valvehouse. Installation will.be-
' coordinated.with Security to ensure the towers are.at acceptable posting locations, do not introduce a means:of bridge access, and do.not-impose on the security fence isolation zone. No Seismic II/I hazards are created by installation of these non-safety p
related towers. The effects of the tower installed proximate to.
- the SSW Basin B Valvehouse on the basin slab and valve room was analyzed and determined to be acceptable. To ensure induced loads '
remain within design allowables, no additional live load will be permitted.in the area encompassed by this tower's legs. The tower that is being relocated (Tower 4) a es not have any combustible materials but Towers 1, 2, and 3 'ncorporate a limited
< quantity of combustible material. The new P.sito combustibles are-p insignificant and acceptable and a minor revision to the Fire
. Hazards Analysis will be generated to reflect the new material.
' Additionally, the 8' x 5' slabs for Towers 1, 2, and 4 were evaluated for impact on site drainage during a Probable Maximum l
Precipitation (PMP) event. These small slabs project only 1" above finished grade and are not in a major flowpath and therefore will have a negligible effect on the existing PMP analyses. REASON FOR CHANGE: These changes were made to enhance plant !. security. 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 evaluated in the Safety Analysis Report. These towers are non-safety related and are not a Seismic II/I hazard. The tower located at the SSW Basin was analyzed for acceptability as to effects on the basin slab and valve room via Calculation CC-NSC83-89006 and found to be acceptabit. UFSAR Appendix 9C provides an analysis of safe shutdown in the event of a major fire. In accordance with 10CFR50 Appendix R and Appendix A to Branch Technical Position APCSB 9.5-1, the combustible materials associated with the new guard towers do not present intervening combustibles between redundant divisions of safe shutdown systems and will not present unacceptable potential fire exposures to safety-related systems. Additionally, the small slabs installed will project only 1" above finished grade and will have a negligible effect on the existing PMP analyses. NLSATTC2/SNLICFLR - 2
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. minimal l-impact with~other plant systs.ns/ components, anatallation i
%, of'these towers'will not' create a possibility,for.an' accident or-
': malfunction of;a different type thanlany. evaluated previously;in- j j
'the Safety' Analysis Report. 1The towers are:not' Seismic II/I. '
12- . hazards:and where'they interact withiother plant components (i;e.,- SSW Slab / Valve: Room),'the affects were' evaluated by' analysis and j
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found to be acceptable. The changes implemented do not create the s i", ipossibility-for'a sir.gle fire' event to adversely affeet more than.
' "" 7 ene division ~of safe shutdown ~and safety-related-systems are
; =sufficiently: isolated from the:affects of potential fire exposure.- {
- !In addition, there is no reduction ~in the margin of safetylasr
- defined.in'the' basis for;any Technical Specification.- !
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y + m y m w, 4 , _C Attachment to AECM-89/0093: + y , SRASN: NPE-89-003 DOC NO: PMI-89-00593 SYSTEM: N/A-m 9
' DESCRIPTION OF CHANGE: This 50.59 is for Plant Staff installation '4 ~of a temporary snubber testing-facility in Area 7, Elevation 166'-0" of'the Auxiliary Building during RF03 work activities.
I E This. temporary wood framed structure will be constructed in , accordance..with Plant Staff procedures and will house the snubber-test bench, control console, computer and printer, work tables. and storage cabinets. .The equipment-layout land general construction requirements have been specified. .In addition to these requirements, the 1000 lb. hoist will be attached to the W24 x 76 identified as Beam B43-1 and the ' snubber lift height will be limited to not more than 8'-0" above-the 166'-0" elevation floor, ,
-Plant Staff will .take precautionary measures to ensure that safety related conduits' attached to Beam B43-1 will not be affected by the hoist attachment. Also, the 10,000 psi pressure gauge restraint and the McMaster-Carr Model No. 5033A3 c-clamps for restraining the two needle valves on the snubber test machine will be securely in place prior to ana during machine operation. The snubber test machine will be securely in place prior to and during .
. machine operation = The snubber test machine will not be operated if any of these restraints are not securely in place.
Power for the temporary equipment to be installed within this facility will be obtained from BOP power receptacles fed from MCC18131 and Lighting Panel IL108. The temporary power feeds (exposed cabling) from the 80P receptacles to the temporary
. equipment shall be-installed in accordance with the separation requirements of Reg. Guide 1.75 as delineated in MP&L-ES-02, Rev. 1. Ampacities of these temporary power feeds shall be in accordance with Article 310-15 of the National Electrical Code and the'added loads shall not exceed branch circuit ratings as shown on drawings E-1072-02, Rev. 23 and E-0658-008A, Rev. 7.
l The temporary structure is to be located in fire area 19, fire L zone 1A403. The following fire protection measures are provided: smoke detection, accessibility to manual hose streams and portable
' fire extinguishers. Fire zone 1A403 contains only Division I safe shutdowa components and adequate separatio.n distance is-provided from Division II safe shutdown components therefore, the requirements of 10CFR50 Appendix R, Section III.G 2 will be maintained. One eight hour emergency lighting unit will be partially blocked by this tamporary structure; however, area walkdowns show adequate emergency lighting will be provided as required by 10CFR50 Appendix R; Section III.J. The temporary structure and its associated components are considered to be transient combustibles and the existing plant procedures will govern. Plant Staff will also coordinate the construction of the structure with the SERI Risk Management Coordinator with respect to adhering to the insurance carrier's (NML) requirements.
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v f ' NPE-89-003. Page 2 REASON FOR CHANGE: This 50.59 is for Plant Staff installation of a temporary snubber testing facility in Area 7, Elevation 166'-0
$ c of the-Auxiliary Building during RF03 work activities.
1
.~ 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 evaluated in the Safety Analysis Report. Fire Zone 1A403 contains only Division I safe. shutdown components with adequate separation provided from Division II-safe shutdown components. Routing of the temporary B0P power feeds per.MP&L-ES-02, Rev. I and proper selection of cable-sizes per Article 310-15 of the National Electrical Code will insure that the separation' requirements of Reg. Guide 1.75 are maintained and that no safety related equipment will be affected by this temporary modification. Installation of the snubber test facility was reviewed for additional floor loadings, the is-ton hoist attachment to existing structural steel, and a load drop over the area serviced by the hoist. Calculation No. CC-Q1T22-87021, Rev. O has been gt.nerated and results show that the structural integrity of the 166'-0" elevation concrete floor i -and the 185'-0" elevation floor steel is not affected by these temporary modifications. Additionally, a hazards review was performed with no unacceptable hazards identified. No safety related systems are modified or affected by this temporary change. Electrical isolation and physical separation of the temporary power feeds in conjunction with proper selection of cable size will insure that accident possibilities -remain bounded by existing analyses. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification. L l NLSATTC2/SNLICFLR - 5
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SRASN: NPE-89-004 DOC NO: CN-89-0049 SYSTEM: N19 j DESCRIPTION OFJCHANGE: During operational transients,.a number of pump trips have been experienced due to low flow trips on the
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condensate pumps, condensate booster pumps, heater drain pumps and
?' feed pumps. These pumps are tripped by the following differential alarm switches:
N1N19FSLK018-Condensate pumps.
~N1N19FSLK028-Condensate Booster Pumps !
N1N21FSLK005A(B)-Feed pumps ! NIN23FSLK046A(B)-Heater Drain pump
- The basis for.the low flow trip is'the prctection of the 1 associated-pump from low flow conditions. However, the present switch settings are too conservative and are causing unwarranted )
pump trips. This CN cancels DCP B5/0066-1 and deletes all work associated with DCP 85/0066 except for the applicable setpoint changess By this CN, the only changes performed by DCP 85/066 are . to increase the' trip setpoints on N1N19FSLK028, N1N21FSLK005A(B), j N1N23FSLK046A(B) and N1N19FSLK018.
- REASON FOR CHANGE: This change increased pump trip setpoints to a value which will minimize pump trips and still allow adequate pump u protection. !
SAFETY EVALUATION: There is no increase in the probability of j occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in.the Safety . Analysis Report. Increasing the differential setpoint at which , the respective pump is shutdown will increase plant reliability l by eliminating unwarranted trips. This will in turn reduce the
. probability of a loss of feedwater flow as addressed by 15.2.7 of g the FSAR. No other accident precursors are affected by this CN.
Increasing the low flow trip setpoints for the condensate, condensate booster, feed, and heater drain pumps has no affect on the ability of the plant to mitigate the consequences of a loss of feedwater flow as addressed by 15.2.7 of the FSAR. No other accident consequences are affected. Revising the sotpoint for the condensate booster pump, heater drain purnp, feedwater pump and condensate pump switches affects the associated pump only and does not affect any other equipment previously evaluated in the FSAR. All existing design features and requirements are maintained by this design change. The switches involved in this DCp affect only the operation of its associated non-safety related pump and therefore do not increase the consequenc,s of a failure of equipment important to safety previously evaluted in the FSAR. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 6 J
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, i Attachment to AECM-89/00935 -i >
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.The affected trip-setpoints are not addressed by the GGNS p
. Technical Specifications.' No new Technical Specification .
D requirements are-introduced by this designLchange. No bases for '
'the margins of safety are affected.
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Attachment'to AECM-89/0093-
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V SRASN: NPE-89-00S= DOC NO: DCP-87-0075-S00-R00 & R01 SYSTEM: .C83 DESC;ilPTIONLOF CHANGE: This change added card readers, access
' switches,' electric strikes and magnetic switches to various doors.
? Security keyed locks are also added. This change also added. security barriers to various HVAC ducts. REASON FOR CHANGE: This change enhanced security.
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[? SAFETY EVALVATION: ,This safety evaluation is classified as ,, r- SAFEGUARDS and is available for review at GGNS. ! i r I i NLSATTC2/SNLICFLR - 8
i , Y"c"i " ". ~" , Attachment to AECM-89/0093 ! w D . +- '
, n SRASN:-LNPE-89-006 DOC NO ; DCP-87-0029-S00-R00 SYSTEM: M30 +
n p DESCRIPTION OF CHANGE: This change provides a handrail at the 1 west end of the' Containment Fuel Pool at Elevation 208'-10". The . handrail is removable and detailed such that it does not interfere with the operation of the Fuel Prep Machines located at the west end of the Containment Fuel Pool, nor does the handrail interfere with the Refueling Platform bridge. REASON FOR CHANGE: The safety handrail is needed for the prevention of personnel-injury from n fall hazard as well as protection from potential personnel contamination, both internal and external. i 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 evaluated in the Safety Analysis Report. The added handrails have been designed seismically to preclude any II/I hazards. The handrail designs are in accordance with applicable codes (UFSAR Section 3.8.3.2) and-used appropriate loads as defined by UFSAR Section 3.8.6 and 29CFR1910, OSHA 2206, Section 1910.23. The addition of handrails will have no adverse impact on safety features or affect the accident analyses listed in the UFSAR. The accidents previously , evaluated-in the UFSAR.were abnormal operation and design basis accidents that involved the possible direct release of radioactive material. The addition of this handrail will not, in any manner,
-directly or indirectly, affect the probability of such accidents.
Materials used for handrails meet the requirements established in UFSAR Table 9.5-11, i.e., structural components are + non-combustible. Therefore, work to be performed by this DCP will not affect the function of any safety related equipment and the probability of occurrence of an accident previously evaluated in the UFSAR will not be increased. The changes to be implemented are not associated with any system or component used in mitigating the consequences of an accident as E analyzed in the UFSAR. The consequences considered for accident analysis are typically related to the boundary performance of the plant to an accident. Accident consequences such as reactor coolant boundary pressure, containment pressure, radiological doses, etc., will be unaffected by the addition of a handrail. Additionally, the handrails will not affect the function of any safety related system and all materials are non-combustible. As such, the consequences of accidents already evaluated in the UFSAR are not increased. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification. NLSATTC2/SNLICFLR - 9
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'SRASN: NPE-89-007 DOC NO: -9645-M-300.2-S01-R00 SYSTEM: NA .
DESCRIPTION OF CHANGE: Specification 9645-M-300.2, Supplement 1 provides.a. tabulation of component support materials which are
- considered'ASME Code exempt, and provides the loading combinations required to' develop normal, upset and faulted loads-for the design of pipe hangers, supports, restraints, and anchors. Loading- ,
parameters have remained unchanged, but combination method has < 4
. been revised from ABS (Absolute Sum) to SRSS (Square Root of the Sum of the Squares). The SRSS method for the combination of dynamic responses has been approved by the NRC in Safety Evaluation Report'(SER), NUREG-0831, Supplement No. 2, Paragraph 3.9.3.
REASON FOR CHANGE: Specification 9645-M-300.2, Supplement 1, was generated to allow the use of Paragraph NF-3392.1 of ASME Section III, Subsection NF,1980 Edition, in lieu of Paragraph NF-3392.1 of-the 1974 Edition. 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 evalucted in the Safety Analysis Report.- This Specification Supplement was generated to clarify materials which are classified as ASME code exempt. This change serves to provide a finite listing of component support materials considered ASME Code exempt ad is in agreement with the material exemptions shown in ASME B&PV Code, Section III, Subsection NF (NF-2121),1974 Edition. Supplement I of Specification 9645-M 300.2 incorporates paragraph NF-3392.1 of the 1980 ASME code which results in a change'in calculation methodology for welded joints for ASrtE Class 2, 3 and MC pipe supports. Class 1 pipe supports are not affected by the use of this paragraph. The 1980 ASME code edition has been approved for use by the Nuclear Regulatory Commission via 10CRF50.55a. The change covered by paragraph NF-3392.1 brings conformance and continuity in methodology to acceptable industry standards; 1.e., l the specification for the Design, Fabrication, and Erection of l Structural Steel for Buildings, February 12, 1969, AISC. The use of the later ASME Code will not reduce code margins, consequently, l' the probability of occurrence of accidents will not be increased. l Additionally, the subject specification supplement addresses required load combinations which must be considered in the design of pipe hangers, supports, restraints, and anchors. Loading . parameters have remained unchanged, but combination method has l been revised from ABS (Absolute Sum) to SRSS (Square Root of the I i Sum of the Squares). The SRSS method for the combination of , L dynamic responses has been approved by the NRC in Safety Evaluation Report (SER), NUREG-0831. Supplement No.2, Paragraph '~ 3.9.3. Therefore, the use of Supplement 1 to Specification 9645-M-300.2 will not affect the function of any safety related equipment. Therefore, there is not creation of a possibility for an accident or malfunction of a different type than any evaluated l previously in tha Safety Analysis Report. l ! NLSATTC2/SNLICFLR - 10 i 3, e ,- - ,. .w--
y :./
}
s
.Attcchment to AECM-89/0093; r
b
- b, ; . .
7; -NPE-89-007 , page 2. :l p The'use of Design Specification 9645-M-300.2, Supplement-I will p not directly or indirectly affect any design basis failure points 'F or Technical ~ Specification safety lindts, i.e., the margin of.
~
L " safety is-unaffected.' Designs based on the subject specification I -- will be in compliance with the requirements of the ASME Code and
=
the_ Code margins will be maintained. Therefore, the use of Design L Specification 9645-M-300.2, Supplement 1, Revision 0, will not I introduce any'new hazards or reduce the margin of safety as
' defined in the basis for any Technicsl Specifications.
l I b F NLSATTC2/SNLICFLR - 11
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1 Attschment to AECM-89/0093- ~ [p g ' N , w , J .. f SRASN:--NPE-89-008- DOC NO: SCN-89-0002- to ES-19, Rev. 6 SYSTEM:- N/A
) [ DESCRIPTION'OF CHANGE: The SCN=to ES-19 involves justifiable p "
1 deletion of E32N059 from the.ES-19 Appendix A equipment list. The list of Electrical equipment required to be environmentally qualified was formerly maintained as-UFSAR Table 3.11-7. As noted
-in the UFSAR Section 3.11.1, this list has been moved'to the EQCF (out~of:the UFSAR). ES-19 is now the controlled environmental qualification master list.
I 'This' equipment list is a subset of the GGNS master equ'pmenti list. This subset must be environmentally qualified in accordance with 10CFR50.49 unless it was previously qualified in accordance with NUREG 0588. Recently, SERI developed-the shutdown logic diagrams
,and safety function diagrams (SLD/SPD Project) in order to document the accuracy of the equipment listing in ES-19.
~
The original listing had been developed by the GGNS A/E and NSSS
~
suppliers. The SLD/SFD project developed logic diagrams which depict all systems required to prevent or to mitigate the consequences of an accident that could result in potential offsite exposures comparable to the 10CFR Part 100 guidelines, bring the
. reactor to cold shutdown, and maintain it in a safe shutdown
' condition. Those logic diagrams also show all necessary Electrical equipment in each system and categorizes the equipment function for each of 11 accident scenarios analyzed in the UFSAR Chapter 15 within the scope of 10CFR50.49.
1 REASON FOR CHANGE: Many items were found to have been conservatively included in the equipment list (for 10CFR50.49) that can be justifiably deleted. The equipment that is being deleted by this-SCN to ES-19 is not the total list of items
. identified as candidates for removal, but it has been addressed in this safety evaluation with validated arguments supporting removal from the list.
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 evaluated in the Safety Analysis Report. The SLD/SFD project analyzed each 10CFR50.49 accident and developed a list of items which must function to mitigate those accidents. Deleticas of equip e nt not in i.he SLD/SFD list from ES-19 cannot increase the probability, since
~,
they are not modified and failure of these items will not affect other equipment important to safety. Items required for operator inforrmtion (UFSAR Tables 7.5-1 & 7.5-2) are not affected by this change. This equipment deletion will not affect the sequence of events tables for any analyzed accident nor will it affect other equipment required to be operable. Equipment exposed to harsh environment conditicos are included in the equipment list in ES-19. Those items not exposed to harsh environments or not required to function for mitigation of the analyzed ( cidents and whose failure is deemed not detrimental to plant safety or accident mitigation need not be listed in ES-19. NLSATTC2/SNLICFLR - 12
. -. p A
'O' <
Attachment to AECM-89/0093 n NPE-89-008 l'
.Page,2.
Deleting this. item from ES-19 will not increase the probability of. failure of any required equipment items.
'No-physical modifications are being made. The changes'being made are deletions-to the-GGNS 10CFR50.49 equipment list (ES-19, Appendix A). .The criteria of 10CFR50.49 are still mat and a; maintained at GGNS. Therefore, there is no creation of a possibility.for an accident or malfunction of a different type 1 than any evaluated previously in the Safety Analysis Report.
Also,.there is no reduction in the margin of safety as defined in
.the basis'for any. Technical Specification.
i L I, Y NLSATTC2/SNLICFLR - 13
7'i y s y Atttchment to AECM-89/0093 4
'SRASN: .NPE-89-009 -DOC NO: MCP-88-1019-S00-R00 SYSTEM:- H31 DESCRIPTION OF CHANGE: The Tritium Monitoring System detects
, leakage in the primary water circuit inside the' generator housing
, that could result in major damage over a period of time. For this purpose, the primary water is enriched with the' radioisotope tritium. -The tritium monitoring system will alarm when a total J1 tritium beta particle _ indication of 60 ipm (indications per minute)t is detected from background sources plus leakage from the primary water system. A second alarm setpoint is annunciated from the Tritium Monitoring System at 90 ipm.
The automatic turbine trip, which resulted from the Tritium Monitoring System via EGP if tritium beta' particle levels reach. 120 ipm, has been bypassed under Temp. Alt. #86-0037. This temp.
~
alt. removed th6 trip function to prevent spurious main turbine trips. The turbine vendor, UPC, has confirmed that the trip function'on the tritium monitor can be eliminated as long as the alarm function reinains and is monitored closely. This MCP clears Temp. Alt. #86-0037 by permanently disabling the automatic tritium trip function. The Tritium Monitoring System will still provide
.the alarms as described above. Only the automatic trip function
.is deleted by this MCP.
REASON FOR CHANGE: To eliminate turbine trips caused by spurious signals from the Tritium Monitoring System. SAFETY EVALUATION: Disabling of the. Tritium Monitoring System ; trip reduces the probability of spurious generator trips. Because j the Tritium Monitoring System is a generator protective function, i no equipment important to plant safety is affected by this change. Primary water leakage into the' main generator is still monitored and alarmed. Therefore there is no increase in the probability or .l consequences of an accident or malfunction of equipment important i to safety. i No new failure modes are created by this change. The alarm l functions are still intact and continue to be monitored. 1 Therefore, there is no possibility of an accident or malfunction of equipment important to safety different from any previously evaluated. A Since the trip functions of the Tritium Monitoring System are not addressed in Tschnical Specifications, there is no reduction in the margin of safety as defined in the basis for any Technical Specification. l NLSATTC2/SNLICFLR - 14 l
~ %v a
H l w
. A, LAtt'achment to AECM-89/0093^
k (<_ ! L U$' u - LSRASN:- NPE-89-010 DOC NO:' DCP-85-4005-500-R00 SYSTEM:- E12 4 7 t DESCRIPTION OF CHANGE: This DCP directs the replacement of the t carbon steel cotter pins in the lifting disk nuts of safety / relief
~
valves Q1E12F055AM with stainless steel cotter pins. These , o ~_ valves provide overpressure protection for the RHR A and B loop heat exchangers during Reactor Steam Condensing Mode operati.ons of : the RHR System (E12)._ , REASON FOR CHANGE: The change'of_ material was made to prevent l C corrosion-induced failure of the cotter pins which might introduce ' y the failure mode, described in IEN 84-33, in which the . x f~ safety / relief valves failed to reset following actuation and reduction of pressure due to the nut turning to a lower position on the valve spindle during actuation. c ; SAFETY EVALVATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of
~
o equipment important to safety previously evaluated in the Safety
- r. Analysis Report. The lifting disk cotter pin is a part of the mechanical lifting device assembly of the safety / relief valvs E 'which does'not perform a function critical to the functioning of r .the safety / relief valve in its role of protecting the RHR heat >
exchangers from excess pressure. The function of the cotter pin i is to hold the nut, which restrains the lifting disk, in place on
, the valve spindle at the point at which it was installed by the .
The manual' lifting device (lever and lifting disk l manufacturer. assembly) is required by'the ASME III Code on air and. steam safety / relief valves so that they can periodically be tested to assure that the valves are free. The lifting device does not control or affect the normal function of the safety / relief valve, and its failure would not prevent the valve from acturting when.
. its setpoint is reached. Replacement of the carbon steel pin with one of the stainless steel improves resistance to corrosion which caused the failure described in IEN 84-33. The change in material does not change the function of the cotter pins. Size (length and
' diameter) remains unchanged from the original design. In any case, failure of this assembly would not cause systems or l components to be operated outside of design limits, decrease system integrity, or prevent operation of the RHR system.
Therefore, there is no creation of a possibility for an accident (i or malfunction of a different type than any evaluated previously I in the Safety Analysis Report. The change does not affect the operation or operability of the system in any parameter which affects the margin of safety as described in the Technical Specifications, such as flow, chemistry, power, setpoint, capacity, level, or pressure, since it leaves the operating characteristics of the safety / relief valve l unchar.ged from the original design. NLSATTC2/SNLICFLR - 15 i
g u :w ,- i r; , " - .y a gwl;
# 1 x Attachment to AECM-89/0093 e
y' + SRASN:1 NPE-89-011 DOC ~NO: EER-89-6066- SYSTEME N/A-DESCRIPTION OF CHANGE: The'.~ performance of an ISI weld inspection i , fon.the Recirculation System piping: requires that'several, pipe y supports be temporarily-disassembled to allow: inspection access to
. pipe. welds. This inspection is to be performed dur_ing ' plant outage during- a cold: shutdown (Operat.ing Conditions' 4 and 5).
^
Eisht. snubbers > and two spring hangers on recirculation loop B are to be. removed, iStress analysis ofLthe recirculation piping with
" E- 1' these supports removed has shown that the structural integrity-of the Recirculation Piping system will be maintained in the unlikely event of'an operating basis earthquake (OBE) or a safe shutdown.
earthquake (SSE). -All applicable ASME Code stress allowables are s
. met. Therefore,,the recirculation system operability in Operating
-Conditions 4 and 5 is not affected by the temporary removal-of pipe supports.
1 Based on th'e'above analysis, the following snubbers may be
^; * . removed: 1833G006S301B, 302B, 305B, 3608, 3618, 3628 363B, and
, i
'376B from loop B during Operating Conditions 4 and 5. I n' 4
addition, spring hangers 1833G002H355B and 356B may also be. n: . removed and replaced with a temporary support as described in PMI 86/06833 during'those operating conditions under the condition . ? ; that only one spring. hanger is not operational at' a time. Also, ~! the change >in temperature'of the-system and-the reactor'does not exceed 50 MF during the time the spring hanger is removed. ? Temporary removal of' snubbers and hangers does not result in any permanent changes-to location,-routing, or type of. supports, nor > does 1.t alter any component performance characteristics, design parameters, or operational parameters of the affected system after the affected s'upports are reinstalled, m y REASON FOR-CHANGE: -This change. allowed access for weld. I inspections during RF03. I 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 evaluated in the Safety Analysis' Report. Since all applicable ASME codes allowable stresses'are met, the probability of occurrence of an accident resulting from a seismically initiated pipe break is not increased. There will be no change to existing designs after the g supports are reinstalled. These temporary changes do not affect the structural integrity of the recirculation piping during cold shutdown. Structural integrity of the recirculation piping system has been confirmed with snubbers removed for Operational Conditions 4 and 5. There are no permanent changes made to B existing designs after the affected supports are reinstalled. L Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 16 l e e .- . . - . . , ,
{ s]. . , f, , Attachm2nt:to AECM-89/0093 F , NPE-89-011 r- Page.2-i; ' Structural integrity of the recirculation' piping system has been K cc, ' : confirmed with snubbers ' removed for Operating Conditions 4 and 5.
'.'.,~ Removal.of:these supports temporarily does not change the limiting
~ ' ' '
conditions.for. operation', applicability..or surveillance requirements as defined in the basis for the-Technical. Specifications,. h N k i [f i I ). 4 5
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NLSATTC2/SNLICFLR - 17
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I L. a
Tw. ,w < j Ii N Attachment'to AECM-89/00934 L #
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. " 9
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! N ~
SRASN: NPE-89-012: DOC NO:- CN-89-0136 SYSTEM:- F13 1
, a ,
i
- v. .
j k' ' DESCRIPTION OF CHANGE:L.MCP-89/1035, Rev. O, required removal,or; 1
; an' acceptable retest of unqualified exposed paint surfaces on-the 1 LContainment Dome Access Lift (CDAL) prior to use on the polar
~
j; crane-iii Plant Operating Modes 1, 2 or.3. CN 89/0136 permits 1
. approximately-60 f t' of unqualified UFSAR 3A/1.54 Category A coating-to be temporarily introduced to containment. -Prior to implementation of the MCP and CN, the coating was classified as Category C. Category A means that a potential post-LOCA debris path to the suppression pool' exists and Category C. indicates no such path exists. Prior to installation of the CDAL on the polar crane, the coating is Category C since all deteriorated coating is :
routed to the drywell: head- storage area floor drain. When , installed on the polar crane, the coating is Category A since deteriorated coating can be postulated to have a path to the i
~l suppression pool.
Per the CN, a11' loose scale was removed from the CDAL but remaining surfaces could not pass adhesion pull testing._ The CDAL will' be used for a .short period during plant operating modes 1, 2, and'3 prior to RF03. After RF03, the CDAL cannot be used intthese modes without a qualified coating in place. The. temporary ! increase in the unqualified paint surface total noted in UFSAR Section 6.1.2.1 is less than 1%. si, REASON..FOR CHANGE: This change permitted the evaluation of 60 sq.
.ft of. unqualified UFSAR 3A11.54 Category A coating on the Containment Dome Access ~ Lift.
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 evaluated in the Safety Analysis Report. Coatings are not postulated to be an initiator j in previously evaluated accidents. Considering the transient u existence of the unqualified coatir.g and its insignificant amount, j the justification of these surfaces provided in UFSAR 6.1.2.1.b is i L still valid. This-section of the UFSAR does not consider-the
-failure of unqualified paint surfaces to be a significant i L contributor to strainer clogging. ECCS suction strainers have !
l' been analyzed for performance assuming a bounding condition taken l- as 50% strainer clogging due to postulated suppression pool debris i. from any source during a LOCA. The additional surface area of ll unqualified paint is insignificant with respect to this limit.
. The only equipment affected by a failed paint surface during a L LOCA are ECCS suction strainers. As such, the possible paint failure is only related to strainer performance capabilities and E not to mitigation of the consequences of equipment malfunction.
l l Paint surface failure is not postulated to occur until after the L occurrence of a LOCA. These effects have already been evaluated. NLSATTC2/SNLICFLR - 18
y q <e , , y:
, i AttCchment t3 AECM-99/0093-a
.ib.
s- g-NPE 89-012-Page 2i The' failure mechanism and effect of the subject paint surface is 4 not unique from other previously postvlated failures. Therefore, an accident of a different type than already evaluated in the
'UFSARLis not created. The paint surface failure is not unique
-from previously postulated failures. Therefore, no new equipment
-malfunction is postulated.
;, The ECCS strainers were evaluated for being 500 clogged due to postulated suppression pool debris from any source during a LOCA.
The edditional, temporary, unqualified surface area of paint is
- s. 'less than 1% of the present total unqualified paint surface area.
In addition, UFSAR Section 6.1.2.1.b does not consider the unqualified paint' surfaces to be a significant contributor to strainer. clogging. .Therefore, th'ere is no postulated affect on ECCS strainer performance and the mL:)in of safety is not reduced.
+
l NLSATTC2/SNLICFLR - 19 1 l =
w1
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Atte.chment to AECM-89/0093 !
.s., l 7
L < SRASN: NPE-89-013 D00 NO: MCP-88-1050-500-R00' SYSTEM: N71
+ i
" (
.. . DESCRIPTION OF CHANGE: .The design Installation Detail specified i P
4
for the installations of IN71-LI-R006A/B (LP condenser water box-inlet level indication) and IN71-LI-R007A/B (HP condenser water 1
+
box outlet level indication) assumed ihat the instrument reference j w leg tubing only contained air. 'Because of.the location of the-reference leg sensing port an accumulation of water has partially filled the reference leg causing the_ instrument to indicatt. downscale against the "0" stop. There is no practical method for ! preventirg this water accumulation. l As described by MNCR 0161-88, the instruments are not required for , 1 system operation._ Plant Staff, with Nuclear Plant Engineering .; concurrence, has dispositioned the MNCR to valve out the : instruments and remove them from service. These instruments serve ;
, no safety function and their removal will not af feet any design '
requirement (e.g., soismic design of supporting equipment). The operational requirement to determine when. the waterboxes are full is satisfied by use of Circulating Water System Operation - Instruction 04-1-01-N71-1 using instruments identified by MPL ! Numbers 1N71-LSH-N023A(B;, IN71-LSH-N024A(B) and . 1N71-LSH-N025A(B). .. REASON FOR CHANGE: The instruments were valved out because they ' were not accurate. 1 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 evaluated in the Safety , Analysis Report. These instruments only provide water level ; indication of the LP condenser water box inlet and HP condenser water box outlet. Removal of this redundant indicating ,
. instrumentation can in'no way contribute to the probability of i occurrence of any accident. These instruments serve no accident ,
Q mitigation function.- All required design functions are maintained by their removal. The instruments to be removed are water _f ! actuated differential pressure gages. The condenser water level . L monitoring function they serve is more reliably provided by other instrumentation. Their removal cannot contribute to any other i l l equipment malfunction. The instrument installations do not - interact with any other equipment or instruments. All necessary . design features and requirements are maintained even with their removal. Therefore, their removal cannot contribute to the consequences of an equipment malfunction evaluated in the FSAR. ! Removal of the instruments does not include removal of the associated closed isolation devices (i.e. root valves, manifold o valves, tubing caps, etc.). Therefore, no new accident precursors and no new accident types are created by this design change. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. l NLSATTC2/SNLICFLR - 20
ci ; $ [,* $ p. n 1- g Attachment to AECM-89/0093 e ;
- 4
.i h,s t i t g ;)' - L- ' NPE-89-0131 4 Page 2 Y^ No Technical Specification margin of safety is based on the ;
i
., ,,, condenser water box level indication provided by this instrumentation. All existing margins of. safety and'their bases g ,
are unaffected by this design change. . I' . .l' ' 3 'Oi ' u ?:; I, B t,[ 6 l !i r. lb t s
?
l NLSATTC2/SNLICFLR - 21 ,
.mim I.,
r c
\
.i Attachment to AECM-89/0093 1 l
SRASN: NPE-89-014 ' DOC NO: DCP-87-4007-500-R00 SYSTEM: N22 : M* . DESCRIPTION OF CHANGE: DCP 87/4007 modifies existing pipe support NIN22G002R05 located on'20"-HBD-760. The design-load capacity of < the subject support is increased by this change. The piping is
, ' the discharge piping. from the precoat filter (N1N220004A, B, and .
I C) to the condensate clean waste tank (N1P45A003). The subject pipe support sustained damage as a result of a water hanner, L identified in MNCR 1183-86. The affected system is the Condensate L Cleanup System (N22). As discussed in UFSAR Section_10.4.6.3 the ! b Condensate Cleanup System provides no safety function. This - design change does not affect the operation, function or the : l design parameters of the affected system (N22). The change does ! not affect other systems including any safety related systems. t L This change is-located in the Turbine Building and no new failure , modes are created. The modified support meets all code requirements applicable to this system. REASON FOR CHANGE: To increase the design load capacity of existing pipe support NIN22G002R05. 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 evaluated in the Safety . Analysis Report. This design change will modify pipe support NIN22G002R05 to' increase its design capacity. The. Condensate ; Cleanup System's operation and function will not be altered. This : system has no safety related function as discussed in ; UFSAR 10.4.6.3. The design change will not compromise any safety related system or components. The pipe support supplied by this l DCP meets all applicable design requirements and will function in its intended manner. The modif'ications performed are in accordance with the applicable l codes and design standards. Therefore, there is no creation of a , L' possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The change does not alter the existing operational or design l- parameters of the affected Condensate Cleanup System (N22). There l- is no safety related system affected by this change. The proposed L change does not modify any equipment used in mitigating the consequences of an accident as analyzed in the UFSAR, nor does it affect any actions taken to mitigate an accident as analyzed in the UFSAR. The pipe support supplied by this DCP meets all L- applicable design requirements. This change does not change the limiting conditions of operation, ' applicability or surveillance requirements as defined in the bases for Technical Specifications. Therefore, the margin of safety is not reduced. l NLSATTC2/SNLICFLR - 22 l L
. - ~ _. . - , __ _ _ ._ _
a Attachment to AECM-89/0093 r SRA$N: NPE-89-015 00C NO: DCP-84-0134-500-R00 SYSTEM: G17 L DESCRIPTION OF CHANGE: At present both the " LIQ RADWST SYS
# TROUBLE'! (G17-VA-L601) and ."CHEMWST EVAP TROUBLE" (G17-UA-L603) annunciators alarm in both the main control room and in the radwaste control-room. Since the radwaste control room is now continuously manned, the Operations staff feels that these annunciators (G17-UA-L601 & L603) are unnecessary and a nuisance in the main control room (Panel SH13-P854). To resolve this problem, both annunciators G17-UA-L601,($H13-P854-01) and G17-VA-L603 (SH13-P854-E1) are removed.from Panel SH13-P854, and ,' their associated wiring determinated in Panels 1H13-P735 IH13-P851, SH22-P317 and SH22-P128. The existing annunciator windows on.H13-P854-01 & El are replaced with a blank mylar window. This design change incorporates both DCR 84/0134 and L 84/0135 into a single package.
REASON FOR CHANGE: These alarms can be removed from the main control room without loss of information. . Problems occur when alarms distract operators unnecessarily during plant operations. Nuisance alarms violate good human factors principles as defined in engineering standard ES-17 and NUREG-0700. 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 evaluated in the Safety Analysis Report. The DCP removes.two multi-input non-Q annunciators which have annunciatic'n in the radwaste control room. The G17 system function and operation is not affected by this DCP. E The panel combustible loads are not affected by this DCP. These modifications will decrease the probability of human error while operating plant equipment, due to nuisance alarms in the main control room. The change will not affect any seismic and
- environmental as well as safe shutdown and fire protection L criteria for these panels.
l l The changes remove two non-Q multi-input annunciators on L SH13-P854. Redundant indication of each annunciator exists in the l radwaste control room. The changes are designed to decrease the i probability of human error while operating plant equipment. This is done by reducing the number of " nuisance" alarms which the operators must monitor. Therefore, there is no creation of a possibility for an accident or malfunction of a different type o than any evaluated previously in the Safety Analysis Report, L The changes are designed to increase the operator margin of I' safety, by reducing the probability of human er or while operating plant equipment. The Seismic qualification, safe shutdown, fire protection and environmental criteria are not affected by this design change. NLSATTC2/SNLICFLR - 23
4 a Attachment to AECM-89/0093
+ ,
SRASN: NPE-89-016 DOC NO: MCP-88-1029-500-R00 SYSTEM: P53 ; i DESCRIPTION OF CHANGE: The' Instrument Air System (P53) is utilized to' charge sixteen accumulators and four air receivers serving the safety relief valves designated as the Automatic
< Depressurization System (ADS). Prior to penetrating the l containment, the. instrument air supply pressure is increased to the higher operating pressure of the ADS pneumatic supply (160 ;
psig nominal) by Booster Compressors N1P530002A/B. The air supply . f< = then passes through a 3-micron filter, N19530024. Downstream of , L the filter, the containment isolation valve, Q1P53F003-A, is the i t
"Q" piping boundary. All of the instrument air piping upstream of .
L the containment isolation valve is non-safety related, designed in' o accordance with ANSI B31.1, and qualified as seismic category II/I. t Instrument air filter, D024, is provided with two. isolation valves. The filter bypass line also contains a single isolation > valve but does not contain a filter. In the existing installation, the piping material specification for the i
. filter / filter bypass loop changes to stainless steel at the ;
upstream filter isolation valve and et the filter bypass valve.
, When filter D024 requires replacement, the filter bypass valve is opened in order to provide an uninterrupted air supply to the ADS. ;
.The carbon steel filter bypass valve, N1P53F482, is considered a possible source of particulate contamination to the instrument air system and the ADS.
In order to eliminate this possible source of particulates, MCP I 88/1029 removes the existing filter bypass valve. In its place, c MCP 88/1029 will install a 3-micron filter and stainless steel isolation valves in the filter bypass line of the instrument air i supply to the ADS air receivers and accumulators. In addition, the piping material specification change will be relocated : l upstream of the filter / filter bypass loop, so the upstream filter l 1 solation valve will also be replaced with a stainless steel . valve. l L REASON FOR CHANGE: The carbon steel filter bypass valve, ! N1P53F482, is considered a possible source of particulate contamination to the instrument air system and the ADS. MCP 88/1029 removes the existing filter bypass valve in order to 9 eliminate this possible source of particulate and installs a 3-micron filter and stainless steel isolation valves. SAFETY EVALUATION: There is no increase in the probability vf occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety ' Analysis Report. Loss of instrument air as an initiating event is evaluated in UFSAR Section 15.2.10. In the evaluation the assumed failure is a pipe break in the instrument air supply. MCP 88/1029 l NLSATTC2/SNLICFLR - 24
^ ' '
-e _
~
h Attechment to AECM-89/0093 p l-
'NPE-89-016=
.Page 2
[ will add a filter and two isolation valves to the instrument air filter-bypass line and replace an existing isolation valve. This t- modification.will improve the quality of the instrument air supply i; to the ADS; however, it will not alter either the design function
or the operation of the original system. The piping system modification has been designed to ANSI B31.1 requirements and is qualified as seismic category II/I. The piping modification meets the design codes and standards of the existing installation.
f- UFSAR Section 15.2.10.3.3 states that failure of instrument ^ air will not interfere with the safe shutdown'of the reactor. UFSAR Section 15.2.10.5 further states that there are no radiological L consequences associated with.this event (loss of instrument air); i.e., the instrument air system is not required to mitigate the consequences of an accident. UFSAR Section 9.3.1.3 states that the instrument air system has no safety-related function and that failure of the instrument air system will not compromise any safety-related system or component and will not prevent safe reactor shutdown. UFSAR Section 7.3.2.14.2 further states that the loss of plant instrument air will nct negate any Engineered Safety Feature (ESP) system safety function. This change will modify components in the filter / filter bypass loop of the instrument air supply to the ADS. The filter bypass line is in use only while instrument air supply filter D024 is changed. Addition of a filter in the bypass line will improve ; the quality of the instrument air by providing the capability for an uninterrupted, filtered instrument air supply to the ADS. , Therefore, there is no creation of a possibility for an accident L or malfunction of a different type than any evaluated previously in the Safety Analysis Report. l- There is no reduction in the margin of safety as defined in the basis for any Technical Specification. Loss of the Instrument Air System will not interfere with the safe shutdown of the reactor, and the Instrument Air System is not required to mitigate the consequences of an accident. Since the Instrument Air System is not credited for any mitigative measures, the margin of safety is not reduced. l? s NLSATTC2/SNLICFLR - 25
=- , ,
Attach:ent to AECM-89/0093 i E i NPE-89-017 000 NO: DCP-83-0568-500-R00 1 SRASN: SYSTEM: C11 )1 1 DESCRIPTION OF CHANGE: This DCP replaces existing CRD Hydraulic i
., Temperature recorder system, C11-R018,. manufactured by Tracor "1 W Westronics.,with a new Tracor-Westronics multi-point temperature )
recorder system. Each of_ the systems consist of a multi point 1 L recorder and multiplexers. The replacement system is to be i L located within the same panel / location as the existing. system, p panel 1H22-P007. The new system will utilize all of the same
; thermocouple inputs'as the present system utilizes. The basic ,
function of the new recorder system is identical to the existing 1 system. The new recorder system performs all of the same design l L functions as the existing system as stated in Chapter 9-Section L 4.6.1.1.2.4.2.4 of the GG UFSAR, with enhanced capabilities. The enhanced capabilities of the new system allows personnel to detect ' further CRD high temperature alarms prior to the clearing of : existing CRD high temperature alarms. No new T/C inputs are added ! by this DCP nor are there any existing T/C inputs or alarm i setpoints or alarm functions deleted by this DCP. Only existing field wiring is to be utilized by the new system with no additional loading of the tield instrumentation or power circuits , above that existing with the present system. The existing , recorder system and the new recorder system perform recorder functions only-with a high temperature alarm output annunciated in ! the contro'l room. The system performs no safety function nor does it effect any safety function addressed within the GG FSAR. REASON'FOR CHANGE: The new system enhancements are, state of the , art technology, individual point alarm (3 alarm point capability i per channel) and Individual channel alarm acknowledge. The , capability of the new recorder system to allow the acknowledgement of individual alarms within the recorder system was the initial , requirement for the DCR. The new recorder system will also - present the recorder data and displayed data in a more useable i format, providing users with easier to read and more precise data for reference and analysis. 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 evaluated in the Safety l Analysis Report. Implementation of the DCP being evaluated .
replaces an existing recorder system with a new electronic, state l of the art recorder system that will perform the identical system design functions as the existing system, with improved reliability and enhanced capabilities. The new recorder system will also i present the recorded data and displayed data in a more useable L format, providing users with easier to read and more precise data for ieference and analysis. Malfunction or failure of the CRD Temperature recorder system does not initiate, contribute to or - otherwise impact accident analyses as previously evaluated in FSAR Chapter 6 and 15. L NLSATYC2/SNLICFLR - 26
7 : .--
- Attachment to AECM-89/0093 L
h
, NPE-89-017 Page 2 Implementation of the DCP being evaluated does not introduce any new system design functions or reduced / increased alarm set points.
The CRD Temperature recorder system is not utilized to prevent or mitigate-the consequences of any accident as previously analyzed
~ - in FSAR Chepter 6 and 15.
[L Implementation of the DCP being evaluated replaces an existing recorder system with a new electronic, state of the art recorder i system that will perform the identic-s1 system design functions as
! the existing system, that of recording control rod drive temperature and providing annunciation in the control room for high temperature, with improved reliability and enhanced capabilities. The data recorded and displayed by this recorder Y
system is utilized by plant personnel in the evaluation of control rod drive temperatures. The new recorder system will present the recorded data and displayed data in a more useable format,- providing users with easier to read, more precise data for reference and analysis. The CRD temperature recorder system has no interface with equipment important to safety that could create a situation whereby a failure or malfunction of the CRD temperature recorder system could cause a credible failure of equipment important to safety. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The CRD temperature recorder is not presently considered as essential to provide Technical Specification Bases with respect to safety margins. The replacement recorder will function and connect to the existing plant system in such a manner that malfunction or failure of this equipment will not affect the existing Technical Specification bases for safety margins. l' Therefore the implementation of this DCP will not reduce the margin of safety as defined in the bases for any Technical l Specification. l 1 l l 1 NLSATTC2/SNLICFLR - 27
g- , Attachment.to AECM-89/0093! q i i SRASN: NPE-89-018 DOC NO: Temporary lead Shielding SYSTEM: N/A j Request ! DESCRIPTION OF CHANGE: The Reactor Water Cleanup (RWCU) system j piping in the drywell will have lead shielding installed during , Operating Conditions 4 and 5 only and must be removed prior to j restart.- Calculations were performed on the subject RWCU piping l with.the added weight of the lead shielding. These calculations j show that the structural. integrity of the RWCU piping system with i the temporary-supports will be maintained in the unlikely event of l an operating basis earthquake (OBE) or a safe shutdown earthquake (SSE). All applicable ASME Code stress allowables are met, Therefore, the RWCU system operability in Operating Conditions 4 -! and 5 is not affected by the temporary lead shielding attached to. j the pipe. ; h Based on the above analysis, the temporary lead shielding may be i installed on the pipe during Operating Conditions 4 and 5. ; However, two temporary dead weight supports must be installed on ; the system before the lead shielding is added and cannot be ; removed until all the shielding is removed. Temporary addition of lead shielding does not result in any , permanent changes to location, routing, or type of supports, nor does it alter any component performance characteristics, cesign parameters, or operational parameters of the affected system after ; the: temporary lead shielding and temporary supports are removed. i REASON FOR CHANGE: The Reactor Water Cleanup (RWCU) system piping in the drywell- requires lead shielding in order to reduce radiation exposure to personnel performing work in this area. ; 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 evaluated in the Safety Analysis Report. All applicable ASME Code allowable stresses are l met. 'There will be'no change to existing designs after the lead shielding is removed. Since these temporary changes do not affect the structural integrity of the RWCU piping during cold shutdown, the consequences of an accident previously evaluated in the FSAR , are not increased. Structural integrity of the RWCU piping system 2 has been confirmed with temporary lead shielding for Operating Conditions 4 and 5. Therefore, there is no creation of a possibility for an accident or malfunction of a different type r than any evaluated previously in the Safety Analysis Report. Installation of, lead shielding temporarily does not change the limiting conditions for operation, applicability, or surveillance requirements as defined in the basis for the Technical Specifications. Therefore, the margin of safety is not reduced. 1 l! NLSATTC2/SNLICFLR - 28 h -_ __ _.._
r . ; y Attachment to AECM-89/0093 ! t: l d SRASN:- NPE-89-019 DOC NO: EERR-89-6103 SYSTEM: F41 ) DESCRIPTION OF CHANGE: The UFSAR lists the equipment used in the j Reactor Internal Vibration Monitoring System, along with the i location of the equipment inside the reactor vessel. A partial i description of this startup test equipment is provided in GE i Specification 21A3854, which states "it is intended that the j equipment above the shroud support plate and above the core support plate be removed during the first refueling outage (RF01)." 2 A study was completed in 1986 which provided a basis for leaving a ! portion of this equipment in place during Cycle 2. Accordingly, I some of.the removable equipment was left inside the reactor at the ; end of RF01. A study in 1987 cited justification for leaving all : of the remaining equipment in place during Cycle 3. Some selected ! items were removed during RF02, which concluded in January 1988, . i The following items remained inside the GGNS Unit I reactor vessel, above the core and shroud support plates, at the start of ; RF03: t
-- four accelerometers & their leads, installed on the shroW surface
-- twelve strain gages & their leads, installed on the jet ,
pump riser braces :
-- clips, couplings, tube guards, shield blocks, and conduits ,
used for sensor leads ,
-- supports and brackets for conduits General Electric performed an evaluation (via Engineering '
Evaluation Request 89/6103) of the allowable residence time for t the above listed components. The NPE response to the evaluation states that the minimum requirements for RF03 is removal of the leads from the strain gages and the accelerometers. The remaining removable equipment is to be taken out during RF04. It should be noted that the clips used for the sensor leads are designed to part in the middle when the leads are pulled out, and are then lef t in the reactor vessel permanently. Removal of a clip is required only if it breaks away improperly during withdrawal of the leads. REASON FOR CHANGE: This evaluation determined what removable equipment could remain within the reactor vessel during Cycle 4. 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 evaluated in the Safety Analysis Report. The only mechanism which has been postulated to NLSATTC2/SNLICFLR - 29
1 Attachment to AECM-89/0093 b *1 i l' NPE-89-019 Page 2
. 4 f cause the subject; equipment to come loose from its mountings and l
, interact with other reactor systems is Stress Corrosion Cracking. 1 An engineering evaluation by GE has determined that this mechanism ,
is not a credible event for the subject-equipment during Cycle 4. a The presence of the subject equipment within the reactor vessel in Cycle 4 will have no impact on the response of the plant to any of the analyzed accidents. There is no credible mechanism to force any of. the subject parts off their_ mountings; therefore, these : ' parts will not play a role in any evaluated plant response. , This equipment poses no concern ' s interference with control rod operation or fuel performance, because it has been shown that the case where any of these parts come loose and circulate into the reactor vessel is not a credible event. Since it has been shown-that the subject equipment will not interact with any reactor system, leaving it in place for Cycle 4 i will have no impact on the plant response to the malfunction of any safety equipment.
.There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. It was shown that the only conceivable mechanism for detachment of this equipment (Stress -
Corrosion Cracking) is not a credible event during Cycle 4, and
..* the reactor coolant chemistry will not be affected by this equipment due to the use of stainless steels suitable for service inside the reactor vessel. Due to the fact that this equipment has been shown to have the durability to remain mounted in place ,
during Cycle 4, there will be no impact on the operation of any plant equipment. L Because the subject equipment no longer serves any function and i does not interfere with the function of any other system, leaving the equipment in place for Cycle 4 has no impact on the GGNS Unit 1 Technical Specifications. L l 1 L l l L NLSATTC2/SNLICFLR - 30 l
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'SRASN:':NPE-89-020'- DOC NO: DCP-84-3016-S00-R00 SYSTEM: H13 a,<.v l- s ,
i LDESCRIPTION OF CHANGE: This evaluation ~ verified engravings, gg j ge r" connections, and signal: separation in. control. room, pan 31st
; ',,3.' ;1H13-P843, 1H13-P801, and_SH13-P854. No procedure changes are 1
9 ,,
; required.-
v ,
-REASON FOR CHANGE:- To ensure the subject control' room panels neet !
the' codes per engravings, connections,'and separation.- [ '.c - SAFETY EVALUATION: There is no: increase in the probability of l occurrence or in.the consequences of an accident or malfunction-of-
^ >
, ' equipment important to safety previously. evaluated in the Safety !
o, rAnalys'is Report. LThe original design'is not affected. These j subject components have no. safety function.- Therefore, there is 1 I no' creation of a possibility for an accident or malfunction of a
, different. type than any evaluated previously in the Safety-Analysis Report. Also, there is no reduction >in the margin of.
. . safety as defined in the basis for any Technical Specification.
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;L , , s-3 7 l DOC NO: MCP-89-1050-S00-R00 & R01 LSRASN:- NPE-89-021 SYSTEM:' B33 e .
yh ! DESCRIPTION OF CHANGE:' A 2"; drain line is provided for each of the: reactor recirculation loops to allow the recirculation piping
~F..
- to be drained for maintenance. The two drain lines, which are 1
>>s J1ocated in;the drydl, are designed tol the reqairements of ASME 3 C Section-III, Class 1'up to the'second isolation valve. After.the .
'second~ isolation-valve,;the applicabla construction code is ANSI "B31.1. The lines terminate into flocr drains'which are connected. '
to the radwaste system. The two affected line numbers are 2"-HCD-69.(the same number is used for both: lines). , y 3 '
% : The. drain lines will be' capped downstream of the second' isolation valve.' The: capping will be performed by removing a section out of
. the non-safety related portion of pipe and welding s'2" pipe cap Lto replace the removed _section. The pipe rap.will be welded to
- each:end of the cut pipe to return it to the original
~
' configuration. By performing the capping in this manner, the
. existing pipe support and design analysis are w t adversely affected.
%, REASON FOR CHANGE: This change provides aGditional assurance that leakuge will not occur.
m
- SAFETY EVALUATION: There is no increase in the probability of occurrence or.in the consequences of an accident or malfunction of L
- f. equipment important to safety previously evaluated in the Safety #
Analysis Report. The addition of the items to cap the lino has
- been evaluated and shown not to' affect the structural $ntegrity of uthe safety related and non safety related portions,of the piping.
This evaluation has.shown that the increase in the pressure that may bo seen by the' piping downstream of the F052A/B valves is 7 acceptable.1 Since all-applicable code allowables are met, the probability of an accident resulting from a seismically initiated pipe break is not increased. . The piping systems will function in their intended manner. The piping has been evalusted and shown to maintain structural integrity during various plant loading conditions including earthquake.and other dynamic loadings. -The existing pipe break analysis for this piping is not affected by
- capping the lines. The drain lines are normally closed during all
~ modes of operation. The capping of the linos will provide additional assurance that leakage in the drein lines will not occur. The drain lines are not included in the mitigation analysis for any equ?pment important to safety nor do the closed drain lines alter the performance of equipment important to safety. The structural integrity of the piping system will be maintained. There is no creation of a possibility for an accident or malfunction of a differont type than any evaluated previously in l the Safety Analysis Report. The pressure in lines 2"-HCD-69 may increase tc reactor pressure if the two, ASME Section III, Class 1 NLSATTC2/SNLICFLR - 32 l A
, = p; _
i' Attcchment to AECM-89/0093 t-Y NPE-69-021
- Page 2 F
isolation valves were to leak. The piping downstream of valve F052A/B has been tvaluated and is acceptable for the postulated conditions. The existing pipe break analysis is not affected by this design. Since all applicable code allowables are met, the probability of an accident resulting from a seismically initiated pipe break is not increased. Therefore,'there is no adverse interaction with any other safety system,.and the piping will perform in its intended manner. No new failure modes are created. . The inctallation of the capping items will not change the limiting
, conditions for operation applicability or surveillance . requirements. An evaluation has been performed to show that the design pressure for this piping may be upgraded to rated reactor pressure and temperature for the piping upstream of the cap.
Therefore, this change will not affect the margin of safety. i l l NLSATTC2/SNLICFLR - 33
r , I ' b ow Attachment to AECM-89/0093 e ;
-SRASN: NPE-89-022 DOC NO: DCP-84-0203-500-R00 &'R01 SYSTEM: C34 !
i DESCRIPTION OF CHANGE: A'O.85mVOC potential exists between the C R SRU signal common and the computer input signal common. It is , necessary to.directly connect the signal common circuits, so that ' feedwater flow readings on the NS$$ computer will agree with the : corresponding voltage reading at the SRU,
. REASON FOR CHANGE: This change was made 50 that feedwater-
- readings on the NS$$ computer will agree with the corresponding '
voltage readings at the SRU. SAFETY EVALUATION: .There is.no increase,in the probability of cccurrence or in the consequences of an accident or malfunction of equipment.important to safety previously evaluated in the Safety Analysis Report. Steam and feedwater flow are controlled by C34 to. assist anticipation of flow error problems. The " reactor water r high level portion" of the feedwater control system employs two-out-of-three trip logic to the RFPT "A" solenoid, RFPT "B" solenoid, and main turbine trip system, Therefore, two identical
, component failures (out of three) within the trip signel are required for an errorieous trip signal or icss of trip signal.
Therefore, no credible accidents are associated with C34 system failures. This change provides improvement of C34 readings at the , computer, and does not affect accident probability, i No accidents are ideretified for the feedwater control system; i although the C34 system is required to maintain reactor water between the high level turbine trip and low level scram setpoints, the level.of control required to achieve this is much less stringent than that required to optimize separator / dryer carryover (of water to steam lines) and carryunder (of steam to water , systems). This optimization must be considered the primary function of the C34 system, and relates mainly to equipment l performance. !! This DCP merely implements direct connection of computer and SRU signal commons within the C34 system ir, order to eliminate a small L. , (0.85mVDC) potential difference between these two common circuits, y Existing equipment is unchanged by this DCP other than the slight I improvement in feedwater control system readings at the NSSS , y computer. l The C34 system is not as important to safety as it is to minimizing carryover and carryunder. If carryover /carryander characteristics are optimized, adequate margin to the high level trip and low level scram setpoints are easily met. The only I safety effect of increased carryover /carryunder is the decrease in core subcooling and concurrent decrease in operating margin to thermal limits caused by too much steam in the water (carryunoer); as above, optimization of carryuader for operational concerns assures that this safety concern is easily met. This DCP provides NLSATTC2/SNLICFLR - 34 9
,- ,4m --
)
k-i Attachment to AECM-89/0093 5 [c?!D9-
- (. < - i <cc NPE-89-022- 3 Page 2 l 1
more accurate NSSS readings only._ Therefore, there is no creation J s of a' possibility for an accident or malfunction of a different i type than any evaluated previously in the Safety Analysis Report-
~4 Technical Specification Basis 3/4.3.8, "A high reactor water s : level, level 8, signal will actuate the feedwater system / main i turbine' trip system.", is unchanged as a result _of_this design change package. Therefore, the margin of safety as det1hed in the ,
basis of' technical specification is not reduced. i F V i 6 Y i NLSATTC2/SNLICFLR - 35
W Attachment to AECM-89/0093
'l f.
SRA$N: 'NPE-89-023 DOC NO: DCP-84-3051-$00-R00 & ROI SYSTEM: C11 DESCRIPTION OF CHANGE: This DCP utilized spare existing cable as a replacement for scheme 1A141011A. REASON FOR CHANGE: Rod 44-37 Channel No.- 1 PIP cable was damaged during CRD removal. This cable was not repairable. Therefore, this DCP replaced it. 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 evaluated in the Safety Analysis Report. There is no change to circuit function or operation. Damaged cable was replaced with the same type of cable. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Also, there is no reduction in the margin of safety as defined in the basis for any
' Technical Specification. :
I NLSATTC2/SNLICFLR - 36
O ' ig y Attcchment to AECM-89/0093 p , ,
' II
- o SRASN:- 'NPE-89-024 DOC NO: -DCP-86-4500-S00-R00 & Rol SYSTEM: M20- !
r DESCRIPTION OF CRANGE: This Design Change Package provides ;
- instructions to repair and modify the Upper Containment Pool removable ladder.' The modification to the ladder consists of i
' making the ladder into two separate sections and shorting the top ,
section of the ladder. These modifications will prevent the ladder from obstructing the movement.of the refueling platform. .; The repairs to the ladder consist of replacing-the damaged lower' '
; portion of the ladder with a new section.
REASON FOR CHANGE: This DCP is generated in response to
-MHCR 0597-85.
SAFETY EVALUATION:~ There is no increase in the probability of t occurrence or in the consequences of an accident or malfunction of ! equipment important to safety previously evaluated in the' Safety
-Analysis Report. The Upper Containment Pool Removable Ladder is ,
not associated with any system or component used in mitigating the consequences of-an accident as analyzed in Chapter 15 of the FSAR. Additionally, the. repairs'and modifications to the'1 adder meet the ! applicable n.aterial~ and construction standards. The repairs and modifications to the ladder meet the original design specifications _for material and construction practices, thereby allowing the. ladder to perform its intended design function. No equipment important to safety is affected by this change. Therefore, there is no creation of a possibility for an accident or malfunction of a different type.than any evaluated previously =l' in the Safety Analysis Report. The Upper Containment Pool Removable Ladder is not addressed by , any Technical Specification nor.does it serve as the basis for any technical specification, therefore the margin of safety as defined is not affected.
.i
+
h NLSATTC2/SNLICFLR - 37
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' Attachment to AECM-89/0093 H
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.SRASN: NPE-89-025 DOC NO: .0CP-82-0636-500-R00 SYSTEM: C41 l I
R DESCRIPTION OF CHANGE: ~This DCP added ladders to the Standby l Liquid Control System Storage (SLC) Tank and foundation. The -! ' ladc'er addition performs no safety function and has no effect on plant operations and safety. REASON FOR CHANGE: The addition of these ladders provide access '! to valves on top of the Standby Liquid Control System Storage Tank. ; SAFETY EVALVATION: There is no increase in the probability of r occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis _ Report. No accidents involving SLC tank ladders or
, . accidents originating in tne SLC system are postulated in the -
UFSAR. Nevertheless, the indder addition will not increase the probability of a failure. Material-and construction standards are , the same,as for an existina ladder. The added weight is smal1~ r when compared with the overall tank weight. - The ladder addition does not degrade or prevent actions described in the " Sequence of Events" tables in UFSAR Chapter 15 since it does not.take part in the sequence of events. Specifically, the ladder addition will not interfere with tF SLC tank function in mit.igating an ATWS as discussed in UFSAP. Section 15.8, 1 The ladder addition is designed as a Category I component with adequate clearance from safety-related components, so that damage to existing components will not occur. The ladder addition has no functional or physical interface with other safety-related portions of the SLC system (except for the tank), or with other systems. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The ladder addition has been designed such that a failure of the ladder or the tank to which it is attached need not be postulated. The ladder addition is a static component attached to the exterior , of the tank and its foundation with no interface with the tank's safety functions. The ladder addition has been analyzed to demonstrate that additional loads on the SLC tank are within code allowables and will not cause deformation or rupture of the SLC tank. The ladder will not fail so as to damage other, adjacent equipment. Neither the SLC tank pressure integrity nor the ladder addition are associated with any specific margin of safety as defined in the Bases section of the Technical Specifications. The ladder addition does not affect process parameters of the SLC system such as flow rate and pressure, the tank's sodium pentaborate solution NLSATTC2/SNLICFLR - 38 l'
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A l q Attachment to AECM-89/0093 j [ b SRASN: NPE-89-026- DOC NO: DCP-84-0011-501-R00 SYSTEM: N43 l- I
! DESCRIPTION OF CHANGE: MNCR 276-85 identified a discrepancy l between the Bechtel design and the vender design for the Primary 1 Water Temperature High alarm (IN43-TAH-L612). This annunciator is j to provide monitoring capability for the main generator stator bar 1 temperatures. The vendor drawings indicated that 6 RTDs imbedded l in stator bar slots should be paralleled to provide the annunciation whereas the Bechtel design utilized I thermocouple downstream of the stator bar primary water outlet header.
.MNCR 276-85 was dispositioned as repair.per a DCP.
DCP 84/0011 Rev. O, provided BOP computer points for the primary water outlet temperature of all 144 main generator stator bars. DCP 84/0011-1 Rev. O installs a circuit from the BOP computer to annunciator IN43-TAH-L612. A computer program is also generated which will allow this annunciator to alarm on the following j conditions: 1) Any stator bar exceeds a predetermined fixed reference temperature, and 2) Any stator bar exceeds a ; predetermined percentage of the average stator bar temperatures. j REASON FOR CHANGE: This design will therefore provide coverage , consistent with the original.bDC design except this design will ; monitor all 144 stator bars instead of six stator slots. -
, ,c 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 evaluated in the Safety
- Malysis Report. This design change will install a circuit from the BOP computer to control room annunciator IN63-TAH-L612. Both physical and electrical separation will be maintained for the new circuit in accordance with Reg. Guide 1.75. Both the annunciator and the primary water system are non-safety related. The performance of annunciator IN43-TAH-L612 is consistent with the intent of the original vendor design. The operation of safety related equipment will not be affected by this design change.
Annunciator IN43-TAH-L612 is not required for mitigation of -
+
consequences of any accident evaluated in the UFSAR. This annunciator was originally designed to monitor the temperature of six stator slots to provide indication of high stator bar temperature. The new design will monitor the outlet water temperature of all 144 stator bars to provide indication of stator bar temperature. The Turbine Trip System is provided with a turbine trip signal for High Primary Water Temperature and for Low Primary Water Stator Circuit Flow in order to furnish generator protection. No system components are expected to operate outside of design limits. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 40
,;.c 4
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Attachment to AECM-89/0093-
. s , :
NPE-89-026 ' Page 2 performance of annunciator IN43-TAH-L612 will be consistent with the original vendor intent, No system component will be expected to operate outside of their design limits. Therefore, the margin : of safety as defined in the basis for any Technical Specification
'ir not reduced. l i
i 4 ! i N h a b Y R l NLSATTC2/SNLICFLR - 41
AttCchment to AECM-89/0093 SRASN:. NPE-89-027 DOC NO: DCP-84-0064-S00-R01 SYSTEMt. M51 DESCRIPTION OF CHANGE: -This design change package (DCP) implements the following changes: (1)1 The division II powered drywell chiller' condenser cooling water control-valves N1P44-TV-F531B and F532B.will'be modified to fail open on the loss of instrument air. The non-divisional drywell chiller condenser cooling water control valves HIP 44-TV-F531A and F532A will remain as fail close-valves. This is because in a LOP, where instrument air 1s lost, plant service water (PSW) normally used as the' condenser cooling medium, is replaced by standby service water (SSW). It is known, from the SSW flow balanco performed during the 1985 Fall Outage, that adequate SSW flow is'available to'only the two division 11 powered. drywell' chillers. Therefore, only these chillers will be available in a LOP. The division II powered drywell cooling fan coil unit discharge dampers N1M51F002, F004, F008, F011, F013, and F017 will be modified per this DCP to fail open on'the loss of instrument air. The division-1 powered dampers will remain as fail close dampers. There is no advantage of failing open division I dampers because none of the drywell chillers or drywell chilled water circulating pumps are powered by division I. As a' result of the above modifications, drywell cooling will be
. available.during a. loss of instrument air, during a LOP, and combined LOP and loss of division I. Drywell cooling will not be available during a combined LOP and loss of' division II.
(2) This DCP will disconnect _the fused circuit micro-switches presently installed in the drywell cooling fan coil unit circuits of N1M51B001A-A, B002A-A, B003A-A, B004A-A, B005A-A, B006A-A, B001B-B, BOU2B-B,-B003B-B, B004B-B, B005B-B, and B006B-B and reinstate continuity of the circuit. All power and contro) circuits passing through penetrations are protected with current limiting Gould MSCP fuses as backup protection. Disconnecting the micro-switches in the above fan coil circuits p will not delete penetration protection or motor protection devices. Thermal protection fuses, magnetic circuit breakers and high amperage penetration protection fuses will still exist in the circuits. This.DCP is safety related because the drywell cooling unit is considered to be associated loads, as defined by IEEE-384. They are supplied from Class 1E motor control centers that are shed on a LOCA signal and are otherwise controlled by the load shedding and sequencing panels (Ref. UFSAR Section 9.4.8.1). l l l-p NLSATTC2/SNLICFLR - 42 I
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'Attcchment to AECM-89/0093 NPE-89-027 Page 2 i
"U (3) .This DCP will also permanently remove the on-line tube cleaning system from the drywell chiller condensers, the plant
; chiller condensers, thy Turbine Building cooling water heat . ,
exchangers, and the component cooling water heat'exchangers. This work was originally planned for DCP 85/0043, Rev..O. However, to avoid a controls interface problem between the implementation of Section 1.3.1.1 and the' deletion of the drywell chiller condenser tube cleaning system, DCP 85/0043, Rev. O has been incorporated
'into this DCP.
Of the heat exchangers equipped with the on-line tube cleaning system, only the drywell chillers are equipped with controls which 3 automatically initiates the cleaning cycle. The automatic-backwash cycle will be eliminated by this DCP since it will no. longer be required. The heat exchangers are presently manually cleaned on an as-needed basis. .This practice shall be maintained to ensure efficient
;n operation of the heat exchangers.
REASON FOR C3ANGE: Due-to the change in scope and design DCP 84/0064, Rev. 1 is being issued to supersede DCP 84/0064, Rev. O in its entirety.
-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 evaluated in the Safety Analysis Report. There'aro no accident evaluations in the UFSAR for'the systems affected by this DCP. .Per FSAR Sections 9.2.2.3, 9.2.7.3, 9.2.8.3, 9.2.9.3, 9.2.11.3, and 9.4.8.3, Systems P42, P71, P44, P43, P72, and M51 respectively have no safety-related functions as defined in UFSAR Section 3.2 other than Auxiliary Building containment or drywell penetration isolation valves where applicable. These valves are not af fected by the hmplementation of DCP 84/0064, Rev. 1.
Other than applicable Auxiliary Building, containment or drywell seismic Category 1 penetrations equipped with redundant isolation valves, the syste.T.s affected by this DCP have no part in mitigating the consequences of an accident. This DCP does not affect any seismic Category 1 penetrations or isolation valves. Therefore, the implementation or performance of this DCP has no direct or indirect impact on any accident previously evaluated in the UFSAR. There is no equipment important to safcty affected by the implementation or performance of this DCP. The only equipment affected by this DCP that is associated with safety related applications is the drywell cooling fan coil unit. These units are considered to be associated loads, as defined by IEEE-384. j They l { NLSATTC2/SHLICFLR - 43 x >
Att:chment t3 AECM-89/0093 NPE-89-027 Page 3 are supplied from Class 1E motor control centers (MCC) that are shed on a LOCA signal and otherwise controlled by the load shedding and sequencing panels (Ref.: UFSAR Section 9.4.8.1). Disconnecting the micro-switches installed in the circuits of these units will not affect the Class 1E MCC and will not delete penotration protection. Thermal protection fuses, magnetic circuit breakers and high amperage penetration protection fuses will still exist in these circuits. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. DCP 84/0064, Rev. 1 changes the fail position of the division II dryw L1 chiller condenser cooling water control valves and the division II drywell cooling unit discharge dampers. These valves and dampers are non-safety related components installed in a non-safety related system. They are provided for flow control purposes, not for abnormal condition isolation devices. This DCP also disconnects the micro-switches from the fan coil units listed in Part 2 of the above description. Disconnecting the micro-switches will not delete penetration protection or motor protection devices. The micro-switches are non-safety realted components installed in a non-safety related system. The tube cleaning systems being deleted by this DCP serve no safety related function. They were originally installed to enhance the performance of the heat exchangers in Systems P42, P43, P44, P71, and P72. Other than Auxiliary Building, containment or drywell isolation devices, these systems have no safety related function. The implementation or performance of DCP 84/0064, Rev. I will not create the possibility of an accident of a different type than any already evaluated in the FSAR. All equipment being modified by DCP 84/0064, Rev. 1 is not related to the safe operation of the plant. Failure of the affected M51, p42, P43, P44, P71, and P72 systems will not compromise any safety related system or component and will not prevent safe reactor shutdown. There are no Technical Specification bases affected by the implementation or performance of DCP 84/0064, Rev. 1. The modifications being performed by this DCP affect only non-safety related components in non-safety related systems. The primary design functions of these systems will not change due to the implementation of this DCP. Therefore, there is no direct or indirect impact on any other margins of safety as defined in the Bases for any Technical Specification. NLShTTC2/SNLICFLR - 44
1 i Attachment to AECM-89/0093 ? SRASN: NPE-89-028 DOC NO: DCP-84-0235-$00-R00 SYSTEM: C41 : l L F DESCRIPTION OF CHANGE: DCP 84/0235 adds a 8" T-bolt hinged ! closure to the top of the Standby Liquid Control (SLC) storage !
- . tank. Currently access to the inside of the tank to draw a sample ;
is obtained by unbolting a 30" manway cover, located on top of the : H tank. A hand rail that will provide fall protection for personnel ; working in the vicinity of the hinged cover is also provided by this DCP. .The hand rail does not attach to the SLC storage tank. The SLC storage tank (IC41A001 containing the boron sclution) is a ; safety related, Seismic Category 1, ASME Section III, Class.2 ; component. The tank design requirements and its classification as : discussed in Table 3.2-1 of the UFSAR have not been changed. The : modification meets all the requirements of the Design l Specification including the seismic requirements. The tank ; capacity and other auxiliary features including instrumentation , .(as discussed in UFSAR Sections 7.4.1.2 and 9.3.5.2) are not affected by this change. The hand rail is riesigned Seismic , Category II/I to preclude it from becoming a fall hazard on other . safety related components including the SLC tank. The change does : not alter existing operational or design parameters of the Standby Liquid Control System. No other safety related systems or components are affected by this change. No new failure modes are t created by this ch a ,e. The changes meet all applicable design requirements of the affected system. REASON FOR CHANGE: This change allows ease of drawing a sample or the addition of chemicals to the tank. 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 evaluated in the Safety Analysis Report. This design change adds a hinged cover to the and adds a hand rail for personnel top of the SLC protection. Thestorage tank;s operation and function will not be SLC System altered by the change. The modification to the tank meets all the requirements of the Design Specification and the ASME Code applicable to the tank. Component Classification of the tank as discussed in Table 3.2-1 of the UFSAR is not altered by this modification. The hand rail is designed Seismic Category II/I. The design change will not compromise any other safety related system or component. The change does not affect any actions taken to mitigate an accident as analyzed in the UFSAR. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NI,SATTC2/SNLICFLR - 45 1
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, This.DCP does.not change the limiting conditions of operation, :
3 applicability or surveillance requirements as defined in the bases 1 l for the Technical Specifications. Therefore, the margin of safety .; lis not reduced, j 0 ) 9 Y E. g' , IN l- . i. I i l i i i i NLSATTC2/SNLICFLR - 46 4
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'SRASN:-^NPE-89-029' DOC NO: MNCR-88-0129- SYSTEM: E12 3.
g'J . DESCRIPTION OF CHANGE: /n engineering' evaluation was performed tol g ,- xdemonstrate.the acceptability of the addition.of fire hose Ladapters to.several small bore pipe test connections.1 The iadapters were installed to-provide'an alternate injection path,for
,2 the reactor vessel. makeup from the P64 system during station
< blackout conditions; A--total of eight adapters _wereLinstalled at gh locetions in E12 E21,' E22 E51 and P11 systems 'The adapters were' installed at the test connections associated'with valves E12-F057C & F056C;' valves E12 F061 & F062;c valves E21 F013 & F014; .
valves E22-F021 & F022; valve P11 F124; valves E12 F058A & F059A;- Y Lyalves E51 F213 & F214; and valves E12 F058B & F0598.
'. REASON >FOR CHANGE:' This MNCR was written to document the i y installation of-the adapters and to allow the updating of.the :
o, , P& ids.and' piping isometrics since their installation constitutes.a ; D des _ign change. i
)
, SAFETY EVALVATION: There is no increase in the probability of. ;
occurrence or in the consequences of an accident or malfunction of- , equipment important to safety previously evaluated in the Safety '
.;g -Analysis Report. The additior, of the eight hose adapters in the.
E12, E21, E22, E51 and P11 systems has been shown to meet ANSI j
~B31.1 code requirements and Seismic II/I criteria. These r connections will provide an alternate injection path for RPV i L
-makeup.from the P64 system during station blackout conditions. ;
!* The operation or function of the affected systems as analyzed in l+ the FSAR are not affected by the-addition of the adapters. i The changes do not modify any equipment used in mitigating the
. consequences of an accident as analyzed in the FSAR, nor do they affect any actions taken to mitigate an accident as analyzed in
~
i the FSAR. The~ adapters will not affect the operation of the E12, E E22, E21, E51 and P11 systems as analyzed in the FSAR. Therefore, i L there is no creation of a possibility for an accident or K malfunction of a different type than any evaluated previously in the Safety Analysis Report. This addition of the adapters to the systems E12, E21, E22, E51 ; ,1 and P11 will not change the function of these systems as defined L by the bases of the Technical Specifications. Therefore, the margin of safety as defined in the basis for any technical u specification will not be reduced. + !(
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I NLSATTC2/SNLICFLR - 47 b. u -
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- SRASN: NPEi B9-030 000 NO: DCP-83-0202-500-R00 ~ SYSTEM:- P53 y ,% ,
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J & DESCRIPTION OF CHANGE: The addition of instrument' air Dewpoint O ' LAnalyzers that provide an analog. readout of tne instrument air
# dewpoint on th'e discharge from both airedryers.
V
% . REASON FOR CHANGE: This DCP provides for Dewpoint Analyzers and tthese instruments will further enhance.the reliability of~the=
instrument air system by allowing continuous monitoring of air quality. _ hAFETY: EVALUATION:There is no increase in the probability of J occurrence or in the consequences of,an accident'.or malfunction of equipment'important to safety previously evaluated in the Safety i
@ Analysis Report.. The installation of the dewpoint analyzers which
@g . perform a passive monitoring-function only will not increase the-probability'ofta major.line break or equipment-failure as- .,
described in'15.2.10-of the FSAR. The installation ~of thel
- dewpoint'-analyzer will:not have any. effect on the ability of plant-equipment tosachieve a. fail safe condition as described in 15.2.10.3.3.' 'The'insta11ation of'the dewpoint'unalyzers will not affect the ability of backup. air supplies as. described in 15.2.-10.3.3 from;being available to equipment important to safety.
, The! installation.of the dewpoint analyzers will-not affect the operationLof alternate air supplies for'importance to safety 0 described in 15.2.10,3.3 because the' dewpoint = analyzers are !
l 4 located in the water treatment building where.there.are no ! v category II/I considerations. The new instruments provide an l indicating. function'only-and do not affect system-operation. . Therefore, there:is no creation of a possibility for an accident ! c 'or malfunction of a different type than any evaluated previously ! .in the-Safety Analysis-Report. There is no reduction in the l L, margin of safety as defined in the basis for any Technical : Specification. This-system is not addressed by the GGNS Technical
.Spe:1fications.
4 l r i L! 4 NLSATTC2/SNLICFLR - 48
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i "A - a SRASN: NPE-89-031 DOC NO: - DCP-83-0034-S01-R00' SYSTEM: .N19- > v
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-DESCRIPTION OF CHANGE: HNCR 00270-84 identified damage to piping ,
in pipe support NIN19G008R03 and concrete-breakout at north base t Lplate caused by transient condition in the system. .This DCP is
+ "- : adding a reinforcing pad over the west side'of pipe, two contour
.r- pads'on.the east side of pipe and re-torquing the concrete- .
'T Q expansion anchors. >
REASON'FOR CHANGE: These changes ensure that the pipe support'. will function correctly. 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 evaluated in the Safety-Analysis Report. System safety analysis has shown thataf'ilure of-1- -
the- condensate system will not comproM:e any safety related system nor prevent.a safe shutdown of the plant. Therefore, the condensate system' serves no safety function. The pipe support-modified by this DCP meets all ANSI B31.1 code requirements and. will' function in its . intended manner. No new modes of failure will occur. Therefore, there is no creation of a possibility for-an accident or. malfunction of a different type-than any evaluated previously in the Safety Analysis Report. Also, there is no' reduction in the margin of safety as defined in
;the basis for any Technical Specification. Since the system is not. utilized in. establishing a margin of safety, the margin of-safety is not affected.
1: L L 1 I' h l NLSATTC2/SNLICFLR - 49
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-SRASN: NPE-89-032l DOC NO: DCP-84-0236-500-R00 . SYSTEM: R25
- m j i
k" ' DESCRIPTION OF CHANGE: .Atpresent-annunciatdrsR25-XA-L607 (34.5 KV BUS 12 INC:FDR 552-1102 TRIP). and R25-XA-L608 (34.5 KV i', fBREAKER 552-2102-TRIP) alarm in the control room on' Panel-H LSH13-P807:when their. respective breaker trips. Both of these-L+ breakers feed Bus 13R. -They have an interlock to prevent being tied to bus 13R simultaneously. A* s To resolve:this problem, inputs for the two annunciators (R25-SA-L607 andiR25-XA-L608) will be wired in series in Panel i ITB001 to form"a singleLannunciator (R25-XA-L607) that will alarm
~
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,, -only when both breakers (552-1102 & 552-2)02) trip. This will allow one breaker to be tripped, while the:other is closed (Tied , 'e to Bus-13R) without'having.an. audible alarm in.the control room.
Thus an alarm,will only occur when a true off-normal condition
' exists. Each breaker will continue to have separate indication on
~
s the~B0P computer. ~ .The annunciator window for breaker 552-1102 will be removed and the new "34;5'KV BUS 13R INCM FDR BREAKERS TRIP" annunciator put in its place on SH13-PS07-2A-A4 and numbered , i
-as R25-SA-L607." The existing annunciator window for breaker 552-2102 on .H13-P807-4A-A6.will be removed and a blank window a installed in its place- .
.' < REASON FOR. CHANGE: This change resolves this problem which occurs ,
during normalcoperations'since one of the two breaker annunciators is always alarmed. This: alarm is a nuisance and a violation ~of
' good human factors as defined-in NUREG-0700.
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 evaluated in the Safety
. Analysis Report. The cha'nges affect indication of two non-Q l ' breakers on the-H13-P807. Benchboard. -Each breaker will continue to have Separate indication on the BOP computer. The changes are A , designed.to decrease the probability of human error while operating plant equipment. This is done by reducing the number of L J " nuisance" alarms which the operators must continuously monitor.
The seismic and combustible loads for the panel are not affected. These modifications.will decrease the-probability of humcn error while operating plant equipment, due to continuous alarms in the , main control room. The change will not affect any seismic and environmental as well as safe shutdown and fire protection L criteria for these panels. The breakers will continue to have separate indication on the 80P computer. The DCP changes do not affect the operation of the R25 system equipment. The combining p f of the alarms will serve to give the operator warning only when a true "off-normal" condition exists. The DCP changes do not affect the decision of consequences of any malfunction of equipment in s the UFSAR. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated
' previously in the Safety Analysis Report.
H l L NLSATTC2/SNLICFLR - 50 L
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(p ' Attachment >to AECM-89/0093 '. -i n A' a L.' ' NPE-89-032~ .r F Page.2: L
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The changes are designed to increase thi operator' margin ~of ( P.' je! safety,-:by reducing the probability of. human error while operating.
@ plant equipment,- The Seismic' qualification,' safe shutdown, fire. i protection and environmentul' criteria are not affected by this f". ' .
- design change.
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, $ SRASNi NPE-891033 DOC NO:~ MCP-88-1038-S00-R01 SYSTEM: -R20
~ DESCRIPTION OF CHANGE: . Minor Change Package (MCP) 88/1038 L provides for the installation of cooling fans inside load center ..
K' enclosure LC14 bel. The fans will effectively, increase ~the rating-i' of the transformer. The modification wili be confined to-the
~ interior of.the non-safety related enclosure. Power will be Ex supplied internally from the non class.IE load: center bus.
Isolation;and physical separation in accordance with IEEE 384^as - [' 'provided by Regulatory Guide 1.75, Physical' Independence of ;
-Electrical Systems, ensures that malfunctions to.the cooling fans
' or fan. circuits will not propagate to any equipment-important to i safety'as previously evaluated in the'FSAR.' Malfunction of the
, load ~centerl d ue
~ to. fan failure will not create the possibility of .i
,a an accident'of-a-different type than already evaluated, as~a
. failure of.any' loads connected to LC14BE1 would be' enveloped by a i -loss of offsite power'as previously evaluated. The loss of ADHR ,
3 pump motorsLto be added by DCP 88/0008 and DCP 88/0008-7 due to
~6~
load center failure was evaluated in safety evaluations CFR-88/0008R00 and CFR88/00087R00. The consequences were found to be acceptable. The installation of the fans and increased rating of the
- transformer will!be reflected'in the one line diagram for the unit l 3
- (E-1081). This drawing is not included in the FSAR. An editorial change which adds. detail by reference to drawing E-1081 will be made to the. main one line diagram (E-0001) which is UFSAR Figure 08.1-001. The change will not affect the technical content l
of Figure 08;1-001. l REASON FOR CHANGE: This change was made to provide forced air ventilation of the load center transformer, i .t D , 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 evaluated in the Safety l Analysis Report. The modification has no physical impact to components, structures, or systems described in the FSAR. An w editorial change will be made to a FSAR figure to add detail by reference to another drawing not included in the FSAR. The modification has no affect on the operation or function of plant systems described in the FSAR. Further, the changes are limited to the interior of a non-safety related enclosure. Separation and isolation per. Reg. Guide 1.75 ensures that malfunctions will not propagate to safety related systems or equipment. The consequences of failure of the load center are enveloped by accid 9nts or occurrences already evaluated. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety
- Analysis Report.
The modification will have no impact on systems, components, or functions that could alter any technical specification safety margins. Therefore, the margin of safety as defined in the basis for Technical Specifications will not be reduced. NLSATTC2/SNLICFLR - 52
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' Attachment to AECM-89/0093
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. < SRASN: NPE-89-034 DOC NO: OCP-88-4501-S00-R00 SYSTEM:- Fil i
i EDESCRIPTION OF CHANGE: A modified HFTS Containment Penetration Closure Hatch.has been designed for the GGNS. The HFTS transfers fuel bundles and control rods between the containment and the auxiliary building and therefore is considered a containment penetration which requires a closure assembly.to provide the< necessary containment isolation boundary. 'This closure provides a
- b. means of making a positive seal.across the HFTS tube'in the region.
$w , where the HFTS passes through.the containment structure' at the-transfer pool. The closure: assembly is welded to the containment Et t .
p ' - penetration sleeve and these form a part of containment boundary
-during reactor operation. ' A design change is necessary due to-the , recent experienced difficulty in opening:and closing the hatch.
The'. system description of the HFTS is contained in Section :i 9.1;4.2.3.11 and; illustrated in Figure 9.1-15-of the GGNS UFSAR. ; A new design concept with a more positive latching mechanism has been developed by GE. Detailed design information is provided~in GE specification 22A4614 Rev. 'l " Containment Penetration Closure"
. and modification Drawing 103E1537 Rev.1 of the same title.- ~ This
, change only involves the closure assembly. No other mechanism of E
7 the HFTS is affected. 1 L REASON FOR CHANGE: This change is necessary to ensure reliable !
- opening and closing of the closure. hatch in order-to provide-the necessary containment isolation boundary.
SAFETY EVALUATION: There is no increase in the probability of i occurrence or in' the consequences of an accident or malfunction of l equipment important to safety previously evaluated in the' Safety Analysis Report. The design change does not bypass or cause any B ' b; _ bypass of system design features that will cause an entry into an ; L accident condition as described under the " Identification of - Causes" section of the UFSAR'as the hatch closure or opening is j not a cause of.any UFSAR accident condition. i' The design change does not degrade overall system reliability as i a)' there is no change in power source, i b) there is no change in instrumentation, c) there is no effect on capability of an auxiliary system to support another system, k d) it does not cause any system or component to operate outside design, Technical Specification or testing limits, e) it does not decrease any system integrity as piping and cable support are not changed and original load requirements are met.
'NLSATTC2/SNLICFLR - 53 {
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@ Attachment to AECM-89/0093'
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. , NPE-89-034 Page 2- l lThisdesignchangedoesnotaffectordegradethesystemor.
~
component performance in mitigating the consequences of an
'E . accident analyzed in. Chapter 15 of the UFSAR. The Technical Specification leak rate test will ensure containment isolation ;
will.b~e assured in the event of the accident.. The design change does_ not degrade or prevent actions described in t;he " Sequence of Events" tables of the analyzed accident since containment isolation will be assured:with this change, m The-design change calls for meeting all original material and a construction requirements (i.e., ASME Section III Class MC) including: ra) aJstress analysis meeting normal, upset and faulted load ) (e.g., seismic) conditions,
.b)- separation criteria are not directly applicable since the L
hatch operates independently and'does not require redundancy,
<c)' environmental qualification criteria will still be met because the design is in accordance with the original-
- material and construction requirements and they will operate
# 'in the same environment.
The design change will'not degrade equipment reliability'as: - e 0 a) ~a ll' normal, upset, faulted loads are analyzed meeting the original design criteria,
~
b)_ the component protection features and component' reliability. are maintained and even improved with the new design, c). no support system performance downgrade is necessary in order to maintain reliability' operation of the new hatch. The design change' allows the hatch to open and close as designed without' any anticipated malfunction as a) there will be no valves or pump involved in the design that might cause blockage, b) there is no instrumentation involved that might give faulty indications. The hatch will operate within the design specification as a) a stress analysis has been performed to ensure all pressure and mechanical loads are met, b) since the hatch is manually activated, reliable power sources are not crucial therefore can be reasonably assured, NLSATTC2/SNLICFLR - 54 4 + - - < e , o
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.ap Attachment to AECM-89/0093
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L, l NPE-89-034 r 1 Page 3-. l 3 c) waterIhammerandvibrationarejudgedtobeminimalin'the. ; operating environments of:the closure hatch. ; The possibility for an' accident of.a different type than any '
'already evaluated in the UFSAR is not created because:
. 1)- The: design' change does not involve a new component that was not considered in the UFSAR.
t 2), This change involves the containment penetration component , which is assured in the' Technical Specification to provide containment isolation to mitigate the accidents analyzed in Chapter 15 of the UFSAR.
, - 3) The Technical Specification leak rate testing will assure containment isolation; therefore, no accident of a different type can be created.
The poss'ibility of a malfunction of equipment important to safety different than previously evaluated in the UFSAR is not created because:
- 1) The designichange cannot cause any-other malfunction of. .
equipment as the hatch operates independently preventing any potential transient induced into control circuits for alarms, annunciators'or equipment initiations.
- 2) The design change also cannot cause a malfunction that would cause a-failure of equipment as they operate independently
-from other equipment that'are important to safety.
The margin to safety as defined in the basis for any Technical Specification is not reduced because:
- 1) The design change does not directly affect the operation of 4 the HFTS in transferring fuel bundles and control rods between the containment and the auxiliary building. It only affects the opening and closing of the hatch.. Therefore, none of the key parameters such as power, flow, chemistry, pressure capacity, setpoint or level are affected.
- 2) The Technical Specification requires containment integrity by insuring containment isolation capability. Containment penetration leak rate surveillance testing is to be periodically performed to meet this requirement. The basis for the Technical Specification requirements is to ensure accident analysis licensing limits are met. Since the objective of this design change is to ensure the leak rate test criteria can be met thereby maintaining containment isolation capability, margin to safety is properly maintained.
NLSATTC2/SNLICFLR - 55 i
D{ f 'n- [ [ AttcchmenttoAECM-89/0093-g q 0:. y *
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*SRASN: .NPE-89-0354- ' DOC NO: DCP-84-3000-S00-R00. SYSTEM: P53 h
e , DESCRIPTION'0F CHANGE: This DCP modifies-the wiring of the p;; ' Instrument Air Compressor Control. System. Power failure causes-
; 'the Instrument Air Compressor to trip. The compressor cannot presently be restarted from the Control Room.
{ t REASON FOR CHANGE: .This change was made to allow the' Instrument
/[l<
h. Air Compressor to be restarted from the Control 1 Room. SAFETY: EVALUATION: There.is no increase-in the probability of occurrence or in the consequences of an accident or malfunction of-
- f. / equipment important-to safety previously evaluated in the Safety "1 Analysis Report. The instrument. air has no safety related function. -The failure of this system will not compromise any.
safety related system or component and will not prevent safe
- reactor shutdown. Therefore, there is no' creation of a possibility for:an accident or malfunction of a different type than any evaluated previously,in the Safety Analysis-Report.
Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification. s
-(
i l-l 1 l 1 lj -
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t i 2_., i
y~ ys g , N , LAttachment to AECM-89/0093' k ,' wm 7 . v ' SRASN:1 NPE-89-036L . DOC NO: iDCP-84-4077-500-R00 SYSTEM: -N64 >
, - DESCRIPTION OF CHANGE: The'offgas system process and-instrumentation ~ diagram M-10928, Revision 15 (Coordinates B ,
Lthrough D-7)' depicts the correct installation of valve NIN64F051B
- as' the "Adsorber Train- B Inlet Valve"'and NIN64F051D as- the '
m Adsorber Train B D0128 Bypass Valve". M-10928 also depicts line D
- 1;1/2"-EBD-102, "Adsorber Nitrogen Supply Line", as being 4
connected downstream of N1N64F0518. System piping isometric M-1352D, Revision 13.however, depicts the actual-installation of NIN64F051B as "Adsorber Train B D012B Bypass Valve" and depicts
- N1N64F0510 as Adsorber Train B Inlet Valve".--As a result..of the
; incorrect location of NIN64F0518, the adsorber train B nitrogen
! supply line, 1 1/2"-EBD-102,.is also incorrectly located. These
< nonconformances are documented in Material Nonconformance Report (MNCR)'772-84.
The design objectives of DCP 84/4077 are as follows: l(1) Retag valves N1N64F0518, NIN64F0510, and associated instrumentation (except affected valve handswitches) to reflect the correct installation as depicted on M-1092B, l y Revision 15, (2) -Connect existing handswitch NIN64M022B to the retagged valve NIN64F051B (previously NIN64F0510), 1
.(3) Connect existing handswitch NIN64M0220 to the retagged valve N1N64F051D (previously NIN64F0518), and
[ (4) Connect the adsorber train B nitrogen supply line 1 1/2"-EBD-102 to downstream of the retagged valve N1N64F051B. 1 FSAR Section'3.2 classifies this system'and all its components as "Other" meaning that loss of system function would not affect safe i shutdown of_the plant. This system is Non-Q, Non-Safety Related, I Non-Seismic and NRC Quality Group D. REASON FOR CHANGE: This evaluation ensured that diagram M-1092B and M-1352D are accurate. SAFETY EVALUATION: There is no increase in the probability of 1 occurrence or in the consequences of an accident or malfunction of , equipment important to safety previously evaluated in the Safety 4 Analysis Report. The postulated worse case failures (Off Gas system leak and gross system failure) analyzed in FSAR l Section 15.7.1 envelope the occurrence of postulated accidents due to the change. The affected components are not safety related and do not perform any function required for performance of a safety function by safety related components, equipment or systems. The affected NLSATTC2/SNLICFLR - 57
pg -
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Attachment:
to'AECM-89/0093: r~ viQ
~
1$r !3I i L NPE-89-036-Page 2-i I- components >do not perform.any functions which mitigate the affects. 1
'of.any accident. analyzed in the FSAR. Therefore.--there is no creation of a possibility for an accident or malfunction of a different typeithan any evaluated previously in the Safety Analysis Report.
n;; j: .The change'does not change.the' limiting conditions for operation, ' F applicability,Jactions, or. surveillance requirements as-defined in the' basis for. Technical Specifications 3/4.11.2 and 6.15.- ' t l l 4 e
,1.
I i 4 NLSATTC2/SNLICFLR - 58 s
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i {i LAttechment.to AECM-89/CD93 fn ;+ g 23;, SRASN: 1NPE-89-037. ' DOC NO: MCP-89'-1055-S00-R00 SYSTEM: 'N12 :f u 1 e; p-it . . . .. . I;F LDESCRIPTION 0F CHANGE: . The header' supplying steam to,the RCIC: 7,, . Turbine'willibe separated from:the normal Auxiliary' Steam System. m~ . .. REASON FOR CHANGE: . This] change allowed an independent' steam !
; , isource"to supply steam to the RCIC turbine.-
il SAFETyLEVALUATION:- There is no increase'in.the. probability of, .
~
p . occurrence lor in-the consequences of.an accident or-malfunction of- i;
. equipment important.to' safety previously evaluated in the Safety
~
?
4 y, JAnelysis: Report. The piping design; meets ANSI B31.1 code requirements. The piping is support,ed for dead weight and thermal' 7~^ loads-only, sinceLit is installed in an area'oftthe Turbine
+ Building which contains no safety related equipment. :The Auxiliary. Steam System serves no. safety function.- Systems~
' analysis has shown that. failure of the N12 System will'not compromisefany safety related systems or prevent-reactor. shutdown.
l ' The operation or function of the N12 ~ system, as analyzed in' the FSAR', is-not affected by the modification of-this MCP. The (" .
', , separation of the header supplying steam to the RCIC turbine from the Auxiliary Steam source will not affect the normal function of Auxiliary Steam System. i W 1 I .Therefore,ithere-is no creation ofca' possibility for an accident.
^
or malfunction of a different type than any evaluated previously n . in the Safety Analysis Report. The N12 System is not addressed-in : [~' the. Technical Specifications,'and-therefore, has not been' utilized-in computing a' margin'of safety. s s
"U .i' NLSATTC2/SNLICFLR - 59
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n., _-
/ '
[' ,' Attachment;to AECM-89/0093J i n.,n ; N l i::' M DCP-84-0020-500-R00-
~
SRASN: NPE-89-038 DOC NO: SYSTEM: -N64 v p s
, l DESCRIPTION OF CHANGE: This change provided for the addition of'a l
' . nitrogen (N2)' purge supply ~ for-the. low temperature off gas, system charcoal:adsorbers for fire suppression: purposes. The~ nitrogen i
supply shall provide.the' cooling medium flow to facilitate- i lowering: charcoal bed: temperatures. , REASON ~FOR CHANGE: .This change will reduce system outage l time.
,' ~
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 evaluated >in the Safety- -
W Analysis Report.. The N2 purge supply will not affect the normal
. operation of theLnon-safety-related N64 system and will not- ,
increase the. probability of occurrence of an accident previously evaluated'in the FSAR. The N2 purge. supply;is being added for fire' suppression purposes. .The piping and pipe supports supplied
.by.this DCP-will meet all.applicalle design requirements and will.
7, .
. function in.their intended manner. No important.to safety 'i components wi.11'be.affected by the addition of a N2 purge supply.
Therefore, there is no creation of-a possibility for.an accident i or malfunction of a different type than any evaluated previously in the Safety Analysis Report. q
> The margin of safety as defined in the basis for any Technical )
. Specification-wi.11 not be reduced since the N2 purge supply is i being added for-fire suppression. purposes. l I
a
,.s NLSATTC2/SNLICFLR - 60
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k Attachment-to'AECM-89/0093 ai
]F >
E ;SRASN: NPE-89-039 000.N0: DCP-84-4032-500-R00_ SYSTEM: N64 E - DESCRIPTION OF CHANGE: This DCP replaced Hiller Position Switches" jF. with NAMCO Switches on 20 valves-in the N64 Off-Gas System.
~
As- q' stated in FSAR Chapter 3.2, the classification of the affected - f
. component is:- Per Table 3.2-1 XXIII the safety class is "Other".
- REASON'FOR CHANGE: The.present switches are'not reliable.-
;,194 ' SAFETY; EVALUATION: There is no increase in the probability of p SY ' occurrence or in the-consequences of an: accident or malfunction'of '
< equipment'important to safety previously evaluated in the Safety Analysis Report.' The new switches were. tested and'they performed' better than _the switches beingi used now, . so~ the probability for an accident:1s not-increased. This change is a' direct replacement of the existing switches with more reliable switches performing the s same function for which they were designed. No new failure modes are introduced. Therefore,:there is no creation.of a possibility
~
2 , - for an accident or malfunction of a different type.than any. evaluated previously in the Safety Analysis Report. Also, there
~
< is no reduction in:the margin of safety as defined in-the basis for any Technical Specification.
=
, e i (-
[. [/. ? 1. NLSATTC2/SNLICFLR - 61
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- Attachment'to AECM-89/0093) g k u-; ,
k ISRASNL: NPE-89-040 DOC-NO: DCP-85-0066-S00&S01-R00 SYSTEM:. N19 Dx n4 DESCRIPTION OF CHANGE:. The condensate booster.. pumps, heater drain " . pumps and reactor feed pumps are equipped with-low flow. trips in-
- order to prevent pump damage' caused by low flow conditions. DCR 85/066' requested that'the single deviation' alarm cards be replaced with! dual alarm cards and that computer alarm points be acided.
DCP-85/066-1 will' replace the single alarm cards Bailey e' , 744110AAAA1 with' dual alarm cards Bailey 744210AAAA2.and add the
. requested computer alarm points-1 4' REASON FOR CHANGE: These computer al~ arm points will increase plant reliability-by alerting the operator of possible trip conditions before an actual _ trip occurs.
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 evaluated in the Safety Analys.is Report. This change increased the differential setpoint !
- at which the respective pump is shut down. .The installation of
- I the dual differential alarm cards and the addition of the' computer l
- alarm points will increase plant reliability by alerting the '
operator of possible trip conditions prior to an actual trip. ii This wi.11 therefore decrease the possibility of a loss of 4
- feedwater accident as described by 15.2.7 in the FSAR. Revising the setpoint for the condensate booster pump, heater drain' pump, 1 and feedwater pump switches and installing a dual setpoint-switch
- j for the condensate pump affects the associated pump only and does ;
not affect any other equipment previously evaluated in the FSAR. J The added alarm function will-serve only as a warning of an ]
- impending trip of the associated pump and will not serve as a-trip 1 0 function or affect any -safety related trip functions. The changes .l being made by this DCP will increase system operational a
reliability of the affected pumps only and will not have any other affect on other plant components. Therefore, there is no' creation
- of a possibility for an accident or malfunction of a different type than'any-evaluated previously in.the Safety Analysis Report.
The affected switches and associated setpoints are not addressed I by the GGNS Technical Specifications. The pumps controlled by the alarm cards and the computer alarm points are not addressed in the GGNS. Technical Specifications. NLSATTC2/SNLICFLR - 62 i W 4 y ..--a e v -.- ,. -
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,, Attachment to AECM-89/0093 l
DCP-83-4528-S00-R01 SRASN:.-NPE-89-0411 DOC-N0i SYSTEM: 291 k * ' DESCRIPTION OF CHANGE: . Cable tray- 1BDTDQ11 is a 12" B0P cable T: m - tray. that'is located below Elevation .111'-0" of the Control Building, Area 125B. The cable tray was identified by MNCR 0899-83 as' requiring an additional support to withstand seismic loadings.
- This cable tray does not pose any II/I hazards -in the present-configuration.- DCP 83/4526 is being initiated per the MNCR
, , disposition to install an additional cable tray support a
, '(A2B-S109-1-1350).
t REASON FOR CHANGE:. This change provides consistency with other cable' tray designs and preclude any, future Seismic II/I hazards. SAFETY EVALVATION: There is no increase in the probability of. - occurrence or. in the consequences 'of an accident or malfunction of. equipment importart to safety previously evaluated in the'. Safety
' Analysis Report. The design for the new cable tray support meets 1
'all FSAR requirements and support design criteria. Installation of the new support will ensure structural integrity of the BOP cable tray. The B0P :able tray and its supports are not required to mitigate the consequences of an accident previously evaluated in'the'FSAR. The seismic design of the cable tray system will prevent any hazardous effects on other plant equipment, structures, and components. The 80P cable tray support configuration is designed such that it will remain structurally adequate during a seismic event and not create any new hazards.
~Therefore, there is no creation of a possibility for an accident 1: or malfunction of a different type than any evaluated previously in the Safety ~ Analysis Report.
l '... The B0P cable tray and its supports are not addressed by any
-Technical Specification nor serve as the basis for any Technical Specification.
NLSATTC2/SNLICFLR - 63
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' 3 4, 3
ff " '
, .Attdchment to AECM-89/0093 a-y ,
h",' ' SRASNil : NPE-89-042 DOC NO: ~DCP-88-0010-S00-R00 SYSTEM: Fil~ i E -! h*- b ;
' DESCRIPTION OF. CHANGE: The purpose ^of DCP 88/010 is to enhance Jthe' operational ~r0.iability of-the Refueling and Fuel Handling
$i ; Platforms. The modifications include 1) replacement of the main ,
<; power / interlock cables on-both platforms with a more flexible b[4 cable, 2) replacement of the existing hydrostatic load cells on the main hoists'and' auxiliary hoists of both. platforms.with electronic load cells, 3) installation of a trolley-mounted
? <
- Auxiliary. hoist on the Fuel Handling Platform (FHP) similar to the !
& ~ existing trolley-niounted hoist on the Refueling Platform -(RP), ' 4) ; modification of the Upender Zone'(UPZ) interlocks _to allowt ! F RP. trolley-jog "left" and'FHP-trolley-jog "right" motion-(retreat motion) when the interlock is encountered upon approach, and 5) ' %- addition.of forced ventilation cooling o'n the RP Main Holst Power N Center:end RP Bridge / Trolley Power center' panels. The platforms
'are-safety related (Class 2) and Seismic Category I from a structural standpoint (UFSAR 9.1.4.1).
REkSON.FOR CHANGE: This change was made in order to support
. refueling outage schedule, ,
2 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 evaluated in the Safety 4 Analysis Report. The only existing UFSAR accident evaluations- ,
with probability: of.-occurrence applicable to the RP or FHP changes
'are the' fuel drop accidents described in UFSAR 15.7.4 (Auxiliary Building. accident) and 15.7.6 (Containment-Accident). In the
;, postulated events, a spent fuel assembly is dropped onto stored. 3 1 spent fuel due to failure of the fuel assembly lifting mechanism.-
None of the changes in DCP 88/010 adversely affect the load
. bearing capability of the RP or FHP' main fuel hoists, and auxiliary hoists are not permitted to handle spent fuel (Technical Specification 3/4.9.6.1.and 3/4.9.6.3). The' changes do have indirect. affect on the main hoists as follows: the power / interlock cable'provides power to the main hoist, the RP Power Center panels provide fuel hoist control, and the main hoist
-load. cells provide load indications and inputs for the load int'erlocks. Replacement of the power / interlock cable is intended to decrease failure frequency an a reliability enhancement. ' Power cable failure causes loss of power to the main hoist which activates the noist brakes preventing a load drop. The forced ventilction cooling of the RP Power Center panels will resolve current overheating problems allowing hoist control per the original design intent. The electronic load cells will perform the same function as the existing hydrostatic load cells. The new load cells will maintain their calibration better than the existing load cells which " leak down".
NLSATTC2/SNLICFLR - 64
yp , 7, . 6 - I
-Attcchment to AECM-89/0093 h.
k L. NPE-89-042~
\ Page 2
;The only evaluated UFSAR accident evaluation applicable to the RP
~
or FHP are the UFSAR 15.7.4 and 15.7.6 fuel drop accidents. In the postulated even's, spent fuel was-dropped onto stored spent l fuel from a drop height of six feet (max.) causing damage to 101 , t fuel rods. . None of'the changes in DCP 88/010 will allow fuel to be raised higher than the existing up-limit switch trip points L . which would result in a drop height of less than five feet. Also, none-of the affected RP or FHP equipment is required to mitigate the consequences described in UFSAR 15.7.4 or 15.7.6s
-DCP 88/010 does not adversely affect the fuel grapple, grapple-E mast, mast support, or main hoist winch. None of the RP or FHP
!? components which are modified can create a fuel drop accident, A. neither will their malfunction result in malfunction of load 4 bearing components that can cause a fuel drop. In fact, the ( '
-DCP 88/010 changes are intented to enhance the equipment reliability by reducing the frequency of equipment failures experienced in RF01 and RF02 (power cable failure, loss of load cell calibration, panel overheating trips, etc.).
Affected RP and FHP equipment shall provide the same operational function as the existing equipment per the original design intent. The modifications will provide increased operational reliability per the original design criteria, and seismic adequacy of the RP and FHP structures are maintained. Also, the RP and FHP p , modifications will not affect any cther safety related equipment in any operational ~ mode. Changes to the UPZ interlock will. allow
- retreat motion of the RP and FHP fuel grapple when the interlock b.
is encountered upon initial entry'into the transfer canal "upender zone". However, these changes will not inhibit the safety feature provided by the interlock. The purpose of the interlock is to prevent possible fuel damage caused as a result of inserting or removing a. spent fuel bundle in the HFTS with a less-than-vertical upender. Such an accident would be different than any previously evaluated in the UFSAR, but its possibility is not created since the interlock will still function as intended in the immediate vicinity of the upender. Therefore, the possibility of a new accident type is not created. < Since none of the RP or FHP modifications adversely affect the
, load bearing capability of the main fuel hoists or the seismic integrity of the overall platform structures, malfunction of the affected RP or FHP equipment will not result in the malfunction of other plant equipment that is important to safety. The changes in DCP 88/010 are intended to decrease the possibility of malfunction of the subject equipment to increase operational reliability.
Since all equipment affected will comply with the original design criteria and/or perform the same function as the existing equipment, the possibility of a new nalfunction of equipment important to safety is not created. NLSATTC2/SNLICFLR - 65
v 3:; , n ; 40 4 Attachment to AECM-89/0093 t <c . 7:, , . Wo -
, NPE-89-042-
..Page- 3 L s- - . .
3
-Changes _tocthe. subject RP and FHP equipment-are applicable to only.
Technical Specification:3/4.9.6 and 3/4.9.12. The margin of
, safety as. defined in 3/4.9.6.is not_ reduced since 4.9.6.3.1d and e and 4~.9.6'.3.2 will-be changed to_ include the applicable .
Y , surveillancefrequirements for the'new trolley mounted ~ Auxiliary hoist on the FHP and clarify surveillance requirements applicable to the Monorai1~ Auxiliary _ hoist only. This change will ensure , v thatialisthe DCP 88/010 modifications will comply with--the bases: ' C' , given for-3/4~.9.6. Also,'the. basis for 3/4.9.12 is not affected since:none of the DCP changes 1 affect personnel access control to - Room 1A525 and safe operation of the. HFTS _is maintained (UPZ . interlocks provide the safety function for.which they were 4
. intended).
P
}
( w p C l: l-i. NLSATTC2/SNLICFLR - 66
WY
,y '
l At'tachment to AECM-89/0093-L SRASN: NPE-89-043 D00.NO: MCP-89-1052-500-R00 SYSTEM: C34 P
~
DESCRIPTION OF CHANGE: Reactor feedwater lines A&B (24-0B0-25) N. have. venturi type flow elements with two' sets of differential pressure. sensing taps. The differential pressure sensing taps measure the same values inside the-flow element but'on opposite sides. . On each flow element only one set of the taps is utilized from which the parallel transmitters for reactor feedwater and the1
- circ. water makeup system receive'their input differential .
I pressures. The. reactor feedwater lines are not required to be ASME Class -1, 2 or, 3, nor are .they Seismic Category I components. Downstream of the' instrument root valves for the reactor feedwater; system transmitter and the circ.-water makeup system, the tubing leading to the cire, water makeup system flow transmitter is to be v cut and capped. This change will leave the reactor feedwater system flow transmitter receiving a signal from the original set of taps. , The tubing downstream of a set of root valves on the spare' set of ; taps will be' cut and the cire, water makeup flow transmitter will j be connected. The cire, water makeup flow transmitter will now !
- receive a differential pressure from taps on the opposite side of !
' the flow elements but will still be measuring differential 1
_ pressure inside-the flow element. There 'is no electrical work involved in this change. REASON FOR CHANGE: After the completion of this tubing change the i R reactor feedwater flow transmitter and the circ. water makeup flow
- transmitters will receive differential pressure through completely L different connections to~the flow element. These values can be compared and can be trended for divergence or convergence-which ,
may be indicative of a flow element problem. ! u SAFETY EVALUATION: There is no increase in the probability of j occurrence or in the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the Safety l Analysis Report. The flow transmitters will be measuring the same values as'before on the same respective flow element but using ,
'different +.aps. These instruments serve no safety function or support equipment important to safety. There are no design requirements associated with the flow element, the differential !
~
L, pressure connections (piping, root valves, tubing, or instrumentation) affected by this change. The tubing affected by this change is non-ASME and non-seismic. No electrical work is to be performed. The instrumentation associated with this tubing modif(cation is not required to function to mitigate the consequences of any analyzed accident or transient. The flow element, differential pressure connections and connected instrumentatin are not required to be designed to ASME Section III or Seismic Category I requirements. The instrumentation is not NLSATTC2/SNLICFLR - 67 t - y e
~ ~
$$?, ' ' I' ;c > Attachmentito AECM-89/0093 f h NPE-89-043. ( ,
-page 2 as ,
,e required to. function to mitigate the consequences:of a malfunction
"~
[g,4 3-of equipment important'to safety. Failure of any of these E!' ' components will not affect the ability of any structure, system or. ;
- component to' maintain reactor pressure boundary, shutdown the
~
reactor, or mitigate the consequences of an accident. Therefore, there is no' creation of a' possibility;for an accident or
- malfunction of'a different type than any evaluat'ed previously in
~
1the Safety Analysis Report.- This instrument tubing will be' modified in accordance with , original-design requirements applicable to the existing-installation.. The readings to be obtained-will be from the same-
-flow element at different but;1dentical points. Consequently, the=
accuracy of'the process merasurement and the confidence level of
- the instrumentation performing this measurement will'not be
'affected by this modification. Therefore, the margin of; safety as defined 71n the basis for any Technical Specification has not been.
reduced. t S
.I._
t ; i NLSATTC2/SNLICFLR - 68
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1 Att chment to-AECM-89/0093
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, =SRASN:" NPE-89-044~ ; DOC'NO: .--DCP-82-5073-S00-R00 ? SYSTEM: B21' ,
s -!. 1:l' ' - i:
#1 '
l DESCRIPTION.OF. CHANGE: . This change added piping taps for(pressure : , =l
< scs. j . transducer ~B21G02201PP and B21G02501PP.- *!'
= REASON ~FOR CHANGE: ; The addition of the piping taps allowed;for e'
e 'theLimplementation~of theLoriginal plant design. 1 E$lU? (( SAFETY EVALUATION:: - There is.no increase in the probability of
, occurrence.or in the consequences of an accident or malfunction of SL '
equipment?important to safety previously' evaluated in the-Safety.
?? Analysis' Report? The. original 1 design intent of adding pipe taps, ;
h 'n - for; pressure transducers B21G02201PP and B21G02501PP.is. maintained ,
- by this DCP.; !The. addition of the
~
"T"_ taps-is non-safety related.
1 The-addition'of'"T" taps per original plant des 3gn does not affect
~.
'O 'the safe:o.peration'nor the safe shutdown'of.GGNS. This-DCP'does :
s not change.the FSAR. .Therefore, there is no creation of a '
-possibility for an accident 1or malfunction of a different'. type than any~ evaluated previously in the-Safety' Analysis l Report.
Alson there is no reduction in the margin of safety as defined in'
?! 'the basis for any Technical. Specification..-Following first cycle- l testingi the "T". taps were plugged and seal welded' 7 ,
9 m-t l.1; ; 1 i 1 F
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'NLSATTC2/SNLICFLR - 69 i
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iSRASN: NPE-89-045 I ' DOC NO: DCP-86-0037-S00-R00 SYSTEM: P7l'
+f g ,
~
'DESCRIPTIONzoF CHANGE: .This DCP will provide the capability,for. ,
P monitoringicorrosion rates and the offectiveness of chemical. g 7; treatment-in the Turbine Building Cooling Water (P43), Component ,
' Cooling Water-(P42),-Drywell Chilled WaterL(P72) and Plant Chilled w> :Waterf(P71)Lsystems by use of the Metal' Specimen (coupon) Test Method.c-The-design will add:a loop to each system'which will'
~
9r P* containicorrosion coupon holders, a' flow indicator and. the . necessary valves for operation of the loop. a' ( j REASON FOR CHANGE: No systematic method of' monitoring corrosion rates'in plant water systems exists at GGNS. A system is required
~
f' '
-in order to determine the life of components and to improve the.
~
b"{ ' ; efficiency of _ the components within_ these systems.: t , 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 evaluated in the Safety
. _ Analysis; Report. The modifications provide a process loop for_the installation of corrosion. coupons. The piping and pipe supports designs meet' ANSI B31.'l code requirements and are qualified as
- seismic II/I, where. applicable. The installation is non-safety "related and will'not affect the safety related portions of these .:
systems. Of'the affected systems, only that-portion of the system 4 g which penetrates the containment is safety related. Failure of (' the-affected systems wil1~not compromise any. safety related. system or-component and will not prevent reactor shutdown. .The operation or function'of the'affected systems as. analyzed in the FSAR is not
, - affected by-the' addition of the corrosion monitoring loop. The 4<
changes do not modify any equipment used in mitigating the consequences of an accident as analyzed in the FSAR, nor do they f L ' affect'any actions taken to mitigate an accident as analyzed in I the FSAR. The modifications will not affect the operation of the P71, P72,.P43 and P42 systems-as analyzed in the FSAR. The flow being diverted through-the corrosion monitoring-loop will not affect the normal function of the systems due to the minimal flow g rates for this. loop. Therefore, there is no creation of a 0 possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The addition of the corrosion monitoring loop to the systems P71, i P72, P42 and P43 will not change the function or operation.as defined by the bases of the Technical Specifications therefore the margin of safety is not reduced. h f NLSATTC2/SNLICFLR - 70 w ,
,c # :- m - r- ,, -
m hg ' ' [gil' f.qtlk y ns of.u + m - AttichmentitoLAECM-89/0093) fl f.p-i '* , k[. I M
~
O} . .>'
'l . -SRASNi'?NPE-89-046"'
1 DOC NO: CN-A30-032 ? SYSTEM: --- A3 0 (ip}" 1g fi[' ,
'. DESCRIPTION OF CHANGE: This change added a Nu. bottle rack'at'-
iL as '
-elevation 133 Lit lat'the turbine building.-
pp: - ,.
' ~ . t 1 GREASON FOR CHANGE: cThis: change secures the No bottles.'in place'.?
llch/ , , g ,, p%' .. , , w . SAFETY EVALUATION: There!is noLincrease in the probability of g -:4-
, 7
-occurrence.or in the consequencesLof;an accident'or malfunction'of-
~
~
$1
![, s ' equipment important-to safety,previously: evaluated in the Safety 1' Analysis' Report. .The bottlerack secures.ethe N,. bottles:in place'.
P > -The rack increases the. security ofcthe N, bottles.-. Therefore,. b there,isino creation of a possibility"for-an accident or malfunction of a different type'than-any evaluated previously in-
,2 6 ~ <
(the. Safety Analysis Report. Also, there;is no reduction in the
. margin of safety'asidefined-in the-basis for any Technical'
~
F e4- -Specification'. g i: s r iI v 4% , s. 3 s 4 h i i .- 5 r. I l e
'l .
- 'j-jt I.- 1 s 3_
n- .l h l -- !y NLSATTC2/SNLICFLR - 71 l'^ x t t .h
V' ' m ..'
- g. < M y j .
, Attachment to AECM-89/00931
;g . >
, :SR'ASNk.NPE-89-047' DOC NO:- 0CP-87-0053-S00-R00L SYSTEM: R63 s
p , _ .< . DESCRIPTION OF CHANGE: The purpose of this design' change is to . d . physically 1 remove the chlorine detectors from service and 1
- electrically disconnect the detectors.from the Control Room E. 1 i isolation. system. .
MV . REASON-FOR CHANGE: .In order to limit exposure to unnecessary isolations'and since the chlorine detectors-are no longer included- 1 in .the' GGNS- Technical: Specifications, the. chlorine detection
! system'should be removed:from service and electrically.
disconnected from the Control Room isolation system. p !! SAFETY; EVALUATION:; There:is noiincrease.in the prcbability of. I' , occurrence or in' the consequences of an accident _or-malfunction of" equipment important:to safety- previously evaluated in the Safety _i Analysis Report.
. The change does not. increase the probability of an'onsite or an offsite chlorine accident. SERI has adequately demonstrated that an onsite a'ccidental chlorine release without automatic _ isolation
- of the-Control Room would result in chlorine concentrations well
/ below the toxicity limit suggested by Reg. Guide 1.78. In ! addition, SERI has shown that the probability of an.offsite chlorine accident is within the SRP Section 2.2.3 acceptance criteria.
The consequences of an onsite chlorine accident are discussed in Section 2.2.3.1.3 of the FSAR. The results of the analysis :t described in that section indicate that SERI is in compliance with. the' requirements of Reg. Guide ~1.95 which requires liquified chlorine in quantities greater than'20 pounds to be stored at
.least 100 meters away from the Control Room'or its fresh air inlets. The results also~show that with no credit taken for ControlRoom_ isolation,tgechlorineconcentratjoninsidethe Control Room of 24.6 mg/m is below the 45 mg/m allowed by the Reg. ~ Guide 1.78. The chlorine detection system provides mitigative measur_es applicable only to chlorine release accidents.
No malfunction of equipment important to safety previously evaluated in the FSAR is predicated on a failure of equipment affected by this design change. All applicable design reviews were performed and have concluded that modification of equipment by-this DCP does not create a seismic II/I hazard or any other
# type of hazard to safety-related components. Removal of the chlori.ne detection system requires no changes in the transportation or use of chlorine onsite or offsite.
Therefore there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 72
"]
- w= '
(h);h ; p. Attachment to AECM-89/00931 Li: -- L i
- NPE-89-047
. Page 2 e '
a
~
~;; . .
With the deletion of.the chlorine detection system, the chlorine-
> t concentration inside the Control" Room will~still; remain below the
! allowed limits established by Reg. Guide 1.78. .Also, the
. probability of an offsite accident-is extremely low, design-
, requirements are not required tol mitigate ~the event as-specified-
"=
in-SRP 2.2.3. Therefore, the change does not reduce the margin of.
' *- 4 tv safety-as defined;in the bases for any, Technical Specification. . + , , :'-e.
r i l i i l 1 > it . L ! a r4 NLSATTC2/SNLICFLR - 73
gyye . m y- g " At ttchment ' to AECM-0?/,0093 1
-Al
. SRASN ' NPE-09-048
' DOC NO: DCP-84-0231-S00-R00 SYSTEM: R61 J
< 4
-DESCRIPTION OF CHANGE: Evacuation warning lights are being installed in high background noise areas.of the plant. The
' 91 rotating. lights ~are activated'byfa receiver / relay connected to the page'line.of.the P.A. System. The existing multitone generator will be. replaced.w!th a similar tone. generator that han an R.F.
ac Transmitter. incorporated into it.' When the evacuation-alarm is
- initiated the normal' sound is heard and.a 50 KHZ R.F. Signal is impressed on the+"Page" pair of wires. This R.F. Signal enters-the receiver / relays and enorgizes the' rotating warning lights. !
- i. '
Power for the receiver / relays and rotating' lights is. connected to the P.A. 120 volt. .To compensate for the additional load on the V.A. 120 volticircuits,'two (2) additional 120 volt circuits are. being added. One of these 120 volt circuits is being added at the > Radwaste Building and.the other is being added to the Aux. Bldg. l t at elevation 139'.
-The amplifier for the Turbine Bldg. roof speakers is now powered-
-from MCC 13B22-11. This amplifier is being reconnected to the uninterruptable power supply to provide yard coverage in the event 1
g/~ of an accident.-
. REASON FOR CHANGE: This change is made in response to LCTS Identification Number 2-A.
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 evaluated in the Safety
' Analysis Report.
The plant P.A. System is not safety related. Evacuation warning lights are being added to the plant P.A. System. All supports for
, ' raceway and equipment in Category I buildings have been seismically supported or reviewed for II/I compliance.
Additions are being made to the P.A. System which is not required to mitigate the consequences of an accident. Failure of the P.A. System will not affect the sequence of events in the accident analysis.
. Separation requirements are being met on all circuit additions.
Seismic concerns in Category I buildings have been addressed. Additional circuits are being added to compensate for additional
'3" loads on the P.A. System.
The warning lights being added by this DCP are designed to operate with the P.A. equipment by the use of an R.F. Receiver / Relay combination. The P.A. System has no safety function. There is no creation of a possibility for an accident or malfunctions of a different type than any evaluated previously in the Safety Analysis Report.
- NLSATTC2/SNLICFLR - 74
Attachment to AECM-89/0093 ! 4- ,: NPE-89-048 : Page 2 , The P.A. System is not analyzed'in Chapter 15 since it is not , safety related. No new failure mode is being created, j The R.F. Receivers and Warning Lights are connected to the P.A. System. Failure of either would only affect the P.A. System. Additional loads on the UPS bus has been reviewed and will not degrade the panel. The P.A. System is not addressed in the Technical Specifications. Therefore, there is no affect on the margin of safety. i [ ; r N t I i r I. NLSATTC2/SNLICFLR - 75
3; g* , Attcchment t3 AECM-89/0093 SRASN:- HPE-B9-049 DOC NO: DCp-85-0163-S00-R00 SYSTEM: N41 ;
.i DESCRIPTION OF CHANGE: DCp 85/0163 installs reflash capability to existing alarm circuits in the GAC cabinet. The alarm circuits ;
affected by this DCP are the inputs to control room annunciator ,
' lights IN41-XA-L614 (Gen. Seal Oil Trouble) and 1N41-XA-L611 '
(Primary Water. Trouble). At present the inputs associated with each of;the listed annunciators are paralleled, forming the common k trouble alarm inputs to the 2 control room annunciators described
.above.
i
.The modification consists of installing 5 Ronan reflash modules and 31 auxil ary relays in the GAC cabinet, removing the common
- inpute for each annunciator and rewiring them to the newly installed reflash modules. These modules ~will be wired to provide the 2 common alarm outputs to the existing control room annunciators. The indication in the control room will be
. ., identical to the presskt indication upon a trouble alarm, except il'! 'that additional alarm inputs received prior to the clearing of the F' '
existing alatm will reinitiate the alarm from the t.cknowledge ;
"g state back to an alarm state, thereby informing control room ?
p" < personnel of anoth9r associated alarm-inpet. This action will , 4 continue until all alarm inputs have been cleared. r , The Generator Monitoring System is not a safety system nor is it tied'directly to any safety system. The purpose of the Generator i Monitoring System is to permit the monitoring of generator support j
. systems for detection and correction of trouble within the '
systems. No system set points.are being changed and no new common inputs are being added to either annunciator. The alarm system will funct$on and report the identical trouble inputs as it did prior to the implementation of this package, with the new - capability of detection of addition alarms common to the annuncaitor prior to the local alarm being cleared. '
' REASON FOR CHANGE: By adding reflash capability to each of these
. . circuits' control room personnel will be able to detect subsequent alarms within the same system prior to the clearing of each alarm at the local station.
4 /' SAFETY EVALUATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of , equipmont important to safety previously evaluated in the Safety f- Analysis Report. p The Generator Monitoring System monitors generator support systems , i for detection and correction of trouble within the generator auxiliary systems. The alarm system will function and report the identical trouble inputs as it did prior to the implementation of this package with the new capability of detection of other alarms common to the annunciator prior to the local alarm being cleared. A failure of this annunciator system will not impact any equipment NLShTTC2/SNLICFLR ~16 y a
Attachment to AECM-89/0093
-NPE-89-049 i Page 2 ,
safety function nor will it introduce other equipment malfunctions that could instigate an accident. The effected system has no
- i. control function or capability and is not directly tied to any
- safety or control-system. The Generator Monitoring system is ,
;. non-safety related and has no direct ties to any existing safety I systems. Therefore a failure internal to the Generator Monitoring .
System will not increase the probability of malfunction of any ' safety system. The effected system is used as an aid in the
- c. evaluation and handling of Generator support system events. i Therefore, there is no creation of a possibility for an accident r
or malfunction of a different type than any evaluated previously in the Safety Analysis Report. ; The Generator Monitoring System is not addressed in the GGNS r lechnical-Specification and will not reduce the margin of safety " as defined in the basis for any Technical Specification. l-l' s NLSATTC2/SNLICFLR - 77
gmg -. - - , , Attachment to AECM-89/0093 SRASN: NFE-89-050 000 NO: DCPL 86-0016-SOC-:t00 SiSTEM: N11 DESCRIPTION OF CHANGE: Turbine extraction steam line restriction orifice condensate drain bypass valves IN11-F005A, B and i IN36-F008A, B will be provided with a means to be closed by 1: overriding the bleeder trfp valves (BTVs) " fully open" interlock ! when all other existing interlock requirements have been met. l ,p REASON FOR CHANGE: This change will provide operations with a ! more efficient means to operate the plant by closing the [ restriction orifice bypass valves when the plant is running for extended periods of time with a steam flow through the BTVs which , will open but not fully open them. - 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 evaluated in the Safety Analysis Report. The restriction orifice bypass valves affected by this change provide a larger pipe diameter to drain condensate or steam from , lines 24-GBD-l'and 16-GBD-32. A turbine trip will cause-the BTVs in these lines to vent the air pressure from the closing spring j which will put pressure on the check valve disc to close. Should steam flow reverse, the check valve disc will close (BTV valve fully closed) which opens the closing circuit to the affected restriction orifice bypass valves thus causing them to open. The open bypass valves will then depressurize lines 24-GBD-1 and s 16-GBD-32 even if the override condition exists. The additional capacity in draining these lines by the restriction orifice bypass valves is not evaluated in the FSAR.
- The most credible abnormal condition associated with the restriction orifice is a turbine trip. During a turbine trip, the ;
restriction orifice bypass valves may be opened by the closing of the line isolation valves IN11-F003A, B, IN36-F011A, B or the closing of the associated BTV valve. The failure of restriction orifice bypass valves IN11-F005A, B and IN36-F008A, B to open during a turbine trip is not analyzed in the FSAR. This change increases the components in the closing circuits for ; valve IN11-F005A, B and 1N36-F008A, B. The components allow the
" fully open" interlock requirement of the associated BTV to be bypassed while still maintaining an interlock requirement that the associated BTV be "not closed". If one of the closing circuits fails to close the associated restriction orifice bypass valve, the result would be less efficient mode of operation.
NLSATTC2/SNLICFLR - 78
gg , s.. Attachment to AECW-89/0093 p'- p I '- 9 NPE-89-050 Page 2 F When closed, there are four ways to open the restriction orifice bypass valves: [L a. Handswitch N11-M602A, B/N36-M617A, B
! - b. "HIGH-HIGH" signal from the associated heater / reheater drain tank
- i. c. An inlet gate valve to the MSR/Feedwater Heater #6 "NOT FULLY OPEN" signal
- d. A BTV N11-F003A, B/N36-F012A, B " FULLY CLOSED" signal.
The affected. restriction orifice bypass valves do not support any V< safety related equipment nor does the chenge in the closing ' circuits of thess valves increase the probability of a malfunction of equipment important to safety previously evaluated. The addition of a bypass relay and additional contacts in the i closing circuit of restriction orifice bypass valves IN11-F005A, B and IN36-F008A, B only affects the opening and closing of these non-safety related valves. The pressure boundaries of the valves , will not be affected. No new accident event precursors are being introduced. Therefore, there is no creation of a possibility for an accident , or malfunction of a different type than any evaluated previously in the Safety Analysis Report. This change affects the closing circuit for valves IN11-F005A, B and 1N36-F008A, B which are not used for the basis of any Technical Specification. All existing design and installation requirements affecting margins of safety are not changeo by this DCP. NLSATTC2/SNLICFLR - 79
, , v y.
x, 4 4 , i ., y .Att:chment to AECM-89/0093 SRASN: NPE-89-051. DOC NO: DCP-86-0055eS00-R00 SYSTEM: P72 [ D ' DESCRIPTION OF CHANGE: DCP B6/0055, Rev.'O, is being issued to
.p . provide the documentation and instructions necessary to install a
. sample point in.the Drywell Chilled Water System which is
-specifically designated and properly equipped for taking water
. samples. Sample point N1P72SXN048 will be installed in the
'drywell chilled ~ water discharge header and routed to a sample sink .
located'nearbyLon Elevation 119'-0" of the Auxiliary Building.
- q. The sample sink drain:(1"-JBD-1023) will be routed to.the Chemical ,
'Radwaste System (CHRW). 'j 4 RNASON.FOR CHANGE: This DCP provides the documentation and !
instruction'necessary to install a sample point in the Orywell i Chilled Water Bystem. [ g'
~,
F 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 evaluated in the safety ,
, Analysis Report.
-The piping / tubing installed by this DCP is non-safety related,. ,
2non-seismic and will in no way affect the moderate energy pipe i break analysis in-'the.FSAR. The sample sink stand has been -i designed to withstand the applicable seismic loads to preclude any II/I hazards. The.Drywell Chilled Water System is a non-saftey -t related system whose failure will in no way compromise any' safety related systems or components or prevent'a safe reactor shutdown. ; Nor is the Drywell Chilled Water System' required to operate to ; help mitigate the consequences of an accident. The piping / tubing modifications specified in the DCP meet all applicable code requirements, and will in no way impact the pipo break analysis in the FSAR. Therefore, there is no creation of a possibility for an accident l
-f or malfunction of a different type than any evaluated previously ; ,, in the Safety Analysis Report.
b Implementation _of DCP 86/0055 will not reduce the margins of safety as defined in the basis for any technical specification. The.Drywell Chilled Water System is not addressed by the GGNS
. Technical Specifications nor does it impact the margin of safety of any systems addressed in the Technical Specifications. The drywell chilled water system does not directly interface with any Technical Specification systems. l
.NLSATTC2/SNLICFLR - 80 .
Attachment to AECM-89/0093
, SRASN: NPE-89-052 DOC NO: DCP-86-0116-500-R00 SYSTEM: 944 .
DESCRIPTION OF CHANGE: The piping to the drywell chiller ' condensers does not currently have-individual isolation valves l L that would allow maintenance to individual condensers. Current , i isolation valves are in a common header to two condensers skids i N1P72B001A and B002A or N1P72B001B and B0028. These valves ,
} isolates both skids when the valves are shut. During power !
operations two complete skids cannet be isolated be:ause two compressorsfromtheskidarerequiredtomaintaindrywgil i temperature below Technical Specification ilmits of 135 F. Due to ! L; the recent concerns with systems of SSW/PSW interface these ! L condensert should be periodically inspected for coo' ling coil i degradation; l OCP 86/0116 provides the necessary piping and support design to install iso'lation valves to allow individual maintenance to the condensers. REASON FOR CHANGE: This change allows individual maintenance to ; the drywell chiller condensers-. ; 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 evaluated in the Safety Analysis Report. The (P44) Plant Service Water System operation and function will not change. The portion of the system affected by this DCP has no safety related function as discussed in the FSAR Section 3.2. Failure of the system will not compromise any safety-related system or component and will not prevent safe reactor shutdown. The piping and pipe supports supplied by this DCP meet all applicable design-requirements and will function in their intended manner. 1' L As per UFSAR Section 9.2.8.3, the failure of the affected portions I of this system will not compromise any safety-related system or l- component. l Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The portion of (P44) Plant Service Water System affected by this DCP has no safety function. The portion of (P44) Plant Service Water System affected by this DCP is not addressed in the Technical Specification and therefore have not been utilized in computing the margin of safety. NLSATTC2/SNLICFLR - 81
e I Attachment to AECM-89/0093'
, 1 i
SRASN: NPE-89-053 DOC NO: DCP-86-0125-$00-R00 SYSTEM: ESI ! L ' DESCRIPTION OF CHANGE: Presently the annunciation on control I l panels 1H13-P845, 1H13-P842, 2H13-P811 and 1H13-P808 do not have : annunciator acknowledge operator push button handswitches. The I r . acknowledge input is jumpered to the silence input and upon L" operator action to silence the alarm it is also acknowledged. ,
. This DCP provides acknowledge push button handswitches to the above listed panels. This provides additional operator .
acknowledgement as to which alarm has just come in should there be.
, uncleared alarms already present. Therefore, af ter implementation ;
of this design, operations will be capable of acknowledging. j incoming alarms independent of. silencing the alarm. Reference l HED's 169, 740 and 610 in support of the above described design i
+
change. Presently the RCIC turbine status lights on the remote shutdown ! panel 1H22-P150 stay dimly lit due te induced voltages on the circuits. The present light configuration per status light is four neon bulbs. Neons are capable of illuminating with the 7 presence of induced voltages due to the low currents neons l require, once the neon bulb turns on. l This design will replace the four neon bulbs per status light with ' two incandescent bulbs per status ligt.t. The incandescent bulbs will not be susceptible-to induced voltages as'the neon bulbs because they require a much larger current to illuminate. In addition to changing the bulbs, a dropping resistor board will be added to drop the lamp voltage to 28V. The resistor board will be ; seismically supported. The status light modules are designed to operate with neon bulbs or incandescent bulbs rated at 28V, therefore, the status light lenses will not be effected by this ' design. REASON FOR CHANGE: This change will allow the operator to acknowledge incoming alarms independent of silencing the alarm on , control panels 1H13-P845, 1H13-P842, 2H13-p811 and 1H13-P808. The neon bulbs on the panel for RCIC turbine status were replaced with incandescent bulbs because the neon bulbs were remaining dimly lit due to an induced voltage. 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 evaluated in the Safety Analysis Report. This design change reduces the potential of operator error by elimination of a glow problem caused by induced voltages on the RCIC status lights located on the remote shutdown panel. This provides more definitive indication of the status of the RCIC turbine. After this design change is implemented, the RCIC NLSATTC2/SNLICFLR - 82 9
-w-*
Attichment to AECM-89/0093
.NPE-89-053
, page 2 r
iturbine status lights will not be continuously illuminated and will instead provide lighted indication should a problem arise in
, a monitored area.
presently on panels 1H13-pB45, 1H13-pB42, 2H13-pB11 and 1H13-pB08. ; y' .an operator action to silence an alarm.also acknowledges the , alarm. This design' change provides a separate acknowledge
- pushbatton'handswitch.in addition to the existing silenco pushbutton handswitch. This provides additional operator i acknowledgement as to which alarm has just come in should there be *
+
uncleared alarms already present. . n; , The~ circuits affected by this design change perform no automatic plant operating control function. The circuits will perform their
-original decign function after implementation of this package.
The RCIC turbine statua light circuits will be composed of a soismically-supported dropping resistor board in series with 2BV
~
incandeccent bulbs as described above in the description. The circuits are now composed of neon bulbs with dropping resistors mounted internally to the bulb.. [ The acknowledge handswitches being added will separate the , t silence / acknowledge function of the annunciation systems as i described'in the design summary. Therefore, instead of having one , handswitch performing both the silence and acknowledge function, '!
.there will ba-two independent handswitches, i
^ Failure of these circuits after implementation of the design will i remain bounded by the consequences of their failing prior to the ;
implementation of the design and thus will not increast the consequences of an accident previously evaluated.in the FSAR. The resistor board mounted in 1H22-?150 per this design change I will be seismically supported. The RCIC turbine status light assemblies will not be replaced; the bulbs within the light , assambly will be replaced. Therefore the probability of a-malfunction of equipment important to safety, caused by these components falling during a SSE, important to safety will not be
'" increased. Also, only internal jumpers (Division 1) are being added in a totally Division 1 panel so there is no Reg. Guide 1.75 concerns.
The acknowledge handswitches are being added to non-class IE . panels in non-class 1E circuits and do not require II/I seismic mounting. Only internal jumpers (non-divisional) are being added in a totally non-class 1E panel so there is no Reg. Guide 1.75 concerns. Therefore, no increase in the probability of a i malfunction of equipment important to safety previously evaluated in the FSAR has been created. I NLSATTC2/SNLICFLR - 83 a u.s
f 0 Attachment to AFCM-89/0093 , t r hPE-89-053 Page 3 The RCIC' turbine status lights provide indication for trouble ! conditions occurring on the RCIC turbine. They provide no control l functions (i.e., they provide no permissive / interlock). This , design will not change the function of the RCIC turbine status , lights but will provide a more definitive indication as to the availability of the RCIC' turbine, j t L ' Presently the annunciation on panels 1H13-P845, 1H13-P842, i
~ 2H13-P811 and 1H13-P808 have and utilize an acknowledge function.
.The acknowledge function presently is jumpered to the silence handswitch such that when an operator silences an incoming alarm ,
'.* it is also acknowledged. simultaneously. This design will provide .
a separate handswitch such that operations will have the ability
/ to silence and acknowledge alarms on these panels independently of each other. Therefore, since this design is not changing the ;
n function of the silence and acknowledge alarm circuits and all four panels which will have an acknowledge handswitch added are non-safety.related (i.e., no II/I concerns), this design will not increase the consequences of a malfunction of equipment important : to safety previously evaluated in the FSAR. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Failure of the RCIC turbine status light circuit will be limited to the loss.of Division 1 (i.e., 1H22-P150) as the single worst ! failure and cold shutdown is possible by using the remaining ! operable Division 11 remote snutdown panel per UFSAR Section 7.4.144.5.3. The four acknowledge handswitches will be added to four non-safety related panels per the above description. The single worst failure could potentially be the loss of the non-divisional power feeding the control circuits in these panels and this is limited to.a loss of offsite power event which is bounded by previously analyzed accidents in the UFSAR. Therefore, this design change will not create the possibility of an accident i of a different type.than any already evaluated in the FSAR. Per UFSAR Sectinn 7.4.1.4.5.3 "In the event of a loss of one ESF division of power (i.e., one remote shutdown panel) as the single failure, cold shutdown is possible by using the controls and monitoring. instrumentation provided on the remaining operable remote shutdown panel". Since this design affects only the RCIC turbine status lights on 1H22-P150 and does not affect 1H22-P151 this part of the design will not create the possibility of a malfunction of equipment important to safety different than previously evaluated in the FSAR. NLSATTC2/SNLICFLR - 84
,c
Attachment to AECM-89/0093 i. n NPE-89-053 Page 4 i' Per Section 3/4.3.5 of GGNS Technical Specification for bases of ,
"RCIC System Actuation Instrumentation" the actuation Instrumentation is provided to initiate actions to assure adequate ;
h.- ' core cooling in the event of reactor isolation from its primary ' I heat sink and the loss of feedwater without providing actuation of , any of the ECCS equipment. Also per Section 3/4.3.7.4 of GGNS L Technical Specification for bases of " Remote Shutdown System i Instrumentation and Controls" the operability of the remote I shutdewn system ensures that sufficient capability is available to permit' shutdown and maintenance of hot shutdown of the unit from [ locations outside the control room. ; I This design'has no impact on the contrel/ actuation of the RCIC ; l- Sys+,em from the remote shutdown panel. Per the above rieseription ; l the resistor board mounted in 1H22-P150 will not have' any adverse ' effect on the controls / instrumentation that are a part of 'i 1H22-P150. The.GGNS Technical Specification does not address the annunciation ; system. The addition of the four acknowledge handswitches per the t above description will not reduu the margin of safety as defined its the bases for ar<y Tat.hr.lcal Specification, f
)
r t NLSATTC2/SNLICFLR - 85
o htt:chment to AECM-83/0093-o SRASN: NPE-89-054 DOC NO: DCP-86-0126-S00-R00 SYSTEMt D17
' DESCRIPTION OF CHANGE: DCP 86/0126, Rev. 0 implements several ;
design changes / modifications resulting from the GGNS-1 Detailed Control Room Raview-(DCRDR). The review was in response to NUREG-0737-Supp.: 1, which deals with emergency response ; capabilities. _ The design changes contained in this DCP are a ~> direct result of Human' Engineering Discrepancies (HEDs).identifled l
'during'the DCRDR. This DCP includes the required information to !
implement design changes resulting from HED-54, HED-238, HED-337, i HED-827 and HED-1088. [ (MED-54)- . Presently there are no SRV position indicating lamps on the. remote [ shutdown panels. Per HED-54 the addition of SRV position [
, indication to the remote shutdown panels is desirable. This ;
design will provide one red indicating lamp per SRV on both the ; Division I and the Division II remote shutdown panels. The lamp a will be installed in parallel with the solenoid coil of the SRV to ; provide positive-indication that an open permissive has been
~
initiated., This design is consistent with the SRV position .; indication in the' control room for the Division II solenoid coils. ; Reactor vessel pressure indicators located on the remote shutdown a panels will still provide indication that the SRV's have opened. + i (UED-238) j The process radiation monitors located on panel 1H13-P604 are !* multi-scale meters. One scale-is blue and one is green with a lighted pushbutton selector switch on the meter face to select > either of these two scales. This pushbutton selector switch is i split faced with a blue lens on one side and a green lens on the other. At present the blue side is very lightly colored and difficult to identify as blue when lit. This change installs blue , filter caps on the two lights that illuminate the blue side of the i switch when the blue scale on the meter is selected. j (HED-337)
. Presently the condensate booster pump discharge pressure .!
indicator's (lN19PIR610) scale reads from 0-600 PSIG. The [ associated pressure transmitter 1N19PTN062 is calibrated from ! 0-600 PSIG. Indicators are most accurate when read between 25% to 75% of their range. Normal designed operating pressures are as high as 579 PSIG which is well above 75% of the indicator range. Per HED-337, indicator 1N19PIR610 "is pegged offscale during ' normal ops". This design change will change the scale on
-indicator 1N19PIR610 to 0-1000 PSIG. The associated transmitter ;
will be recalibrated to 0-1000 PSIG. This will increase the > accuracy of the condensate booster pump discharge pressure indication such that normal operation will lie between 25% to 75% , of the indicator's range. This instrument loop provides no control function, only indication. NLSATTC2/SNLICFLR - 86
Attachment to AECM-89/0093-J L. NPE-89-054 l Page 2- ;
, (HED-827) Part 1 L At present the existing CTMT/Drywell Temperature Recorders i (IM71-TJRS-R604A and 1M71-TJRS-R604B) do not have the capability i to reflash their associated annunciators. The consequences derived from this deficiency are that when a single monitored point goes into alarm, (CTMT/Drywell Temp Hi) further monitored ,
k ' points go unannunciated until the original point returns to L normal. The new recorders will have the ability.to reflash their ! associated annunciator windows. Thereby-allowing annunciation for p each subsequent alarming point prior to all points in alarm - cg returning to normal. , The purpose of this design is to replace existing CTMT/Dryweil Temperature Recorders (IM71-TJRS-R604A and th71-tJRS4604B) with two new programmable multi-point temperature recorders, The modification consists of the removal of existing temperature 4 recorders located in panel 1H13-P870 and installation of the new
- recorders. Once the old recorders have been removed the new ~
recorders are to be installed and er.isting thermocouple wiring re-terminated. No modification to alarm design function, Signal source or power source will be performed by this DCP. At present the existing Division I and Division II "Drywell Temp Hi" annunciators have inputs from both an alarm contact within the existing IM71-TJRS-R604A/B recorders and a contact from alarm module trip units (IM71-TSH-N608A, B, C, 0), for drywell temperature E1, 166'. The drywell temperature at El. 166is ' recorded on recorders 1M71-TR-R602A/B and 1M71-TR-R603A/B. Two new annunciator windows will be added for "Drywell Temp. Hi El. 166'" (Division I and Division II). This annunciator will be
. brought in by Alarm Module Trip Units 1M71-TSH-N608A, B, C, D.
L_ The "Drywell Temp. Hi" annunciator will be brought in by an alarm I contact from recorders IM71-TJRS-R604A/B. The addition of these annunciator windows is required because alarm modules l 1M71-TSH-N608A, B, C & D do not have a reflash function to allow l subsequent alarming points to be annunciated. The drywell El. l 166' temperature inputs will remain on recorders 1M71-TR-R602A/B o and IM71-TR-R603A/B (Post-Accident Monitoring Recorders). The
" Containment Temp Hi" annuncitors (Division I and Division II) will continue to get their input from recorders 1M71-TJRS-R604A/B. ,
(HED-827) Part 2 l The purpose of this section of the DCP is to replace existing L recirculation pump (A&B) temperature recorders IB33-TJR-R601 and 1833-TJR-R622 with one new programmable multi point temperature recorder. At present the existing recorders do not have the capability to reflash their associated annunciators. The i consequences derived from this deficiency are that when a single monitored point goes into alarm bringing in annunciators "Recire. E pump /MTR A/B Temp Hi" further monitored points go unannunciated until the original point returns to normal. The new recorders
~
NLSATTC2/SNLICFLR - 87 t 5 , c - - - - . r... _ , - , - . - , .--.i
" s :
Attachment to AECM-89/0093 ; i e NPE-89-054 i Page 3 p l will have the ability to reflash their associated annunciator < i windows. This will allow annunciation of further alarms that may occur prior to all points in alarm returning to normal. The replacement of recirculation pump (A&B) temperature recorders, (IB33-TJR-R601 and IB33-TJR-R622) consists of removal of existing : temperature recorders and installation of one new multi-point s programmable recorder in panel 1H13-P614. Once the old recorders c - have been removed the new recorder will be installed utilizing P existing thermocouple and RTD inputs. The new recorder is capable , of accepting thermocouple and or RTD inputs. The existing-recorders accept thermocouple inputs or-RTO inputs, but not both. , This is why there are currently two recorders, which will be ; e*. - repisced by one recorder. No modification to clarm design Lfunction, signal source or pcwer source will be performed by this t recorder replacer.1ent. , (HED-1088) ! Presently there is no " Diesel Generator Ready to Load" status lamp in the control room. Per HED No. 1088 there is no direct indication in the control room for diesel generator ready to load 9 except line voltage vs. generator voltage. This design will provide a descriptive status light on panel 1H13-P864 for Division e
-I and Division II diesel generators to alert operations that the i diesel generator is ready to load. This status lamp will receive its signal from the same source that the existing local diesel generator. ready to load status lamp receives its signal. ;
REASON FOR CHANGE: These changes were made to implement the changes recommended in the DCROR in response to NVREG-0737. SAFETY EVALVATION: There is no increase in the probability of 4 occurrence or in the consequences of an accident or malfunction of l l equipment important to safety previously evaluated in the Safety l Analysis Report. The addition of the SRV status lamps to the remote shutdown panels ' i will provide operations indication that power is available and the i permissives have been met to open the SRV's. Failure of the status' lamps to either short or open will not initiate a spurious SRV operation. The status lamps will be in parallel with the SRV solenoid coils. Failure of the status lamp in the open condition will return the control circuit to its present design. Failure of the status lamp in the shorted condition will short the solenoid coil and open the fuse feeding power to the SRV's solenoid coil. L The design will not increase the probability of occurrence of an accident, specifically accidents initiated by inadvertent SRV actuation as discussed in Section 15.1.4 and 15.6-1 of the VFSAR. l l NLSATTC2/SNLICFLR - 88 l u
m
,p '
Attachment te AECM-89/0093 l J. [ NPE-89-054 Page 4'
. ; Addition of color filter caps to the radiation monitoring pushbutton selector switch will not increase the probability of !
3 occurrence of'any accident previously evaluated in the UFSAR. The i b . filter caps are a non-electrical, non-active component for which a j credible failure could not increase the probability of occurrence ( of.an accident previously evaluated in the UFSAR. l Presently con'densate booster pump discharge pressure transmitter
,, IN19PTN062 and indicator IN19PIR610 are calibrated from 0-600 i > PSIG. Indicator readings are most accurate when operated between L 25% and 75% of their range. Presently, normal operating pressures ;
are as high as 579 PSIG which is well above 75% of the indicator range. This change will increase.the accuracy of-condensate booster pump ditcharge pressure indication by recalibration of the i 7 . transmitter and changing the indicator scala such that normal , , b . operation will lie between 25% - 75% of the indicatur range. This , instrument loop.is non-safety related and provides no control W function. The function of this instrument loop will not be changed. This change will enhance the accuracy of indication of the pumps discharge pressure for operations. Failure of the " Diesel Generator Ready to Load" status light added by this DCP will not increase the probabflity of occurrence of an accident previously evaluated in the UFSAR. Failure of the status light " shorted" will open the fuse feeding the diesel generator , status light matrix located on panel 1H13-P864. An "open"~ failure will return the circuit back to its existing design. These conditions will not increase the probability of an accident previously evaluated in the UFSAR. Replacement of.CNTM/Drywell and Recirculation Pump / Motor - temperature recorders will not increase the probability of an accident previously evaluated in the UFSAR. .The existing !. monitored points will remain the same. No monitored points will ; be added or deleted. The existing setpoints for annunciation of
~
' H1. temperature will remain the same. The CNTM/Drywell Temperature Recorders (Division I and Division II) will be purchased Class 1E l and will be seismically mounted. The Recirculation Pump / Motor ! Temperature Recorder (non-divisional) will not be purchased Class 1E but will be seismically mounted. Existing signal / power cabling will be utilized for the replacement of these recorders i.e., Reg. Guide 1.75 requirements will be maintained. The recorders are located in the control room therefore there are no environmental l- qualification concerns. The new recorders have a better accuracy L than the existing recorders. Credit is taken for both automatic and manual actuation of the SRV's in Section 15.2 of the UFSAR " increase in reactor pressure".
# The SRV's/ ADS will meet the single failure criteria (per IEEE 279) and perform their safety function to mitigate the consequences of
' ' an accident previously evaluated in the UFSAR. - NLSATTC2/SNLICFLR - 89 { l'
7 ,
~a n Attachment to AECM-89/0093 l n . NPE-89-054 Page 5 p +
~
The filter caps added by this design will help distinguish which
! scale has been selected on the radiation monitors identified in ;
p _ the description. These filters have no active component and will ' not increase the consequences of an accident previously evaluated i in the UFSAR. Per HED No. 337, indicator IN19PIR610 is " pegged offscale during , normal' ops". This design will provide operations with a more 4 accurate indication of concensate booster discharge pressure
%[y during normal operation. . Indication of transient discharge res Per UFSAR Section 15A.2.6 p'nonsures safety will be equipment more accurate.
grade is not used in the FSAR analysis to ;! mitigate accidents. However, when the assumption of a > y non-safety grade equipments use results in more severe ' consequences during an accident than-its malfunction, the
~
non-safety grade equipment is assumed to perform its intended function". Pressure indicator IN19PIR610 is a non-safety grade m
. piece of equipment. This design will not change its intended function and therefore this change will not increase the consequences of an accident previcusly evaluated in the UFSAR.
The status lamps, which will be used by this DCP for " Diesel
- Generator Ready To Load", are presently installed in the control room.. Two 9x9 matrix status lamp assemblies, one per division, ;
for diesel generator status presently use 7 of the available 9 ' l lamps : Spare conductors of existing cables which meet the r equir,ements of Reg. Guide 1.75 will be utilized in addition to existir.g spare contacts for the signal. The jumpers installed : will meet the requirements of Reg. Guide 1.75. The status lamps A will be powered by the circuit which provides power to the ;
; existing 7 status lamps. The circuit is fused such that the added ,
load will not overload the circuit, j L Should the lamp fail " shorted" (worst case) these status lamps for the diesel generator would be lost in the control room. This remains bounded by existing status lamp failures which remains L bounded by previous analysis in the UFSAR. L Replacement of' Temperature Recorders IM71-TJRS-R604A/B, l IB33-TJR-R601 and IB33-TJR-R622 will not increase the consequences L of an accident previously evaluated in the UFSAR. The replacement recorders do not introduce any new failure modes beyond those of the existing recorders. The effects of a failure of the
- replacement recorders remains bounded by previous analysis in the UFSAR. The new recorders are capable of reflashing subsequent alarming points to their associated annunciators. This will make
- operations more aware of a problem with the monitored equipment /
3 plant locations. The existing recorders do not have reflash capability. ( i NLSATTC2/SNLICFLR - 90 l, . m _ _. ___ , _ . . _ - . _ . , , _ .
m Attachment to AECM-89/0093 NPE-89-054 ; Page 6' Failure of the SRV status lamps added by this DCP, in the open ' condition, has no effect on the control of the SRV's. Failure in the shorted condition will have an effect depending on the means by which the SRV actuation signal is generated. Each of the six SRV's' that will have status lights added at the remote shutdown panels have individual power circuits (fuses) for control from the control room and from the remote shutdown panels for both ; divisions of SRV solenoid coils. If the actuation signal is automatic or manual from the control room, a shorted failure will open the fuse, in the control room, suppling power to the shorted divisional SRV solenoid coil but ti,e other divisional SRV solenoid , coil will still open this SRV. If the actuation signal is manual from the remote shutdown panel, a shorted failure will open the ; fuse suppling power to the shorted divisional SRV solenoid coil but the other divisional SRV solenoid coil will still open this SRV. Manual' initiation from the remaining remote shutdown panel , will still open this SRV. These SRV status lamps will be seismically mounted. Status lamps , with the same model number as existing remote shutdown panel handswitch status lights will be utilized and will be purchased class IE quality level 2 The status lamps are designed for anticipated abnormal environments and are qualified to meet IEEE , r 32h1971 per UFSAR Section ?.4.1.4.4 Reg Guide 1.75 requirements for separation. crit erit will be maintained.- The total load on the SRV circuits by the addition of these status lamps is still well below the fuse rating feeding these SRV's. The filter caps will be added to the pushbutton selector switch,
. used to select the scale on radiation monitors presently installed
- on panel 1H13-P604. The equipment contained in 1H13-P604 serve no safety function. Failure of the filter cap such that it falls off the lamp will not increase the probability of a malfunction of equipment important to safety.
The function of the condensate booster pump discharge pressure indicator will remain unchanged. The design will only change the meter scale and recalibrate the existing pressure transmitter. The existing pressure transmitter can be calibrated to 0-1000 PSIG. Since this design is not changing the function of existing equipment, adding or deleting system components, this change will - not increase the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR. UFSAR accident analysis postulate failure of one of the redundant Division I or II diesel generators for which the addition of diesel generator ready to load status lamp design remains bounded. The design will utilize class IE material, tne raceway and status lamp assembly is seismically supported and Reg. Guide 1.75 criteria will be maintained. Thus the proposed activity will not increase the probability of occurrence of malfunction of equipment important to safety previously evaluated in the UFSAR. . NLSATTC2/SNLICFLR - 91
Attachment to AECM-89/0093 m i i NFE-89-054 f page 7 L [ The temperature recorders will be seismically mounted. Existing 1' J. signal / power circuits will be utilized i.e., Reg. Guide 1.75 L , requirements will be maintained. The two CNTM/Dryew11 Temperature ; Recorder (Division I and Division II) will be purchased Class 1E, _ Quality Level 2. Therefore,-the replacement of recorders , IM71-TJRS-R604A/B, IB33-TJR-R601 and iB33-TJR-R622 will not , increase the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR. { The addition of SRV status lamps to the remote shutdown panels i will not increase the consequences of a malfunction of equipment , important to safety previously evaluated in the UFSAR. Should an SRV fail open or closed the status lamp installed in parallel with + the solenoid coil of the failed SRV, even if it should fail will not prohibit operations or an. auto signal from providing a. signal to return the valve to the desired position. , The filter cap performs no active function in a safety ur ' non-safety related system. The filter cap is installed in a non-safety related panel and will not increase the consequences of i a malfunction of equipment important to safety previously evaluated in the the UFSAR. The design for changing the condensate booster pump discharge pressure indicctor scale will not ine.rease the consequences of a malfunction of eculpment important to safety. The scale has no active function. No active component is being added/ deleted from . this instrument loop. The diesel generator status lamps added by this DCP will be in - parallel (powered by the same circuit) with existing diesel t generator status lamps. This circuit performs no control i function, only indication. Failure of the new diesel generator status lamps will ramain bounded by failures of the existing j diesel status lamps which is bounded by previous analysis in the . UFSAR. Upgrading existing temperature recorders to provide reflash capability for their respective annunciators will not increase the consequences of a malfunction of equipment important to safety as previously evaluated in the UFSAR. The new recorders will utilized exiting signal / power circuits; perform the same function as the existing records and be installed using the same design standards as the existing recorders. Failure of the SRV status lights added to the remote shutdown panel (opened or shorted) will not create the possibility of an accident of a different type than any previously evaluated in the UFSAR. Failure of the status lamps are bounded by a short across the associated SRV solenoid coil and an open status lamp circuit. The short would open the fuse feeding power to the SRV solenoid NLSATTC2/SNLICFLR - 92
o g Attachment.to AECM-89/0093 (i [ ,
]
I NPE-89-054 j;;: Page 8 l coil which would disable one division of the SRV. The SRV could 1 still be opened by the other division. The SRV would not open as i a result of the short without operator action or an auto signal. . d An open status lamp circuit would return the SRV circuit back to ! its present design (i.e., no SRV status indication at the remote ' shutdown panel). Failure of the filter caps, i.e., fall off the pushbutton selector switch-lamp, will return the pushbutton selector switch back to ,
. its present design, i.e., lightly colored and difficult to !
identify as blue when lit. l l Changing the scale on 1N19PIR610 and r.ecalibrating the associated pressure transmitter IN19PTN062 to obtain proper operation of this instrument loop will remain bounded by existing accident analysis ' previously evaluated in the UFSAR. The nw status lamps will be added to existing diesel generator ' status lamp circuits. Failure of the new diesel generator status
- lamps remains bounded by failure of the existing diesel generator
- status lamps which is bounded by previous analysis in the UFSAR.
The recorders which will replace existing CNTM/Drywell and i Recirculation Pump / Motor Temperatura Recorders will perform the i same design. function as the existing recorders. The new recorders will provide a reflash capability to their respective annunciators l should subsequent monitored points rise above their alarm set point. The existing recorders do not provide this function. This a additional function will help maintain the awareness, of a ' potential problem, for operations. Failure of these temperature recorders,will not create the possibility of an accident different than failure of the existing recorders which is bounded by existing UFSAR accident analysis. The single worst failure that could be postulated by the addition ( of the SRV status lamps would be a status lamp short in the event of a control room evacuation. Upon operator action to open the SRV's from the remote shutdown panel with the shorted status lamp; the. fuse feeding the SRV solenoid coil from this division would i open the circuit. Therefore operations would have to utilize the ' redundant remote shutdown panel to lift this SRV. Per UFSAR Section 7,4.1.4.5.3 "In the event of a loss of one ESF division of power as the single. failure, cold shutdown is possible by using the controls and monitoring instrumentation provided on the remaining operable remote shutdown panel." Therefore the possibility of a malfunction of equipment important to safety different than previously evaluated in the UFSAR will not be created. i
, NLSATTC2/SNLICFLR - 93
Attachment to AECM-89/0093 ! y > NPE-89-054 Page 9 i
-The filter caps.have no interface with safety related .
systems / components. Failure of the filter caps as discussed above ! will not create the possibility of a malfunction of equipment l
.important to safety different than previously evaluated in the -
-UFSAR.
Failure of the condensate booster pump discharge pressure indication instrument loop, after the replacement of its indicator scale and recalibration of its transmitter will remain bounded by ; failure of the existing instrument loop. Therefore, failure of this instrument loop after impit; mentation of the design will ' remain bounded by previous analysis in the UFSAR. j As stated above, the status lamp design for " diesel ready to load" will utilize existing status lamp assemblies, circuits, etc. presently utilized for diesel generator operation. Per UFSAR t Section 15A.6.3.3 under event # 29 - Loss of offsite power, there ,
" are a variety of possible plant - network electrical component ;
failures which can affect reactor operation. The total loss of offsite power is the most severe. Per UFSAR Figure 15A.6-29
" Protection sequence for loss of normal AC power" the standby AC power systems are designed with single failure criteria.
- Therefore this design will not create the possibility of a malfunction of equipment important-to safety of a different type than any previously evaluated in the UFSAR. ,
This design will replace existing temperature recorders with temperature recorders that meet the same design standards i.e., - seismic, separation, environmental qualification, accuracy as the existing recorders. Therefore, replacement of these recorders will not create the possibility of a malfunction of equipment important to safety different than the previously evaluated in the UFSAR. , Addition of Class IE, seismically mounted SRV status lamps to the , remote shutdown panels will not reduce the ability to " shutdown and maintenance of hot shutdown of the unit from locations outside of the control room" (Ref. Technical specification bases 3/4.3.7.4). The addition of " diesel generator ready to load" status lamps in the control room (Division I and Division II) will utilize , The design will meet existing equipment installed in the plant. the separation requirements of Reg. Guide 1.75. Failure of the status lamp will not reduce the ability of the diesel generator system from ensuring that " sufficient power will be available to supply the safety related equipment" (Ref. Technical Specification bases 3/4.8.1, 3/4.8.2, 3/4.8.3). NLSATTC2/SNLICFLR - 94
I l (- Attachment to AECM-89/0093 j l' NPE-89-054 i Page 10 I - The addition of blue filter caps to the pushbutton meter. scale ; p selector switch on the radiation indicating meters on panel i j' 1H13-P604 will not reduce the margin of safety as defined in the basis for any Technical Specification. These filter caps perform no active function; the meters will function with or without these filters. Changing the scale on pressure indicator 1N19-PI-R610 and recalibration of its associated pressure transmitter IN19-PT-N062 will not reduce the margin of safety as defined in the basis for j any Technical Specification. The indicator will still provide , indication in the control room for condensate booster pump-discharge pressure. The change will provide for indication in the mid-scale range of the indicator and not in the top scale range ; which is less accurate. Replacement of temperature recorders 1M71-TJRS-R604A/B, IB33-TJR-R601 and IB33-TJR-R622 with upgraded temperature recorders will not reduce the margin of safety as defined in the bases for any Technical Specification. Per Technical - Specification sections 3/4.6.1.8 and 3/4.6.2.6, the bases for the r limits on containment and drywell average air temperatures is to , not' exceed design temperatures during LOCA conditions, consistent with safety analysis. This change will have no effect on temperature limits established in the Technical Specifications. P [ P i NLSATTC2/SNLICFLR - 95
!T-M , y Attachment to AECM-89/0093 ,
c 5 SRASN: NPE-89-055 000 NO: DCP-87-0055-S00-R00 SYSTEM: V20 , L' DESCRIPTION OF CHANGE: Access to the condenser waterboxes is
- obtained through manways located on the top and the bottom of each !
~
waterbox. Access is required to perform work inside the waterbox and to perform testing and inspections as specified in VFSAR ! Section 10.4.1.'4. To avoid personnel safety hazards, extensive scaffolding must be erected. Valuable manhours must be used to erect and to remove this scaffolding which causes the duration of i the= job to be lengthened. This change provides access to the condenser waterbox manway with permanent ladders and platforms. , I REASON FOR CHANGE: This change removes the personnel safety hazards and eliminates the manhours required before and after the ' required-job for scaffolding erection and removal. The permanent access platforms and ladders will also reduce the number of manrems expended during the duration of the job. SAFETY EVALVATION: There is no increase in the probability of occurrence or_ in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety i Analysis Report. The platforms are being installed to provide safe access to the : waterbox manways. The platforms, structural steel, and waterboxes have been evaluated in Calculation CC-NIU20-88022, Rev. O and found to be structurally (dequate. The attachments to the condenser waterbox stiffeners have taken into account the thermal growth of the condenser by providing slotted holes. The Turbine Building still maintains the ability of having no adverse effects on adjacent Category I structures with the addition of the platforms, i The platforms are being installed to avoid personnel safety hazards. The platforms were designed to meet AISC Codes, Uniform
-Building Codes, and OSHA Standards which meet all the requirements for Turbine Building internal structures. In addition, the design - considered thermal effects on the condenser and waterboxes and ensures their structural integrity.
The platforms have been designed for a 100 psf live load, dead load, and VBC seismic loads and are adequate for their intended use for access to and from the condenser waterboxes. The platforms are also designed to allow thermal movement of the condenser during operation of the plant. The access platforms are not equipment important to safety, but , are designed to maintain their structural integrity. The Turbine Building with-the additional platforms is structurally adequate and will still have no adverse effects on the adjacent Category I structures. This design change considered the seismic and thermal effects on the condenser and in no way impair its ability to function as intended. NLSATTC2/SNLICFLR - 96
(- Attachment to AECM-89/0093 NPE-89-055 Page 2 l Therefore, there is no creation of a possibility for an accident i or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Access platforms are not addressed in the GGNS Technical I Specifications. Therefore, the installation of the platforms will . not reduce the margin of safety as ovfined in the basis for any i Technical Specification. b I
'{
NLSATTC2/SNLICFLR - 97
g - Attachment to AECM-89/0093 i _. 9 SRASN:- NPE-89-056 DOC NO: DCP-87-0090-500-R00 SYSTEM: U20 [. DESCRIPTION 0F CHANGE: This change-provides for a stairway type tower (approximately 39 ft. tall) to be installed on the Turbine
- Building communication bay roof elevation 206'. The tower will be supported by the Turbine Building roof and the Turbine Building siding structural support steel. The design is non-seismic, and . non-safety related, however, the affected Turbine Building steel was analyzed ~ for the tower loads imposed by a 100 mph wind. This change will not require a revision to FSAR Section 13.6 or the r Grand Gulf Physical Security Plan, REASON FOR CHANGE: This change was made to allow security response to fence alarms on the Turbine Building roof elevation 232' from the Control Building.
I" SAFETY EVALVATION: There is no increase in the probability of occurrence or in-the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the Safety i Analysis Report. The only credible accident scenario related to t- this change is due to tornado generated missiles and this tower is bounded by the existing design basis missile analysis (Ref. FSAR Section 3.5.1.4). 'The tower is not a seismic II/I hazard and does not create a new design basis missile.
'The affected Turbine Building structural steel has been analyzed for the new loads an:1 found to be adequate. In the event of a ;
tornado, the siding support steel is not postulated to remain in place (Ref FSAR Section 3.8.4.1.2.1). The only significant loads , are those'due to wind. Therefore, the probability of malfunction ! I of the Turbine Building is not increased. The consequences of , tower failure have no impact on safety as it is not a seismic II/I hazard. 'l The tower provides an alternate method for security response to alarms and its failure is not important to safety. The consequences of Turbine Building failure are not changed by the < addition of this tower, l The tower is a passive non-mechanical structure that does not interact with any plant systems or structure other than the Turbine Building, and the Turbine Building has already been evaluated for accidents. Since the tower only affects the Turbine Building and the Turbine l Building has been analyzed for the new loads and the tower does not create a new design basis missile, no new malfunction scenario is created. Malfunction of the tower itself is not important to safety as it is not a seismic II/I hazard. I NLSATTC2/SNLICFLR - 98 s
, y, . . - - , .,_ , - - , . . - _ ~ _
h;' 3: o . ,- Attachment to AECM-89/0093 i I i- . L: ! L NPE-89-056 : L Page 2 l l Therefore, there is no creation of a possibility for an accident l l or malfunction of a different type than any evaluated previously l l- in the' Safety Analyt,is Report. , Neither the tower nor the Turbine Building steel is used as a i basis to define any margins of safety for the Technical 3 Specification.
- o f
l- . lJ .6 4 k' i l l l
-NLSATTC2/SNLICFLR - 99
y . L Attachment to AECM-89/0093 SRASN: NPE-89-057 DOC NO: DCP-87-4001-S00-R00 SYSTEM: B21 i DESCRIPTION OF CHANGE: The purpose of this design change is to : j' 4 upgrade equipment required for RPV level monitoring in the Fuel Zone and Shutdown Range Instrument loops. i
. Reg. Guide 1.97 requires redundant instrumentation for RPV water I level instrumentation ranging from the bottom of the core support F plate to the ceriterline of the main steam line. The existing GGNS l
- instrumentation measuring this required range consists of three .
! zones (Wide Range,. Fuel Zone, and Shutdown Range) with overlapping ranges. : o Currently, the Wide Range instrumentation provides a trip function , for the ECCS and meets Reg. Guide 1.97 requirements. However, the ! Fuel Zone and Shutdown Range Instruments, used only for L monitoring, are not seismically or environmentally qualified.
- L Furthermore, redundancy requirements are not met by-the Shutdown Range instruments, and neither Fuel Zone nor Shutdown Range instrument loop cables and wiring are Class IE.
- As a result, SERI-has committed to make the appropriate modifications to ensure compliance with Reg. Guide 1.97. The e existing instruments affected are Fuel Zone (B21-LT-N044C&D, B21-LR-R615, and B21-LI-R610) and Shutdown Range (B21-LT-N027 and ,
B21-LI-R605). The three level transmitters installed are i Rosemount 1152D and will be upgraded to 1153D models to assure i qualifications to 10CFR50.49 and IEEE 344-1975. An additional Rosemount 1153D transmitter will be added to the Shutdown Range ' instrumentation to meet the redundancy requirements of Reg. Guide 1.97. Also, the Fuel Zone level recorder will be replaced with a : qualified recorder receiving inputs from the Division 1 Fuel Zone and Shutdown Range transmitters. A second two-pen recorder will replace the level indicators (B21-LI-R605 and B21-LI-R610) in the , P601 panel and will receive the inputs from the Diviston 2 Fuel Zone and Shutdown Range transmitters. These two pen recorders (B21-LR-R615A,B) will be identified as Post Accident Monitoring (PAM) instruments and, in conjunction with the wide range recorders (B21-UR-R623A,B) will satisfy the Reg. Guide 1.97 range requirement from the bottom of the core support plate to the i centerline of the main steam line. In addition, the Fuel Zone and Shutdown Range power cables will be replaced with Class 1E cables, , s and the power supplies will be upgraded to Class 1E, UPS. This upgrade of Fuel Zone and Shutdown Range instrumentation will increase the reliability of the current system by providing qualified, redundant, Class 1E equipment to the Nuclear Boiler System. REASON FOR CHANGE: This change was made to upgrade equipment required for RPV level monitoring in the Fuel Zone and Shutdown Range instrument loops. NLSATTC2/SNLICFLR - 100
gz, _ Attachment to AECM-89/0093 NPE-89-057 Page'2 3 SAFETY EVALUATION: There is no increase in the probability of L . occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. The instruments modified by this design change will be upgraded to the category one requirements of Reg. Guide 1.97. Overall system reliability will be increased due to the improvements in redundancy, qualification, and power supply availability. Also, no trip functions are affected; only post accident indication is improved. i
' Equipment upgraded in this DCP is not currently addressed in the ;
FSAR as equipment used to mitigate the consequences of an ; accident. However, Section 7.5, " Safety-Related Display
, Instrumentation" will be revised to address not only the Wide ,
- Range level instrumentation but also the Fuel Zone and Shutdown Range instrumentation.
No malfunction of equipment important to safety previously evaluated in the FSAR is predicated on a failure of the instrumentation affected by this DCP. This upgrade of the Fuel Zone and Shutdown Range instrumentation meets all the r qualification and divisional separation requirements for Class IE circuits and equipment. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in
;the Safety Analysis Report.
The instrument upgrade required by this design change does not : alter the original design scheme of coolant level measurement. The transmitters replacing the Rosemount 11520s are qualified.for improved radiation exposure conditions. The additional transmitter will provide a redundant channel for Shutdown Range indication. The level recceders will provide redundant, e divisional level indication. No' trip functions are affected by l this design change. The original level monitoring function of l these instruments remains unchanged; no new failure modes are , I- created. This upgrade of the Fuel Zone and Shutdown Range instrumentation meets all the qualification and divisional separation requirements for Class 1E circuits and equipment. L Plant systems, isolation, and ECCS actuation instrumentation, as . (_ well as RPS instrumentation are not affected by this design L change. No trip functions are affected by this design change. Therefore, the margin of safety as defined in the basis for any L Technical Specification is not affected. NLSATTC2/SNLICFLR - 101
,7 7
- Attichment't3 AECM-89/0093 ys q
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SRASN:: NPE-89-058 DOC NO: DCP-87-4018-S00-R01 SYSTEM: . P33 DESCRIPTION OF CHANGE: _ The liquid and atmospheric sampling , subsystems of the Post Accident Sample System (PASS) are installed
^
in=the PASS grab sample panel (N1P33-P001). Both subsystems [: require modifications to 1:nprove the overall system's ability to . meet the requirements of NUREG-0737. This panel, in its present
~
condition, is a sionificant maintenance issue due to its
' overcrowded condition and poor ALARA design. i
-DCP 87/4018 is to install a new atmospheric sampling cystem in a j separate panel' totally dedicated to atmospheric sampling. The new I panel will be located an the PASS room adjacent to the existing ;
IP33-P001 panel. The' existing atmospheric grab sample-point is designed such that the operator is exposed to approximately 36" of unshielded sample line. These lines are the sample inlet and outlet line between i [ the cask and panel. To reduce operator exposure, the method for taking an atmospheric grab sample is being modified. The existing , atmospheric grab sample point will be eliminated. The new atmospheric grab sample point will consist of a replaceable septum l located behind a removable' lead plug in a lead block shield wall. . This lead plug will be fitted with a guide tube which will allow the operator access to the septum via a long needle and syringe. l The' sample will'be drawn into a lead shielded syrir.ge attached at one end of a two' foot mechanical arm. The syringe will be opened
,and closed by the operator from the opposite end. Design and fabrication of the mechanical arm shall be the responsibility of i Plant Staff. The arm shall be fabricated to meet plant +
requirements. [ Twenty-four hours following the onset of an accident, a drywell h and containment atmospheric sample is required to be taken via the PASS. Per Environmental Specification 15026-E-100.0, Rev. 4 at ' the time the first sample is required the drywell sample _will be saturated steam at 250**F and 30 psia. The atmospheric sample ' inlet lines are heat traced to minimize condensation and the loss of particulate sample. However, because the sample is saturated it is likely that some minimal condensation will occur. Any j condensation that could occur for whatever reason will collect in the sample panel since the panel is the low point in the system, f At present, the system provides no means for removing condensate. A sufficient buildup of condensate could reduce system effectiveness. , o l-l k NLSATTC2/SNLICFLR - 102 1
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4 , Att:chment'to AECM-09/0093 e .
!?H h 1 gm,' '
NpE-89'-050, page.2' e The ner atmospheric _ sample system design will include a sample cooler, a condensate diversion tee, and condensate' collection tank. The sample will'be cooled from 250"F to approximately 105"F downstream of.the grab sample-point. The diversion tee will: separate the condensate from the atmospheric sample. The collection tank will include a level switch and local alarm to make-the operator aware when the condensate tank should be pumped co down., The condensate will'be pumped to the suppression pool'via the existing liquid system return line. . The atmospheric sample L return line'is presently-heat traced to 125"F.- This will-further i reduce the possibility of condensate forming in the atmospheric- i [' sampling system. l l _The sample _coolerL i s being added to reduce the power consumption of the heat tracing _on the sample return lines necessary to prevent condensation from occurring. Connections provided by DCp _
'87/4006 in the component cooling water (CCW) system are available ;
for.providing cooling water for the sample cooler. , i Heat tracing is presently installed on the drywell-and containment j sample inlet lines and on the common sample return line. Each -! line has different temperature requirements. The new atmospheric sample panel and affected sample lines will be heat traced by- 1 splicing into the existing heat tracing circuits. The overall .I' heat tracing circuit lengthe will not exceed the maximum circuit lengths as specified by the vendor. ,
, I provisions are being provided to flush the condensate portion of !
the new atmspheric system with demineralized water and to purge l the sample inlet and outlet lines with instrument air. ; i NUREG-0737, Item Il.B.3, Criteria lla and the clarification of 11a i addresses the need for heat tracing and purging of sample lines. _'a A The clarification also states that purge velocities should be
' considered. This DCp finalizes modifications required to the heat [
tracing system. The heat tracing system has been designed :to 1 minimize the loss or distortion of the atmospheric sample to ! ensure that grab samples are representative. Air purging of the sample lines is accomplished via the instrument air system with a system pressure of approximately 100 psig. The atmospheric sampling subsystem, the solenoid banks and the ! liquid pump starter are being removed from the PASS sample panel (1p33-p001) to eliminate the overcrowded condition of the panel and thus better accommodate maintenance activities. Drip pans are being installed in the bottom of the 1p33-p001 panel to collect and contain any minor system leakage and thus, restrict any possible contamination due to leakage into the panel. i NLSATTC2/SNLICFLR - 103 _ 4
f? ' Ql ,' O ' g, , f Att:chmentEto AECM-89/0093 4 _NPE-89-058
'Page 3 h.'
Theli n-line instrumentation in the liquid sampling subsystem, conductivity, pH-and oxygen monitors, are being replaced with new e Jand~ improved instrumentation...A hydrogen monitor is.also being b added-to the in-line instrumentation to provide additional
-sampling capabilities-for dissolved gases.
~
REASON FOR CHANGE: This change was made to improve the overall y system's ability:to meet the requirements of NUREG-0737.- ~
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SAFETY ~ EVALUATION: There is no increase in the probability of occurrence or in-the consequences of an accident or malfunction of ~'y equipmant.important to safety previously-evaluated in the Safety Analysis Report. PASS has beenLdesigned in accordance with the requirements of the GGNS operating license, the Technical Specifications'and as cornitted to in the UFSAR. This DCP provides enhancements to PASS which will enable the system to better perform its design functions. The design has considered and incorporated those pertinent design features which will prevent the failure of PASS-from affecting systems required for safe shutdown of the plant , following an' accident. The PASS design has also incorporated j
-design features for ALARA considerations. ,
, Therefore,-there in no creation of a possibility for an accident !
or malfunction of a different type than any evaluated previously in the Safety Analysis 1 Report. The margin of safety as defined for any Technical Specification remains unchanged. The Technical Specifications are not affected.- PASS has been designed in accordance with the requirements of the ( GGNS operating license, the Technical Specifications and as
-committed to in the UFSAR.
u f ? .> k NLSATTC2/SNLICFLR - 104
9 m g g7u ip Attachment to AECM-89/0093 A l . e4 e 1
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L SRASN:--HPE-89-059 DOC Not. DCp-88-0002-S00-R00 SYSTEM: LN32 ! [ , i
~ DESCRIPTION OF CHANGE: The turbine _overspeed protection system '
! -does not perform a safety function.. Furthermore,.the turbine overspeed control' system equipment, electrical cables, and *
' hydraulic lines are not_ required to safely shut down the reactor.
~
F However,_should an overspeed-condition exist, missiles which could damage. equipment'important to safety might occur. Therefore, the turbine is provided with a highly reliable and-redundant . mechanical overspeed trip system to trip the turbine in the event j ) of a turbine overspeed condition.
/To ensure reliability,_the mechanical overspeed trip system is
, tested every 14 days with'the automatic turbine test program. E When tested with the automatic turbine test program, an. electric backup overspeed trip is in effect to prevent overspeed. Should the automatic turbine test fail during performance of the test, i the mechanical overspeed trip system may need.to be tested using 1the controls from the turbine front standard. This alternative method of testing the mechanical overspeed trip system requires that the actual overspeed trip be hydraulically bypassed during the= test. Under these circumstances, the turbine is operated for
.a1short period of time without overspeed protection. This DCP provides electrical turbine overspoed trip protection to the turbine at all times when the automatic turbine tester panel is !
energized with the key lock switch. i The: existing electric overspeed trip system is limited to_ tripping the turbine while the mechanical system is being tested with the automatic turbine test. Thic DCp will modify the enisting , electric overspeed trip system to be in effect at all times. This will be done by adding relays d118 and d218 which will be energized in a overspeed condition. When these relays are energized they will energize solenoid valves SJ24-SO41 and 8042 (one solenoid valve and one relay per channel) which drains the oil pressure from the main trip = system causing _the turbine to trip. Solenoid valves SJ24-SO41 and SO42 are the valves currently used when the turbine trip pushbutton is used. A relay d119 will be added to allow blockage of the extended electrical trip system so as it.may be tested during the automatic turbine test program.
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The electric overspeed trip system will also be changed to trip at a speed.112%. This will allow for the mechanical trip system to be taken out of service and tested by tripping it at 110% to
, verify its propor operation while the turbine is provided with electrical overspeed trip capability although the turbine is in operation.
NLSATTC2/SNLICFLR - 105
c v Y , . Attachment to AECM-89/0093 ~ t d. .
'NPE-89-059 Page 2' ;
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It must te sted that implementation of this design change will not degrade the reliability of the turbine trip system. No. component or system which currently provides a turbine trip signal
- l will be-degraded by,this design change. Consequently ~, the turbine will continue to be protected and will continue to be tripped when necessary. However,.since the electric overspeed trip'will.be in effect at all times, failure of the electric overspeed trip system-
~
can potentially cause a turbine trip which is not necessary and 1 thereby decrease plant availability. Consequently, special
~
attention must be given to maintaining the components associated . with the electric overspeed trip system following implementation of this design change. REASON FOR CHANGEi This change provides for electrical turbine
/
,overspeed trip protection to the turbine at all times.
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 evaluated in the Safety Analysis Report. . The design change implemented'herein is consistent with the design of other instrumentation which also provides-a turbine trip. The ; electrical overspeed trip system provided by this design is a redundant, two channe1~ system which only requires one channel to provide a~ turbine trip when required.' This design is identical in
- redundancy and channel independence to other turbine trip i instrumentation such as the reactor water level 8 turbine trip.
4 Therefore, the reliability of the turbine trip on an overspeed condition is equal to or better than the turbine trip for other
. monitored conditions.
A turbine trip is evaluated in Chapter 15 of the FSAR. As. stated , above, only one overspeed trip system channel is required to function to provide a turbine trip when required. However, since this design is identical in redundancy and channel independence to other turbine trip instrumentation, the probability of initiating an unnecessary turbine trip is no greater for this electric overspeed instrumentation than for a turbine trip initiated by instrumentation. The addition of this electric overspeed trip system to be in effect at all times does not alter the existing mechanical overspeed trip. Furthermore, neither any other turbine trip signals nor the turbine trip system as a whole is altered in any way by this design change. Consequently, the consequences of an accident previously evaluated in the FSAR are not increased. Therefore, the capability to trip the turbine when required is not affected by this design change. Consequently, the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR are not increased. NLSATTC2/SNLICFLR - 106
py n . Attachment to AECM-89/0093 H 1 i ty . t . NPE-89-059-Page 3 ; There:is no creation of a possibility for an accident or , malfunction of- a .different type than any eva19ated previously in the Safety Analysis Report. - L The electric overspeed trip system operates. independently of the mechanical overspeed trip system and trips the. turbine via existing turbine trip fluid solenoids. The electrical power supplies, relays, cables, etc. added or altered by this design change are not safety related. Furthermore,:this design change . o does not alter the design function of any component to impose any J new safety related design function. . t The basis for Technical Specification 3/4.3.9 is to ensure that: t the-overspeed trip system is available to trip the turbine ~to prevent missiles which might damage equipment important to safety.. This addition of this electric overspeed trip system to be.in-affect at all: times ensures that overspeed protection is provided even when the mechanical overspeed trip system is removed from
-service for' testing. Consequently, this design change only-ensures that the margin of safety implied by this Technical Specification is not reduced when the mechanical overspeed trip system is removed from service for testing.
i ( l' t t NLSATTC2/SNLICFLR - 107
Attachmont to AECM-89/0093 w i m
'tRASN:: NPE-89-060 DOC NO: DCP-88-0008-S01-R00- SYSTEM: E12 DESCRIPTION 0F CHANGE: This DCP, in conjunction with the Base DCP
,1 .88/0008 and supplement DCP's 88/0008-2.thru 88/0008-7,. provides the design details for the installation of the Alternate Decay Heat Removal System'(ADHRS). This supplement provides the design for.the ADHRS suction piping and pipe supports. The tie-in for the 12" ADHRS suction line is upstream of valve G41 F057 in line 18" HBC-222. This tie-in is
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. located in the RHR "C" Room. From this point, it is routed into the. Piping Penetration Room adjacent to the RHR "C" Room where it branche's to the two new ADHRS pumps. The installation of the suction piping will require the addition of 6 new supports and the i modification of one existing support.
This' supplement.provides the design for the new valve G41 F348,
-which is being provided to minimize the amount of piping requiring lead shielding by isolating the branch line,18" HBC-229, from the !
ADHRS suction path (ref. DCP 88/0008-2 for lead shielding i requirements). Additionally, 5 pipe supports of Stress Problem 45
' requiring modifications as a result of valve G41 F348 and the lead shielding being added are included'in this DCP. y 1
This safety evaluation addresses only piping and pipe support modifications resulting from the installation of the ADHRS suction line and the valve G41 F328. This safety evaluation does not j address the operation or function of ADHRS and its impact on the' j FSAR.' .i REASON FOR CHANGE: Thfs change minimizes the amount of piping requiring lead shielding by isolating the branch line, 18"
-HBC-229, from the ADHRS suction. path.
1 SAFETY EVALUATION: There is no' increase in the probability of , occurrence or-in the consequences of an accident or malfunction of i equipment important to safety previously evaluated.in the Safety j Analysis Report. The piping and pipe supports designs meet ASME Section III l requirements and are qualified as seismic category I. The 'I addition of the piping and pipe supports does not affect the
. integrity of any other safety' system. The piping and pipe supports will function in their intended manner. ,
4 Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. , The installation of the piping and pipe supports does not change , the limiting conditions for operation applicability or surveillance requirements. The piping and pipe supports meets ASME Section III requirements and will function in their intended manner. Therefore, this change will not affect the margin of safety. NLSATTC2/SNLICFLR - 108 4
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y o > - 3 x . F' , Attachment'tolAECM-89'0093/ a y
;l n SRASN: NPE-89-061: : DOC NO: DCP-88-0008-502-R00 SYSTEM: E12 tu -
. LCESCRIPTION~0F CHANGE: This DCP, in conjunction with the Base DCP 88/0008 and supplement DCP's 88/0008-1 and 88/0008-3 thru
^
88/0008-7, provides the design details for the installation of the - , Alternate Decay Heat Removal System (ADHRS).. iThis supplement provides design modifications to the existing RHR and FPCCU piping which will be utilized as the suction line of the ADHRS. These design modifications include. installation of radiation shielding on piping, installation of motor operators on valves Q1E12F066A and B, and modification of pipe supports. The installation-of lead shielding is on existing RHR and FPCCU piping-located in the main corridor in Areas ~7 and 9 at Elevation 93' of the Auxiliary Building. The shielding is necessary to reduce radiation' levels when the ADHRS is operating. The~ installation of motor operators on F066A and B allows various flow paths to be established for ADHRS. OCP 88/0008-7 provides
.the' electrical power and controls.for operating the valves, h .The pipe support modifications are necessary as a result of the-additional weight imposed on the piping by the shielding and valve ll operators.
This safety evaluation addresses only the aforementioned modifications. This safety evaluation does not address the operation or function of ADHRS and its impact on the FSAR. REASON FOR CHANGE: This change. modifies existing RHR and FPCCU
. piping to be used as the suction line of the ADHRS.
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 evaluated in the Safety Analysis Report. The piping and pipe supports designs meet ASME Section III requirements and are qualified as seismic category I. The - modifications to the piping and pipe supports do not affect the integrity of the interfacing piping or any safety systems. The piping and pipe supports will function in their intended manner. This DCP does not change or degrade or prevent actions as described or assumed in the FSAR. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 109
4 2 7g D, '
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.. Attachment to~AECM-89/0093~
c 't t
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NPE-89-061:
.Page 2i l
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The ' modifications to the' piping and pipef supports do not change the limiting conditions for operation: applicability or-p . surveillance requirements. - The ~ piping'and pipe supports: meet ASME
- t-- LSection III requirements and will: function in their intended ::
., manner. Therefore,1this-change willinot affectLthe margin of '
esafety. i
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I? I I. Id t e NLSATTC2/SNLICFLR - 110
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[- g 4 Atttchment to AECM-89/0093 gs , s
.q+
L FW SRASN:. NPE-89-062' DOC NO: DCP-88-0008-S03-R00 SYSTEM:'.E12 y A ~ DESCRIPTION OF CHANGE: _ This'DCP,-in-conjunction with the Base DCP
,, 88/0008 and supplement.DCP's 88/0008-1, _2 and- 4 thru 7, willE C ; provide the design' details for the' installation of the Alternate Decay'HeatLRemoval System (ADHRS).
This supplement provides design details for the installation of the-ADHRS piping and pipe supports from the ADHRS Heat Exchanger
. discharge to'the RHR' tie-in. .This piping is located in.the piping
. penetration room adjacent to the RHR "C" Room. Tne piping design 1
, consists of one B" line from each of the two ADHRS Heat Exchangers where it joins into one 10" common line prior to entering into line 18" GBB-58 downstream of E12 F029C. This discharge piping will require the installation of 6 new supports'and tho'
I
; modification of one-existing support.
, REASON-FOR CHANGE: This safety evaluation addresses.only piping and pipe support modifications resulting from the installation of the ADHRS discharge line from the ADHRS Heat Exchangers to the RHR tie-in. This safety _ evaluation does not address the operation or function of ADHRS and its impact on the FSAR. .l 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 evaluated in the Safety
' Analysis Report.
The piping and pipe supports designs meet ASME Section III ! requirements and are qualified as seismic category I. The addition of the piping'and' pipe supports does.not affect the
, s integrity of any other safety system. The piping and pipe supports will function in-their intended manner. l Therefore', there is no creation of a possibility for an accident ,
J- or malfunction of a different type than any evaluated previously ! in the Safety Analysis Report. l 1 The installation of the piping and pipe supports does not change ; the_ limiting conditions for operation applicability or j'
' surveillance requirements. The piping and pipe supports meets ASME Section III requirements and will function in their intended manner. Therefore, this change will not affect the margin of safety.
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'NLSATTC2/SNLICFLR - 111 4
d
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$ s Attachmsnt to AECM-89/0093' L (
, .SRASN: 'NPE-89-063 DOC NO:' DCP-88-0008-504-R00 SYSTEMi T21 DESCRIPTION.0F CHANGE: The DCP scope includes the' installation of support pads and anchor bolts for-the Alternate Decay' Heat Removal-i , System (ADHRS) pumps and heat exchangers at Elevation 93'-0" of-
'the Auxiliary Building. Rigging: instructions for handling the-ADHRS heat'exchangers, pumps, AC-Unit, radiation monitor, and other v'alves, pipe spools,.or miscellaneous equipment to be lowered through the RHR_C hatch are provided.in the DCP. ,
p" x Details are provided to implement permanent .and temporary structural modifications required >to provide rigging attachment points to.the existing Auxiliary Building structure. Modification of cable trays in the RHR C pump room to facilitate future rigging L efforts is-also provided for in the DCP.-
. NUREG 0612, " Handling of Heavy Loads at Nuclear Power Plants",
s identifies two acceptable methods.for-evaluating the handling of heavy loads. 'ihey are:to take steps to assure that' load mishandling will not occur or to evaluate the consequences-of mishandling, Compliance with the NUREG is assured during the implementation:of the-DCP by limiting the areas in which equipment can be; handled, by evaluating the affect of unanchored equipment in.the anticipated locations, and by evaluating the structural
-integrity of-the Auxiliary Building structure under the worst case load drop scenario. In addition, compliance is further assured-by the requirement that the operational impact of equipment-damage
.possible during load handling be-evaluated as part of the implementation work package prior to implementing rigging.
instructions provided in the DCP. REASON FOR CHANGE: Implementation.of the instructions are evaluated in this safety evaluation only for affect on building structural integrity. The DCP requires that evaluation of the actual-rigging evolution for' operational effects be performed as part of.the implementation work package. F SAFETY EVALUATION: There is no increase in the probability of occurrer.ce or in the consequences of an accident or malfunction of
- equipment important to safety previously evaluated in the Safety -
Analysis Report. The presence of the equipment temporarily unanchored at its design L location prior to final bolting and alignment does not create a ? hazard to other plant systems in such a way that previously evaluated accident occurrence probability increases since no safety related equipment is located in close proximity to the new i equipment. l. p 1 L L NLSATTC2/SNLICFLR - 112
y: , q , iAttqchment-to'AECM-89/0093) s 4
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.NPE-89-063 Page 2:
4 The presence of unanchored equipment cannot increase the - y' . consequences of previously evaluated. accidents due to their' lack of proximity.to other_ safety-related equipment.' In addition,'
-floor' loads have-been found to'be. acceptable along the load paths.
to beLtaken during equipment movement.
'The presence of.the ADHRS equipment temporarily unanchored.does
-not increase the probability of equipment malfunction due to displacement in a seismic event _since it-is either physically isolate.from or lacks proximity to other safety related_ equipment when installed at.the design location. . Cable. tray and support modifications are designed to ensure cable-tray integrity in a ,
. seismic event. The worst case load; drop postulated along the !
' assumed load path was' evaluated and determined to not' jeopardize j U ~ building. structural' integrity.
a
'The consequences of previously evaluated safety related equipment
. malfunction are not increased since the Auxiliary Building structure has been determined adequate to resist in all postulated loads imposed during. implementation of the DCP.
The Auxiliary Building structure is evaluated and modified as -l
. required to.' support rigging operations described in the DCP.
Guidance.is also given with respect to floor. loading to prevent exceedance of' design live load during load handling. Due to lack
~
of proximity to safety related system components, no new 11/1 ! O . hazards-are created by the presence of the subject unanchored .; ADHRS equipment. !
.o Therefore, there is no creation of a possibility for an accident j or malfunction of a different type than any evaluated previously.
f in the Safety Analysis Report. The margin of safety for technical specification bases 3/4.5.1 is ; not reduced since the implementation of action statements are not l affected by cable tray modifications and surveillance requirements can still be met. No'other technical specifications are affected 1 by the DCP. ]
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NLSATTC2/SNLICFLR - 113
T
^- @ . 'Attachmont to AECM-89/0093' q
SRASN: iNPE-89-064 DOC'N0: 'DCP-88-0008-S05-R00' SYSTEM: P44 3
, DESCRIPTION' 0F CHANGE: This DCP, .in. conjunction with the Base,DCP 88/0008'and supplement DCP.'.s 88/0008-1.thru 4, 6 and 7, provides the design details for the installation _of the' Alternate Decay-Heat Removal System (ADHRS). '
u This supplement provides the design of'the Plant Service Water ~ System piping.and pipe' supports to,the ADHRS Heat Exchangers. Also,Cthis supplement provides for the designs for the installation of; and PSW supply for, the ADHRS A/C Unit and'ADHRS ,
; Radiation Monitor Unit.
c , The n_ew PSW loop for the ADHRS Heat Exchangers is in a parallel path with the CCW Heat Exchangers N1P42 B0010. Both the; supply and return. lines'are_12" diameter' pipe. The tie-ins to the existing PSW~are' located in the northwest corner of the Auxiliary
. Building, north of the CCW Heat Exchangers. - From.the. tie-in
points it is routed in a north of then east direction-toward the , Piping Penetration Room adjacent to the RHR "C" Room. It. enters _* the. piping Penetration Room through two penetrations on the' north' wall of this room and continues to the ADHRS Heat Exchangers~ . Inside the room an additional loop is provided to supply PSW to l the ADHRS A/C Unit. On the-return line from the ADHRS Heat I ~ _ Exchanger a supply and= return;1oop is provided for the Radiation
' Monitor.
L This safety evaluation addresses only piping and pipe-support E modifications resulting from the installation of the ADHRS PSW - i l supply. This safety evaluation does not address the operation or
- function of' ADHRS and its impact on the FSAR or Technical .
l Specifications. , REASON FOR CHANGE: This change reflects the addition of Plant Service Water System to the ADHRS Heat Exchangers, the-ADHRS A/C Unit and ' ADHRS Radiation Monitor Unit. 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 evaluated in the Safety Analysis Report. L The piping and pipe supports are designed to meet ASME Section L III/ ANSI B31.1 requirements, as applicable. The designs for ASME L Section III piping are qualified as seismic category I and the l designs for ANSI B31.1 piping are qualified for seismic category The addition of the piping and pipe supports does not II/I. affect the integrity of any other safety system and will not change or degrade or prevent actions as described or assumed in the FSAR. The piping and pipe supports will function in their
- l. intended manner.
L L NLSATTC2/SNLICFLR - 114 l
G-g 1 Attachment-to AECM-89/0093. (+ NPE-89-064-Page: 2-Therefore, there is.no creation of a possibility for an accident or. malfunction of a different type than any. evaluated previously in the-Safety Analysis Report.
'The installation of the piping and pipe supports does not change 7 ..
the limiting conditions for operation applicability or
"" surveillance requirements. The piping and pipe supports meets ASME Section III requirements and will function in their intended manner. Therefore, this change will not affect the margin of safety.
3 v t NLSATTC2/SNLICFLR - 115
. , l Attachment-.to AECM-89/0093 .
t.- o SRASN: NPE-89-065- '000 NO: DCP-88-0008-S06-R00 SYSTEM: E12 y~ DESCRIPTION OF CHANGE: This DCP, in conjunction with the Base DCP > 88/0008 and supplement DCP's 88/0008-1 thru 5 and 88/0008-7, provides the design details for the installation of the Alternate Decay Heat Removal-System (ADHRS). . This-supplement provides the design of the ADHRS pump discharge to heat exchanger piping and pipe supports. This supplement also provides the design details for installation of the ADHRS pump and heat exchangers. Structural modifications are required to the supports on the heat exchangers in order for their installation. No modifications are required to the pump base for their installation. Foundations for the pumps and heat exchangers are provided in DCP 88/0008-4. The ADHRS will utilize two 50% capacity centrifugal pumps and two 50% capacity shell and tube heat exchangers. This equipment will s be obtained from the Unit 2 Fuel Pool Cooling and Cleanup System. Cooling will be provided by Plant Service Water (PSW) and the PSW tie-in will be provided by supplement DCP 88/0008-5. The ADHRS pumps and heat exchangers will be installed in the Piping Penetration Room adjacent to the RHR "C" Room at Elevation 93'-0" of the Auxiliary Building. This safety evaluation addresses only piping, pipe support, heat
.exchangers and pumps installations of ADHRS as described in the above paragraphs. This safety evaluation does not address .the operation or function of ADHRS and its impact on the FSAR.
REASON FOR CHANGE: This change provides for the addition of the ADHRS pumps and heat exchangers and associated piping.
> 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 evaluated in the Safety.
Analysis Report. The piping and pipe supports are designed to meet ASME Section III/ ANSI B31.1 requirements, as applicable. The designs for ASME Section III piping are qualified as seismic category I and the designs for ANSI B31.1 piping are qualified for seismic category II/I. The heat exchanger and pumps are designed in accordance with ASME Section III requirements. Their installations are qualified as seismic category I. The addition of these components does not affect the integrity of any other safety system and will not change, degrade or prevent actions as described or assumed in the FSAR. The components will function in their intended manner. NLSATTC2/SNLICFLR - 116
cm- , 1c: ,e .,._ , Attachment to AECM-89/0093
.. gg.
. -)
NPE-89-065 ' Page 2l
- s 9-I .Therefore,-ther$isnocreationofapossibilitsforanaccident.
or malfunction of a different type than any evaluated previously in'the Safety Analysis Report. The installation of the piping, pipe supports, pumps, and heat' exchangers does not change the limiting conditions for operation applicability or surveillance requirements. The piping.. pipe supports, pumps, and heat exchangers meets ASME Section III ,
. requirements and will function in their intended manner.
E
~Therefore,.this change will not affect the margin of safety.
l I s i NLSATTC2/SNLICFLR - 117
p ' l
- I
. Attachment to AECM-89/0093 1
SRASN:- NPE-89-066 -DOC NO: DCP-88-0008-507-R00- SYSTEM: ~E12 DESCRIPTION OF. CHANGE: This DCP supplement makes various instrumentation and electrical modifications required for the addition-of the: Alternate Decay Heat Removal System. This includes the following: a)' Pressure tap points 1E12-PPN049A, N049B, N0410A, N4108, N411A, N4118 and N412 will be added to the RHR system to support temporary gauges used in system performance testing. b) Two temperature loops and associated indicators (1E12-TIR631 and R632) will be added to provide ADHRS heat exchanger inlet and outlet temperature monitoring from the control room. . c) A pressure switch IE.12-PSL-N030 will be added to provide low suction pressure protection for the ADHRS pumps. d) All electronics required for the heat exchanger effluent +
; radiation monitoring of the PSW system will.be added.
, This includes installation of a radiation detector (ID17 RE-N015), a skid. mounted flow switch (1D17-FSI-N073), an annunciator card (ID17-RITS-K615), a high alarm curcuit
.(ID17-RAL-L656) and a flow alarm circuit (1017-FAHL-L657),
e) Power will provided for the process sample pump and flow 4 indicator mounted on radiation monitoring skid 1017-J015, as well as cabling -for control room monitoring / annunciation and BOP computer point retransmission, f)- Cabling will be installed for ADHRS heat exchanger inlet and outlet temperature monitoring. g) Power feeds and control circuitry will be added for the ADHRS pumps (Q1E120005A-N & B-N) and the associated pressure switch (E12-PSL-N030). i h) Addition of power feeds and control circuitry for RHR flow control valve Q1E12F424-N. 4 i) Addition of power feeds and control circuitry for RHR Fuel Pool Cooling Assist Suction Valves Q1E12F066A-A & B-B. j) Addition of power feeds and control circuitry for the RHR C Pump Room Aid Handling unit N1T418014-N. k) Addition of control circuitry for the interlock between the RHR C Pump Q1E120002C and associated suction valve Q1E12F004C.
- 1) Addition of Control circuitry for the ADHRS mode trip enable.
m) Addition of all associated ADHRS conduit and tubing supports. NLSATTC2/SNLICFLR - 118
/p -
Attachmont to AECM-89/0093 L NPE-89-066 Page 2 m
'OCP 88/0008-7 and this evaluation are thus limited to the r _ activities as outlined above, o
REASON FOR CHANGE: This change will reflect the addition of the Alternate Decay Heat Removal System's associated instrumentation and electrical modifications. SAFETY EVALUATION: There is no increase in the probability of occurrence or in the consequences of and accident or malfunction of equipment important to safety previously evaluated in the. Safety Analysis Report. Safety related and non-safety related circuits have been provided in accordance with Regulatory Guide 1.75 to insure that failures or events that occur in non-safety related systems are not ! propagated'into safety related systems in an manner that has not been previously evaluated. For non-safety related systems, the worst case failure / event would be a loss of offsite power, and-this has been previoulsy evaluated both in UFSAR analyses and the 50.59 evaluation for DCP 88/0008, Rev.0. The only safety related circuits affected by the ADHRS are MOV circuits for */alves Q1E12-F066A-A and F066B-8, and the protective interlock between the RHR C suction valve and RHR Pump C. These circuits have been ! previously evaluated in the 50.59 evaluation for DCP 88/0008, Rev. ' 0, and found to be acceptable. All ADHRS circuits, controls, power supplies,-and instrumentation meet the requirements for physical and electrical separation (Reg. Guide 1.75) and equipment qualification where applicable (IEEE-323). All raceway design provided has been evaluated to seismic or II/I criteria and found to be. acceptable. Circuit protective devices are adequately sized / coordinated to provide l their intended function without creating the potential for
^
spurious trips. Control room controls and instrumentation have been provided that are consistent with present design philosophy as related to operation of plant equipment from a human factors standpoint. Accident consquences which consider overall installation of the ADHRS have been evaluated by the 50.59 evaluation for DCP 88/0008, Rev. O. It has been determined that l the referenced evaluation envelopes assessment for accident L consequences for the scope of activities as outlined in the description for this evaluation (88/0008-7), and the results of the evaluation were found to be acceptable. l-NLSATTC2/SNLICFLR - 119
9' ; ~ Attachment to AECM-89/0093; 'x NPE-89-066 Page 3 Non-safety related circuits have been provided with necessary isolation, separation, and seismically evaluated supports 1
' sufficient to insure that failure /maloperation of these circuits !
does not propagate into the class IE power system. Class IE-circuits are provided with appropriately qualified materials to insure proper function under the environmental conditions to which-they will be subjected. Class IE circuits are isolated, ! separated, and supported to prevent maloperation.in one division from propagating into the other divisions. . Equipment added or . modified by the overall ADHRS design has been evaluated for ! malfunctions of equipment important to safety by the ADHRS ! interaction evaluation as addressed by the 50.59 evaluation for 1
.DCP'88/0008,- Rev. O. The results of this evaluation have been i determined to envelop modifications performed by this supplement. !
Per the referenced evaluation and the above discussion, it has been determined that there will be no increase in the probability of a malfunction of equipment important to safety as previously evaluated. Installation of safety related ADHRS circuits utilizing IEEE-379 single failure criterion insure that malfunctions of equipment- ! important to safety are bounded by existing analyses for loss of a j single ESF division (i.e. failure of a diesel generator). Use of
~
dual contact' logic in the RHRC pump interlock added by this design
'will further reduce the possibility if loss of an ECCS loop due to-a single contact failure. Consequences of a malfunction of equipment important to safety due'to the addition of the overall '
ADHRs design are evaluated by the ADHRS interaction evaluation to l _ support the 50.59 evaluation-for DCP 88/0008, Rev. O. Reveiw of this evaluation has determined that changes as provided by DCP
-88/0008-7 are enveloped by the referenced evaluation and that ,
m there will be no increase in the consequences of a malfunction of l equipment important to safety as a result of this change. , There is no creatien of a possibility for an accident or l malfunction of a different type than any evaluated previously in l: the Safety Analysis Report. Use of Reg. Guide 1.75 separation, IEEE-379 single failure criterion, IEEE-323 equipment qualification where required, and ! seismically evalusted raceway design for the installation of electrical circuits, raceway, and components for the ADHRS will insure that new accident possibilities will be bounded by pervious t analyses. The possibility for creation of new accident types by the installation of the overall ADHRS design has been evaluated by the 50.59 evaluation for DCP 88/0008, Rev. O as provided by the ADHRS interaction evaluation. This DCP supplement (88/0008-7) has been reveiwed and it has been determined that activities conducted by DCP 88/0008-7 are enveloped by the referenced evaluation. Therefore, no new accident possibilities will be created by performance of the activities as described. NLSATTC2/SNLICFLR - 120
rn; , Attachment to AECM-89/0093 v NPE-89-066
-Page 4
- s. -
Interfaces with equipment'important to safety as previously-evaluated and ADHRS equipment have been reveiwed for the potential L of creating new malfunction possibilities. For the non-safety, I -related circuits added/ modified by ADHRS, physical and electrical separation provided by Reg. Guide 1.75 has been utilized to insure that non-safety related circuits will not degrade the operation of
, equipment important to safety. In.the case of added/ modified safety related circuits, potential malfunction have been evaluated by the 50.59 evaluation for DCP- 88/0008,-Rev. 0, and found to be acceptable. Overall-ADHRS equipment design was evaluated for new-malfunction possibilities throughout a'11 reactor modes as addressed by the ADHRS interaction evaluation and the results have been determined to_ envelop equipment added/ modified by this supplement.
Technical. Specification safety margins will not be reduced due to the scope of activities conducted under this evaluation..
- Electrical power system independence, capacity, and redundancy will be retained consistent with existing design. Electrical protective functions are provided where required to insure proper system operation for affected systems due to changes in system operation due to the ADHRS (e.g. the added RHRC pump to l A suppression pool suction. valve interlock). Also, ADHRS major . 'j components, will be deenergized during reactor modes 1, 2,.and 3 to further reduce the possibility for inadvertant use/maloperation of such equipment. .For an overall ADHRS system evaluation, the 50.59 evaluation for DCP 88/0008, Rev. O is bounding and the
, results were found to be acceptable. l , NLSATTC2/SNLICFLR - 121
pps , p,- 1 ['
-Attacht4ntLto'AECM-89/00931 y
Y- > f ,
~SRASN:' NPEi B9-067 DOC NO: DCP-88-0009-S00-R00 SYSTEM: Fil l
EDESCRIPTION OF CHANGE: The purpose of DCP 88/009 is to enhance the operational" reliability of the Horizontal Fuel Transfer System (HFTS) in order to support the refueling outage schedule.. The- '
. primary modification is to replace the existing cable / drum drive
' system with a pull-pull cable drive arrangement. This N '
modification'will include replacement of the winch with a dual-3
; winch. assembly,-' structural modifications;to the Auxiliary Building f-HPTS structure as required to mount-the new sheave assemblies, and-installation of:a horizontal sheave assembly in the transfer tube near the tube closure hatch. Secondary modifications will include f replacing the underwater:upender limit switches with proximity L switches, replacing the underwater carriage-position limit
[ switches with above water geared-limit switches (at the winch assembly), making HPU changes to speed up the Upender hydraulic system, "hardpiping" the underwater hydraulic lines, and. routing the underwater electrical cables in stainless steel tubing. The HFTS (F11-E015) is non-safety related, Seismic Category II/I, except for the transfer tube, which is safety-related,-Seismic-Category I since it is part of the primary containment boundary. All cable and raceway installations will be made in accordance
, with the separation requirements of Reg. Guide 1.75. The HFTS- ,
(F11-E015) is presently powered by a 480V circuit from MCC 12B52.
.This power' feed will remain the-sole power source-for the system, and the feeder circuit breaker is appropriately sized to protect ,.
[ the feeder cable. Modifications performed by this DCP will not i affect the Fire Hazards Analysis or any 10CFR50-Appendix R analyses. REASON FOR CHANGE: This change to the Horizontal Fuel Transfer System is to support the refueling outage schedule. 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 evaluated in the Safety ! Analysis Report. .l
-f None of the existing'UFSAR accident evaluations, including fuel .;
drop accidents in 15.7.4 and 15.7.6, have a probability of occurrence based on the reliability of the affected HFTS components or structures, including the transfer tube. In addition,.no system, structure or component identified in any previously evaluated UFSAR accident is dependent upon the affected q HFTS equipment for mitigating the probability of the postulated j accident. Cable and raceway installations will be made in ' accordance with Reg. Guide 1.75 to insure no safety related components will be affected. NLSATTC2/SNLICFLR - 122
*' 'Attachmsnt to AECM-89/0093 i
/
NPE-89-067 Page 2 , No' 0FSAR accident was evaluated for consequences of an accident caused by failure of the affected HFTS components or structures. Also,'the HFTS is not required to mitigate the consequences of any previously evaluated UFSAR accident, and changes to the transfer tube will.not adversely affect the structural or pressure retaining integrity of the primary containment boundary. None of the affected HFTS equipment is safety related except'the ' transfer tube which serves a primary containment function. All ' applicable HFTS equipment is designed to preclude seismic II/I concerns in order to ensure integrity of the transported fuel and protect the transfer canal liner plates from damage caused by structural failure.of the applicable support structures. Primary containment is not adversely affected since the changes do not affect the closure hatch and the modifications inside the tube 4- will comply with the original ASME design criteria and seismic / pressure retaining integrity is maintained. Except for the transfer tube, the affected HFTS equipment is not safety related and failure would not affect any safety related equipment required to prevent or mitigate the consequences of any previously evaluated UFSAR accident. The transfer cart and rail L support structures will retain their seismic adequacy to prevent damage to the spent fuel. Thus, damage to more than 101 fuel pins L as described in UFSAR 15.7.4 and 15.7.6 is not postulated. . Also, L the consequences of equipment malfunctions that rely on primary E containment for mitigation of resulting accidents are not r increased'since the transfer tube will maintain its structural and pressure retainW integrity. The HFTS equipment is only required during Mode 4 or 5 - refueling, except that the primary containment function of the transfer tube is required during Modes 1, 2, and 3. All HFTS L equipment affected by this DCP shall provide the same operational
- function as the existing equipment in accordance with the original design intent. All additional cabling and raceway installations will be made in accordance with Reg. Guide 1.75 to insure electrical . isolation and physical separation from all safety related components. The modifications are intended to provide increased operational reliability in accordance with the original design criteria, including the ASME criteria for the transfer tube changes. Also, this DCP does not affect the transfer tube closure hatch. Thus, primary containment is not adversely affected for Mode 1, 2, or 3. In addition, none of the affected HFTS equipment is required for RHR or spent fuel cooling capabilities during Mode 4 or 5. Therefore, the possibility of a new accident scenario is not postulated.
NLSATTC2/SNLICFLR - 123
T, ; Attachment =to AECM-89/0093 NPE-89-067 [ Page 3 : n Changes to the transfer tube will maintain the adequacy of the primary containment boundary it serves. All other HFTS equipment changes provide enhanced operational reliability while complying-with the same design criteria and performing.the same function as the existing equipment. Also, these-changes do not affect. safety
-related equipment, nor will operational failure of the subject n HFTS equipment affect any other equipment that can Jeopardize RHR or spent fuel cooling capabilities during Mode 4 or 5. Structural
~ failure is prevented by the Seismic II/I design to ensure that 1) fuel integrity is maintained in Mode 4 or 5, 2) safety related C equipment is not jeopardized in any mode, and 3) transfer canal liner plate-integrity is maintained in all modes. The basis for Technical Specification'3/4.9.12 is to control personnel access to the potentially high radiation areas (Room 1A525 . the transfer tube access area, Auxiliary Building Elevation 185'-0") immediately adjacent to the tube and to assure safe operation of the HFTS. The HFTS changes do not affect personnel access control to Room 1A525. Also, the HFTS changes are intended to enhance the operational reliability of the system while maintaining the same logic and. interlock features that ensure safe operation. Thus, the basis for Technical Specificatior, 3/4.9.12 is not affected. In addition, since transfer tube changes will not adversely affect primary containment, the margin _of safety is not reduced for any Technical Specification. NLSATTC2/SNLICFLR - 124
, l . Attachment to'AECM-89/0'093 P
SRASN: NPE-89-068 DOC NO: DCP-88-0012-S00-R00 SYSTEM: N19 DESCRIPTION OF CHANGE: The condenser waterboxes are provided with access'manways_to permit' entry to perform inspections which are required.in VFSAR Section 10.4.1.4. The access manways are currently _ sealed using a cover designed by-Southwestern-Engineering Company. Sealing the manway cover is extremely difficult with the current design. The current gasket joint design is such that internal-pressure tends to move the gasket. , The narrow sealing surface of the gasket against the manway pipe . causes' the gasket to roll out 'of the gasket groove on the cover; To alleviate this sealing problem, a new manway cover will be designed.using standard slip on, small tongue and groove flanges and rubber or compressed asbestos gaskets. The Standard ANSI B16.5 flange will provide a better bolting arrangement to allow a more positive method of obtaining seating stress-on the gasket. REASON FOR CHANGE: This change allows for a more positive method of obtaining seating stress on the gasket, thereby lessening the chance for leakage. 3 SAFETY EVALVATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. Leaking waterbox manway covers would result in the loss of circulating water into the Turbine Building. Section 10.4.5.3 of j the UFSAR evaluated the effects of a leaking expansion joint which j would envelope the case of a leaking manway. Calculation '
'CC-NIN19-88042, Rev. I designed the new manway covers and showed the structural adequacy of both the manway covers and the condenser waterbox. The new design provides a small tongue and groove flange which will provide a' uniform and stable seating surface for the new gasket. The twenty bolt flange will also ,
provide a better means of applying the gasket seating stress which-will decrease the chances of leaking. Therefore, the probability 4 of an accident previously evaluated in the FSAR will not be increased. , UFSAR Section 10.4.5.3 evaluated several different scenarios of losing circulating water from the circulating water system. One ; scenario, postulated a-gross failure resulting in the flooding of i both Unit I and 2 Turbine Buildings and the Control Building to an elevation of 93' - 8 1/8". This scenario would envelope any possible scenario resulting in the leaking or even failure of a manway cover. Additionally, Calculation CC-NIN19-88042, Rev. 1, NLSATTC2/SNLICFLR - 125
I - Attachment to AECM-89/0093 NPE-89-068-c Page 2 showed the structural adequacy of the condenser waterbox with the
.new manway cover and the manway cover itself. Therefore, the consequences of an accident previously evaluated ~in the FSAR will, i not be increased.
UFSAR Section 10.4.5.3 evaluated the possibilities of gross failures in the circulating-water system, resulting in the g flooding of the Unit 1 and 2 Turbine Buildings, the Control Building, the Radwaste Tunnel, and the Radwaste Building. . _The-evaluations showed that no equipment essential;to attaining-and ., maintaining a cold safe shutdown would be affacted by the _! resulting flooding. Any possible leaking from the manway covers would be bounded by the evaluations performed. Therefore, the. probability of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. The loss of circulating water from the condenser waterbox with a ; failure of the manway cover would result in flooding which would be bounded by the worst case of gross failure floodinq. , Therefore, the consequences of a malfunction of equipment !
.important to safety previously evaluated in-the FSAR will not be .
increased. [ The purpose of the waterbox manway is to provide a means of access i to inspect the interior of the waterbox and the condenser tubes. The manway cover's function is to prevent the release of the ' circulating water from the waterbox into the Turbine Building. So the failure of the manway cover would only result in the loss of : the circulating water, which is evaluated in the UFSAR and the new 1 manway covers and waterbox have been shown to be structurally adequate in Calculation CC-N1N19-88042, Rev. 1. Therefore, the installation o' the new manway covers will not create the , possibility of an accident or a malfunction of equipment important > to safety of a different type than already evaluated in the FSAR. The manway covers are not used as a basis for the GGNS Technical Specifications and have no need of being used as basis in any GGNS Technical Specifications. The new design of the manway cover is being provided to prevent the chance of leakage around the manways which has previously occurred. This design provides a better gasket, a more uniform gasket seat, and a more secure method of bolting than the original design. Therefore, the new manway covers will not reduce any margin of safety as defined in the basis for any GGNS Technical Specification. NLSATTC2/SNLICFLR - 126
Attachment to ALCM-89/0093
'SRASN: NPE-89-069 DOC N0: DCP-88-0036-S00-R00- . SYSTEM: P41 DESCRIPTION OF CHANGE: GGNS committed to flow balance alternating l divisions of SSW during future outages. Presently, annubars and- ;
, DP cells are required to be temporarily installed and removed i during every outage in the two divisions for flow balancing. ,
=Three of the most critical locations, in which permanent annubars l are required, are the Division I Jacketwater Cooler (P41-N063A), M Division II Jacketwater Cooler (P41-N063B) and the Fuel Pool Cooling Heat Exchanger "B" (P41-N082). +
r. In order to minimize impact on the system, this DCP will provide three new flow elements as substitute for the flow points mentioned above. These new flow elements will be provided with a safety related 2" gate valve and safety related piping to allow "Q" annubars to be installed at the ends. No DP cells'and associated hardware will be provided in this DCP. Operation of these flow points will require the_ annubars to be in the withdrawn position (root valve closed) during times when flow balancing is ; not being performed. This is necessary since leaving the annubars in the inserted position would lead to fouling of the annubars. REASON FOR CHANGE: Permanent installation of annubars are needed to reduce outage critical path item delays caused by the installation of temporary instrumentation used for flow balancing during an outage and to reduce time required for weekly and i quarterly flow-test that are performed on components using SSW f or-cooling water. SAFETY EVALUATION: There is no increase in the probability of . occurrence or in the consequences of an accident or malfunction of I equipment important to safety previously evaluated in the Safety Analysis Report. As stated in UFSAR Section 9.2, the safety function of the SSW ' system, containing the plant ultimate heat sink (VHS), is to provide a reliable source of cooling for plant auxiliaries that are essential to a. safe reactor shutdown. The SSW system is designed to perform this cooling function following a design basis loss of coolant accident (LOCA) automatically and without operator action, assuming a single active failure coincident with a loss of offsite power. The annubars and associated piping have been designed to ASME Section III, Class 3 requirements and are qualified as seismic Category I. Installation of the annubars will not adversely affect the UFSAR Sections 3.6 and 3C.3 analyses for piping f ailures in a fluid system. For the annubar in the Auxiliary Building, the existing analyses have demonstrated that the existing hazardous effects NLSATTC2/SNLICFLR - 127
m - V~
' Attachment to AECM-89/0093 l
L:. V [ NPE-89-069 Page 2
; from moderate energy pipe cracks (i.e., area flooding, low
' velocity wetting of equipment in the-area, and II/I concerns) will
-encompass the effects of any additional postulated leakage. For the annubars located in the D/G Building, UFSAR Section 30.3.4 ,
states that no pipe cracks have been postulated due to flooding . concerns in these rooms. Therefore, the annubar installations ' have been designed per UFSAR Section 3.6.A.2.1.c.4(d) to stress . levels-in which no. cracks are required to be postulated.
.The operation or function of the affected system as analyzed in-the FSAR is not affected by the addition of the annubars. The SSW system original design as described in the UFSAR has not changed.
Therefore, there is no creation of a possibility for an-accident i or malfunction of a different type than any evaluated previously ; in the Safety-Analysis Report. The addition of the annubars will not change the function or operation as defined by the bases of the Technical Specifications, therefore,-the margin of safety is not reduced.
~
l l l l NLSATTC2/SNLICFLR - 128 l
g
~'
Attachment'to AECM 89/0093 SRASN: NPE-89-070 DOC NO: MCP-88-1027-S00-R00 SYSTEM: M61
}
s DESCRIPTION OF CHANGE: This minor change package provides. ! installation instructions.for two modifications to the Integrated Leak Rate: Test (ILRT) pressurization skid, a mobile test platform used in-the Drywell Bypass Test discussed in Technical-
-Specification 4.6.2.2. These modifications are: a-2-inch manual valve in parallel-with the present 6-inch air control' valve in order to better control air flow rate, and a 3-inch bypass'line with a manual- valve between the air compressor connection and the exhaust silencer in order to provide a way to control delivered air pressure independently from flow rate. These changes will be-depicted in the piping diagram for the ILRT skid, which is included in the P&ID for the Containment Leak Rate Test-System (Drawing M-1111A, FSAR Figure 6.2-76). ,
REASON FOR CHANGE: These changes were made to ensure better control of air flow rate and to control delivered air pressure-independently from flow rate. 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 evaluated in the Safety Analysis Report. 1 The piping and pipe support designs meet ANSI B31.1 code requirements. The installation is non-safety related and will not 4 affect the safety related portions of this system. The operation or function of the affected system as analyzed in the FSAR is not affected by the addition of this piping. The affected system in this MCP is non-safety related except where the system' penetrates containment. The failure of the affected system will not compromise any safety related system or component and will not prevent reactor shutdown. The addition of the bypass loops made by this MCP do not affect the-analysis of the system as j described in the FSAR. The changes do not modify any equipment used in mitigating the
. consequences of an accident as analyzed in the FSAR, nor do they affect any actions taken to mitigate an accident as analyzed in the FSAR. The modifications will not affect the operation of the M61 system as analyzed in the FSAR.
L Therefore, there is no creation of a possibility for an accident L or malfunction of a different type than any evaluated previously l in the Safety Analysis Report. p l Since the operators will be able to more closely control the test l of the containment and drywell in accordance with the ! requirements, then this test will be a more reliable measure of actual system performance, thus ensuring that the margin of safety for the referenced Technical Specification is maintained. No other specifications are affected. NLSATTC2/SNLICFLR - 129 l i
,- ~ , , . - , ~ . . , .
e ~ , P Attachment to'AECM-89/0093' , + , i 1 -SRASN: NPE-89-071 DOC NO: DCP-88-4500-500-R00 SYSTEM: R14 , DISCUSSION OF. CHANGE:: The referenced MNCR was generated to identify the main generator 'B Phase' step-up transformer < (N1R145001B) as-having failed during startup after RF02. -The subject DCP.is being issued to make electrical and fire protection changes necessary to facilitate-replacement of the failed transformer with a similar Unit 2 transformer. The replacement
- transformer has equivalent electrical ratings and protective
- features,_and.is designed for'the service environment to which it will_be subjected, The DCP provides instructions to modify- :
vari _ous electrical connections to account for differences in-physical dimensions between' the original unit'and the replacement transformer. DCP 88/4500, Rev. O also addresses modifications and: , subsequent inspection and testing requirements for the. deluge water-spray system-(N1P64D110B). This system is designed to extinguish a fire involving the 'B Phase' step-up transformer. ;; The' deluge system is designed and' installed in accordance with
~
NFPA 15. _ Additionally, this system is non-safety related and is not required to be seismically designed. Per NFPA 15', the deluge system must provide an impingement rate of 0.5 gpm per square foot
- of projected surface area of the transformer tank and coolers. .
REASON FOR CHANGE: These changes were made to make electrical and fire protection changes necessary to facilitate replacement of the subject transformer. 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 evaluated in the' Safety
-Analysis Report.
Per UFSAR Section 9.5.1.-2.2.4, the deluge water spray system must L be designed and installed per NFPA 15. The deluge system will not 7 be affected by implementation of this'DCP. Only minor piping and l nozzle redirection changes are being made to accommodate the l- replacement transformer. The changes implemented per this DCP i will not result in a system hydraulic demand in excess of the L existing fire water supply capability. This system provides l protection to the 'B Phase' step-up transformer which is non-safety related and is not required for safe shutdown. The replacement transformer has equivalent electrical ratings and protective features, and is designed for the service environment to which it will be subjected. Differences in physical dimensions between the original unit and the replacement have been incorporated in the DCP. The deluge system piping changes implemented serve to accommodate the replacement transformer. The modified system will continue to provide 0.5 gpm per square foot of projected surface area of transformer tank and coolers as required by NFPA 15. l System detection and actuation logic is unaffected by implementation of the subject DCP. NLSATTC2/SNLICFLR - 130
3g , t ,
'N .AttO;hment't3 AECMiB9/0093-
- 3. I
?Y .
E tt@J NPE-89-071- :!
- page 2 .
Additionally, this deluge system can be isolated by. Valve I 4 NSP64F081B ylthout affecting any other lire protection / suppression I l systems. Tnis modified system in combination with'various two ! Lhour fire barriers will protect buildings containing' safety !
-Velated equipnent from exposure or spill fires involving the .. step-up. transformer-thereby satisfying the requirements in Section l;jp D.I.h of Table 9.5-11 of the UFSAR.
l 1 Ther6 fore, there is ru) creation of a possibility for an accident- r or malfunct-lon of a.different type than any evaluated previously-in the Safety Analysis Report. .! Technical Specification'3/4.7.6.2 governs spray and/or sprinkler. j systems at GGHS.. More specifically, this Technical Specification i governs-systems which provide fire protection to components ' essential for safe shutdown. .The deluge spray system (N1P64D110B)
-is-located in Fire Area 59 (i.e.,' site yard) and provides fire- 1 protection to the 'B Phase' transformer. Both the deluge system !
and the transformer are non-bafety related and neither is required for safe shutdown. Therefore, the margin of safety as defined in l the basis of any Technical Specification is not reduced by t .L implementation of DCP BB/4500. l i k h i k M, 7 t (.. l f _; 1 m I i tu r NLSATTC2/SHLICFLR .131 w _ , _-...- . -- . _. _
f I Attachment to AECM-89/0093 SRASN: NPE-89-072 000 NO: MCP-88-1052-500-R00 SYSTEM: C61 DESCRIPTION OF CH/dGE: At present, both seppression pool level indicators at the remote shutdown panels are not consistent with the control room recorders and do not have sufficient range to allow completion of all Emergency Procedure (EP) steps as intended by the BWR Owners' Group EP Guidelines (EPGs). This was identified during the Detailed Control Room Design Review (DCRDR)/EP Upgrade efforts. Presently, the meters have a range of 12.33' to 24.33', which does not match the control room recorder range of 10,5' to 25.5'. SERI committed to change the indicator ranges to match the control room and allow completion of intended
, EP steps.
This MCP expands the range of C61-LI-R402A on H22-P150 and C61-LI-R402B on H22-P151 to 10,5' to 25.5' to allow completion of intended EP steps and conform to good human factors practices. This will require replacement of the present 12.33'-to 24.33' scales on indicators C61-LI-R402A(B) with new 10.5' to 25.5' scales and recalibration of their associated transmitters C61-N402A(B). REASON FOR CHANGE: To enable the Operations staff to complete all the EP steps _as intended by the BWR Owners' Group, and to conform to good human factors practices. This change will provide more consistent information it, the control room and at the remote shutdown panels. SAFETY FVALVATION: This is an indication range change only. No equipment function is changed, and no new equipment or component is added. No seismic, fire protection, or equipment qualification criterion is affected by this change. Therefore, there is no increase in the probability or c:,nsequences of an accident or malfunction of equipment important to safety. Nor is there a possibility of an accidtnt or malfet.ction of equipment important to safety different from any previously evaluated. This design change providos the operator with additional information at the remote shutdqwn penels which is consistent with indication in the control room. It is intended to reduce the probability of operator error by providing more information and consistency between the control room and the remote shutdown panels. Therefore, there is no reduction in the margin of safety as defined in the basis for any Technical Specification, i l NLSATTC2/SNLICFLR - 132 1 l
?'
g: Attcchment t> AECM-89/0093' i . E' SRASN: NPE-89-073 DOC NO: MCP-89-1010-S00-R00 SYSTEM: P41 1' f '
. DISCUSSION OF CHANGE: The purpose of this MCP is twofold:
[ provide a removable spool piece in the SSW makeup water line to Basan "B" and install the injection line in Basin "B" for the ; future SSW chemical injection system (DCP B6/0092). [ c EER 88/6066 requested a removable spool be provided in the busin p makeup water supply line in order.to support the chemical cleaning effort during RF03. [ The SSW makeup supply line, 8" JBD-174, is modified by the f g . installation of 2 pa$r of flanges just downstream of Valve . NSP41F504B. Also, line 3/4" JBD-1205 and support NSP41G014C01 l will require minor design changes to allow installation of the 1 flanges. ! The installation of the' injection lint, JZD-40, for the future i 44 = chemical injection syctem, is a 2" diameter pipe made of carpenter i f 20 alloy. It originates outside of the pump house on the north ; side .- This outside portion comicts of a blind flange and a plug ! valve. The lane enters and exits the pump house through two new. ; penetrations. It descends to Elevation 76' passing through the debris screen to a point between the SSW pump Q1P41C001A-A and the .
- HPCS SW pump Q1P41C002-C. It is located and supported as to '
IL preclude any possible failure that could affect the operation of the SSW system. l: ' REASON FOR CHANGE: These changes support the chemical cleaning of f the SSW basin and provide a futura chemical injection line. ! [ SAFETY EVALUAT10'h There is no increase ~in the probability of ! occurrence or in the consequences of an accident or malfunction of [ equipment important to safety.previously evaluated in the Safety ; !- Analysis Report. . c' As stated in UFSAR Section 9.2, the safety function of:the SSW i System, containing the plant ultimate heat sink (UHS),-is to provide a reliable source of cooling for plant auxiliaries that are Ossential to a safe reactor autdown. The SSW system is designed to perform this cooling function following a design basis loss of coolant accident (LOCA) automatically and without operator ; action, assuming a single active failure coincident with a loss of , offsite power. ! The operation or function of the SSW system, as analyzed in the f FSAR, is not affected by the described piping system ' modifications. The piping and pipe supports installed by this MCP have been designed to ANSI B31.1 requirements and are qualified as ; seismic category II/I. The SSW system original design as . described in the UFSAR has not changed as a result of the l installation of the described flanges or the injection line. ; L i l NLSATTC2/SNLICFLR - 133 , u < i
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e ; Attachment to AECM-89/0093 i l NPE-89-073 Page 2 i
- The addition of the flanges to the $$W makeup line will not change or affect its function. The design of the chemical injection
, line's discharge sparger, which will be located in the $$W basin sump, is consistent with the design of the debris screen over the sump with respect to preventing particles greater than 1/8" t diameter from entering the SSW pump suction. Also, the discharge
- sparger is located and supported as to preclude any possible F failure that could affect the operation of the SSW system. The 1 SSW system operational reliability per the original design as described in the UFSAR has not changed. Plant operation with this i
piping in the SSW system will have no adverse effect on the functionality of systems required to mitigate the consequences of postulated accidents or malfunction of equipment important to safety. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The addition of the described piping will not change the function or operation as defined by the bases of the Technical Specifications, therefore, the margin of safety is not reduced. 6 9 a NLSATTC2/SNLICFLR - 134
r~ Attachment to AECM-89/0093 SRASN: NPE-89-074 DOC NO: MCP-89-1045-$00 ,00 SYSTEM: G33 : DESCRIPTION OF CHANGE: The objective of this MCP will be to add a . whip restraint to line 4" DBZ-22 at a point just upstream of the flange connection to the RWCU Heat Exchanger NIG33B0001A. 4 REASON FOR CHANGE: The addition of a whip restraint will limit the pipe movements and maintain the stresses in the piping at an ' acceptable level which will assure the operability of valve F251 ' [ following a postulated terminal end break at the Heat Exchanger G33 B001A. SAFETY EVALVATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the Safety Analysis Report, The addition of the whip restraint will not - affect the operation or function of the G33 system. The whip restraint installed by this MCP meets all applicable design " requirements, Installation of the whip restraint is in conformance with the methodology utilized for evaluatiun of - postulated pipe break effects for Grand Gulf Unit 1. The addition of the whip restraint will provide the necessary hardware modifications to assure operability of valve F251 following a postulated terminal end break at the Heat Exchanger G33 B001A. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report, The whip restraint is not addressed in the Technical Specifications, and therefore, has not been utilized in computing a margin of safety. This MCP will not affect the operation or function of the G33 system, and therefore, the margin of safety is not reduced. 1 , 6 NLSATTC2/SNLICFLR - 135
T Attachment to AECM-89/0093 SRASN: NPE-89-075 000 NO: MCP-89-1001-S00 900 SYSTEM: FIS DESCRIPTION OF CHANGE: The Refueling Platform is operated manually by an operator requiring expert judgement and skill in positioning the fuel platform grapple over the fuel assembly and in moving the platform to and from the fuel transfer locations. This modification provides an operational enhancement positioning system (Lasertrac) to be " plugged-in" to the refueling platform.
, The Las9ttrac system allows the operator to input a move or c file L move cornmand to any pre-defined fuel location and the system will move the bridge and trolley simultaneously at the maximum allowed speed to that location. The file move command allows a set of moves to be entered from a file and executed at the command of the ! operator.
This modification does not affect the existing no mal manual operation of the platform system or restrict the operator in making any platform moves during operations. This modification does not bypass any of the existing system interlocks. REASON FOR CHANGE: Lasertrac when " plugged-in" will automatically position the platform and fuel grapple over the fuel assembly and in moving the platform to and from the fuel transfer locations. 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 evaluated in the Safety Analysis Report. None of the existing interlock logic is invalidated or affected. When Lasertrac is " plugged-in" all existing interiocks remain intact and effective. Implementing this modification does not change, degrade or prevent actions described in the " sequence of events" Table 15A.6-36 of the UFSAR. The refueling platform operational parameters are not affected by this modification and therefore safe shutdown capability as required by 10CFR50. Anti-collision reliability is increased when Lasertrac is " plugged-in" and Lasertrac is in the semi-auto or auto modes of operation. The design function of the refueling platform system remains unchanged by this modification. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The design function of this modification to the platform system retains all existing control and interlocks and the margin of safety is not affected. i NLSATTC2/SNLICFLR - 136 L
e.. Attcchment t3 AECM-89/0093
- SRASHi -NPE-89-076 DOC NO: DCP-84-4013-S00-R01 SYSTEM: T4B ,
a,
, DESCRIPTION OF CHANGE: The purpose of this package is to install- < metal shroude over the atmospheric pressure sensing process ports :
It ' of the Pressure Differential Transititters (PDT). The subject r
. shrouds will be located at approximtely the 280' Elevation of the !
Enclosure Building.D Two will be located'on the North Wall and two + on the South Wall. l L I REASON FOR CHANGE: This change prevents wind or rain from damaging the transmitters, j l- SAFETY EVALUATION: There is no increase in the probabilitv of occurrence or in the consequences of an accident or malfunction of , F equipment'important to safety previously evaluated in the Safety Analysis Report. The covers were added to the siding of the > Enclosure Building to protect the PDTs from wind and rain.- f- Although the siding to which the covers will be attached is part ! s of secondary containment; the covers will not have adverse effect on the Enclosure Building due to light. loads. In addition, the : shrouds have been designed to meet all material and construction ! requirements applicable to light gage cold formed steel. The i l' reliability the Enclosure Building ~is not. degraded because the. ; appropriate seismic 11/1 requirements have been met. ; o Although the change is associated with a safety related system and - structure used in mitigating the consequences of an accident, the ; enclosures have been designed so that the Enclosure Building will. ' > continue to maintain the integrity of secondary containment. The ; l- change will also not prevent the Standby Gas Treatment System from- . functioning as intended. Therefore, there is no creation of a ; 5 possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. - The enclosures have been designed such that they will not adversely affect the ability of the Enclosure building to maintain secondary containment. Therefore the change is in keeping with the basis of GGNS Technical Specification 3/4.6.6. p HLSATTC2/SNLICFLR - 137
Attach;ent to AECM-89/0093 I SRASN: NPE-89-077 DOC NO: DCP-88-0018-500-R00 SYSTEM: U22 DESCRIPTION OF CHANGE: The originally planned floor at Elevation 166'-0" of the Turbine Building between Column Lines C & E and 18 and 19.8 will not be installed, resulting in a larger equipment hatch. The existing equipment hatch was intentionally left unfinished around its perimeter during construction to permit greater construction access. Consequently, slab reinforcing had been left exposed and a temporary wood platform erected around the g perimeter of the hatch to cover the unfinished slab and exposed ; reinforcing. Also, a temporary wire barrier. exists in lieu of a ' standard handrail. The subject DCP has been initiated to finish the concrete slab around the railroad equipment hatch (south and ' s west ends) and to provide a handrail around the hatch. A checkered plate deck shall be installed at the north end of the hatch. The implementation of this DCP will have no impact on - plant operations. , REASON FOR CHANGE: The reason for not previously finishing around the subject equipment hatch was to allow graater construction , access. 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 evaluated in the Safety Analysis Report. No accident previously evaluated in the UFSAR has a probability of occurrence which would be increased by changes implemented by this DCP. The handrail designs used appropriate loads defined by 29CFR1910, OSHA 2206, Section 1910.23. The added steel decking is a non-Category 1 structure and is designed in accordance with NPE's Civil Structural Design Criteria Manual, Section 8.3 and VBC. Concrete slab additions are designed for loads delineated by Section 3.8.6 of the UFSAR. The Turbine Building is structurally adequate for the increase hatch size. The increased hatch opening will not affect the accident analysis listed in the UFSAR. The addition of handrails, decking, and the finishing of concrete at the periphery of the equipment hatch will have no adverse impact on safety features or affect the l accident analyses listed in the UFSAR. The accidents previously L evaluated in the UFSAR were abnormal operation and design basis L accidents that involve the possible direct release of radioactive L material. The increase hatch size, the addition of handrails, decking, and concrete will not, in any manner, directly or indirectly, affect the probability of such accidents. Materials used for handrails meet the requirements established in UFSAR Table 9.5-11, i.e., structural components are non-combustible. Therefore, work performed by this DCP does not affect the function of any safety related equipment. The probability (frequency class) of a malfunction of equipment, or an increase within a frequency class for a malfunction of equipment are in no way increased due to the addition of handrails, steel decking, and concrete slab sections in the Turbine Building. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 138
1 Attachment to AECM-89/0093 l 1 i f NPE-89-077 , Dage 2 The addition of a handrail, and installation of steel decking and concrete around the equipment hatch in the Turbine Building will not directly or indirectly affect any design basis failure points or technical specification safety limits, i.e., the margin of safety-is unaffected. The changes to be implemented by this DCP are such that they are not introducing any new hazards or reducing' . t the margin of safety as defined in the basis for any Technical l Specifications, ' l 1 I i L 1 NLSATTC2/SNLICFLR - 139 w g-- w = r -- - -,-- ---y--y .- m
L
-Att chment to AECM-89/0093 r i
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SRASN: pLS-89-001 < DOC NO: 06-ME-1M10-R-0003 SYSTEM: M10 DESCRIPTION OF_ CHANGE:l This change performs the Drywell-Bypass Test (Surveillance procedure 06-ME-1M10-0003) with the reactor l vessel. head and the drywell head removed. , REASON FOR CHANGE: .This change to the test was made to allow for ; Drywell Bypass Test to be completed during refueling when the reactor vessel head and the drywell vessel head were removed. ! SAFETY EVALUATION: Although both the reactor vessel and drywell head are open during this test, other normally open passages are closed, and enough water remains in the refueling pool. to ensure
-no leakage of radioactivity. Therefore there is no increase in a the probability of an accident. ;
The only accidents which could be attributed to performing the ' Drywell-Bypass Test are accidents affecting the reactor vessel, the drywell pressure boundary, end the cooling of irradiated fuel, i The'only way for these types of accidents to occur is by loss of structural integrity of the drywell roof and refueling bulkhead.
- The. components of the drywell roof are stressed for full integrity under all levels of water in the refueling pool. No damage to the. ,
refueling bulkhead or.drywell roof occur due to the performance of
.this test. Therefore there is no increase in the consequences of .,
an accident, and no possibility of an accident different from any t evaluLted in the FSAR. For the above reasons, there is also no increase in the probability or consequences of a malfunction of equipment ; important'to safety and no possibility of an equipment malfunction ; different from any previously evaluated in the FSAR. Technical Specifications do not specifically identify any quantitative.or qualitative margins of safety that could be affected by performing this test with the reactor vessel head and drywell head removed. For this reason and the reasons listed above, there is no reduction in the margin of safety as defined in -' the basis for any Technical Specification. i NLSATTC2/SNLICFLR - 140
p Attachcent to AECM-89/0093 s i SRASN: PLS-89-002 000 NO: 06-RE-1C51-0-0001 SYSTEM: C51 DESCRIPTION OF CHANGE: Reactor Engineering performs a Local Power Range Monitor (LPRM) calibration pursuant to Technical Specification 4.3.1.1 every 1000 MWD /T. This corresponds to approximately once every five weeks. The normal sequence is performed over a two day period. On the first day a complete Transversing Incore Probe (TIP) set is completed. This provides an axial profile of flux at each LPRM string location in the reactor core. .For each LPRM detector in the core, a Gain Adjustment Factor (GAF) is calculated by relating the TIP reading to the LPRM reading as the TIP passes the LPRM readings to actual flux levels. On the second day the GAFs are used to adjust the LPRM amplifiers. This makes the LPRM console readings once again correspond to true flux. Once the adjustment is made to the LPRMs, another TIP set is run to calculate a new set of GAFs for the computer's use. On January 12, 1989 the first TIP set was run for an LPRM calibration. The TIP set was successfully completed. On January 13, 1989 the GAFs calculated on January 12 were used to adjust the LPRM amplifiers. After completion of the LPRM calibration the second TIP set could not be completed because TIP Machine C became inoperable. At this point the LPRM calibration was complete; however, GAFs (for computer use) could not be updated for the LPRMs. GAFs could not be updated for the LPRMs because of: (1) the computer requires all traces'to be completed before it calculates any GAFs; and (2) no information was available for 24 LPRM detectors that could not be scanned by TIP Machine C. Surveillance procedure 06-dE-1C51-0-0001 (Local Power Range l Monitor Calibration) was marked " partial procedure completed" with all technical specification acceptance criteria and all other steps / data acceptable. Also, process computer software change form PCSCF No. 89-01 was written to solve both constraints and GAF calculation. REASON FOR CHANGE: This change allowed the GAF's to be calculated and the LPRM's to be properly calibrated. 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 evaluated in the Safety Analysis Report. The TIP System and the Core Performance Monitoring System are power generation systems and are classified as not related to safety. The subject instrumentation and control systems are operational systems and have no safety function. NLSATTC2/SNLICFLR - 141
(T ~ , Attachm:nt to AECM-89/0093- i t PLS-89-002 , Page 2 l The TIP System activities performed did not change any function or operating characteristic which would result in creation of an t interface that~would have an adverse affect upon safety related components, equipment or systems. The Core Performance Monitoring activities performed did not change any function, parameter or operating characteristic which would result in creation of an interface that would have an adverse affect upon safety related components, equipment or ' L systems.- Also, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated i previously in the Safety Analysis Report. The activities performed did not adversely affect the LPRM calibration or the power distribution calculations, and therefore, does not reduce the margin of safety as defined in the basis for any Technical Specifications. f T e NLSATTC2/SNLICFLR - 142 _m -. . _ _ ._ _ .
p _ f- Attachment to AECM-89/0093 l SRASN: PLS-89-003 DOC NO: FCN-89-003 SYSTEM: P75 4 DESCRIPTION OF CHANGE: This change clarifies VFSAR Section I 8.3.1.1.2.4. The FSAR change does not require a technical i' specification change, because the technical specifications do not specifically require the performance of a combination LOP /ECCS : Actuation Test from standby conditions. Technical Specification i 4.8.1,1.2.d.7.a).2), requires that the diesel generator be ' subjected to a test which simulates a combined loss of power /ECCS actuation signal every 18 months, but does not specify the temperature conditions at which the test is to be performed. , Technical specification 4.8.1.1.2.d.9, requires that the ! combination loss of power /ECCS actuation test specified in i Technical 4.8.1.1.2.7.a).2), be performed at full-load temperature conditions. Regulatory Guide 1.108 requires that the combination i loss of power /ECCS actuation test be "re-run" at full load teniperature conditions, which implies that this type of test is performed twice, once at standby (ambient) temperature conditions and again at full-load temperature conditions. Since the ; technical specifications do not specify that the loss of power /ECCS actuation test has to be "re-run", the performance of surveillance requirement 4.8.1.1.2.d.7.a).2), loss of power /ECCS actuation, following surveillance requirement 4.8.1.1.2.d.9 will meet the technical specification requirements if it is performed once, at full-load temperature conditions. Therefore, the deletion of a combined loss of power /ECCS actuation test from standby conditions, with respect to engine temperatures, will not require a change to the GGNS Technical Specifications. REASON FOR CHANGE: This change to UFSAR Section 8.3.1.1.2.4 was made to clarify that the diesel generator would not have its performance tested twice under LOP /ECCS actuation at both standby conditions and at full-load temperature conditions. The test will be performed only under full-load temperature conditions. SAFETY EVALUATION: The change described does not introduce, modify or delete any events, components, or tests which will make any of the accidents previously evaluated in the UFSAR more likely to occur. The deletion of the combined loss of power /ECCS actuation test from standby conditions is intended to increase diesel generator reliability by eliminating a redundant test. The ability of the diesel generator to respond to a combined loss of power /ECCS actuation signal from standby conditions and full load temperature conditions has been satisfactory in past surveillance i tests. In addition, response of the diesel generators to this testing from both standby and full load temperature conditions is virtually identical when compared to each other. The testing l l which subjects the diesel generator to a loss of power signal from l standby conditions requires the diesel generator to respond to a 1 significant transient upon initiation of the standby service water pump motors, which is virtually equal to, or greater than the transientofthedieselgeneratorassumingitslargestsingleECCS pump starting load. The diesel generator s ability to
,, NLSATTC2/SNLICFLR - 143 l
l
r Attachment to AECM-89/0093 PLS-89-003 Page 2 L respond to a loss of power, ECCS actuation, and combination loss of power /ECCS actuation (from full load temperature conditions), will continue to.be demonstrated through surveillance testing. . The deletion of the testing described in the change does not make-the probability of a diesel generator loss more likely to occur. Therefore, there.is no increase in the probability of occurrence ' or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis l Report. ! Also, there is no creation of a possibility for an accident or o malfunction of a different type than any evaluated previously in the Safety Analysis Report. No new events are added which would , worsen a malfunction of the diesel generators. The change does not introduce any new testing which will result in a malfunction more severe than any previously evaluated. No new mooes of
; operation which could result in different types of failure are introduced. This change does not alter existing operating methods of the diesel generators. ;
The change described does not reduce the margin of safety as , defined in the basis for any technical specifications because the ability of the diesel generator to respond to a combination loss - of power /ECCS actuation signal is not reduced. The testing being deleted will be, in effect, duplicated by other tests which will remain due to technical specification requirements. The deletion of the combined loss of power /ECCS actuation test from standby temperature will not reduce the margin of safety because in the past the results of this same test at full load temperature conditions have been virtually identical by comparison. No failures of either the Division-I or II diesel generator have occurred which have been caused by physical changes due to stresses induced by temperature differences. Failures which have occurred have been of a coincident nature, and have not been linked to temperature induced changes, n l NLSATTC2/SNLICFLR - 144 l
Attachment to AECM-89/0093 I SRASN: PLS-89-004 DOC NO: 05-1-02-III-3, Rev. 18 SYSTEM: B33 DESCRIPTION OF CHANGE: As part of the BWR Owners Group efforts to , analyze the LaSalle Thermal Hydraulic instability issue, GE was
- contracted to perform analyses to alleviate NRC concerns. To date, a very limited scope of work has been performed with TRAC-BWR which is qualified for LOCA analysis. This model has not i been qualified for stability analyses, and therefore, there is a high degree of uncertainty in the calculational results when used for these purposes. This model demonstrated that transition >
boiling (i.e., a penetration of the MCPR safety limit) can develop > very rapidly on large core flow reductions that result in core flow being less than 40% when on or above the 100% rod line. A . more conservative approach has been mandated by NRC Bulletin ' 88-07, Supplement I, which will require a manual scram when operating with a calculated decay ratio line of greater than 0.9 as determined by the ANF reload analysis. To prevent operating in this region until formal calculations and reviews are complete, a power to flow map with more conservative operator ections is being incorporated. t This power to flow map requires the operator to manually scram the ' plant if operation is determined to be in the scram required region as defined in the figure. The detect-and-suppress Region I has been incorporated into Region IV and requires operator action to reduce power by inserting rods until leaving Region IV. . Completing this action in two hours or less eliminates the need for performing the LPRM noise surveillance of Technical Specification 4.3.10.1. I In addition, a manual scram is required on the trip of both recirculation pumps to DEE while in RUS mode. REASON FOR CHANGE: This change was made to satisfy NRC concerns . which will require a manual scram when operating with a calculated l decay ratio line of greater than 0.9 as determined by the ANF reload analysis. SAFETY EVALVATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. The restrictions do not create any conditions of operations not previously analyzed. No instrumentation setpoints or component operating conditions are modified. The operation is within parameters previously analyzed. FSAR Section 15.3, Decrease in reactor coolant system flow rate, evaluates the trip of two recirc pumps to off and assumes a reactor trip on level 8. This bounds a manual scram being inserted on two pumps tripping off. This operation and availability of systems used to mitigate the consequences of an accident previously evaluated are not changed by described restrictions. NLSATTC2/SNLICFLR - 145
p Attachment to AECM-89/0093 n PLS-89-004 L Page 2 i' Since no new modes of system operation are introduced and the effective power to flow operating map is bounded by the analyzed map, the possibility of an accident of a different type than previously evaluated accidents are not increased. The proposed restrictions do not physically change any equipment nor does it change any of the methods of manners in which the recirculation system is operated as described in the UFSAR or Technical Specifications. Since there are no changes to the equipment, there is no increase in the probability of occurrence of a malfunction of equipment important to safety, m The proposed restrictions meet or are more conservative than those described in Technical Specifications and the UFSAR. Responses to component failures or transients that result in operation in the Region-I or Region IV of Technical Specification Figure 3.4.1.1-1 still comply with the actions and LCO of Technical Specifications on the recirculation system. Consequences of a malfunction of equipment important to safety is not increased. The proposed restrictions do not introduce any new operating modes of equipment important to safety. They do not make any physical changes to any equipment. There is no possibility of a malfunction of equipment of a different type than previously-evaluated in the FSAR. Incorporation of this power to flow map does not create a condition of recirculation system configuration which would invalidate the LC0 for the recirculation system. It does not change the methods and manners in which the recirculation system is operated. All changes meet or are more conservative than existing Technical Specification LC0 and Action Statements for the recirculation system. Therefore the margin of safety as defined in the Technical Specification basis is not reduced. ll
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NLSATTC2/SNLICFLR - 146 /
p t Attachment to AECM*89/0093 SRASN:. PLS-89-005 DOC NO: Aux. Steam System SYSTEM: N12 DESCRIPTION OF CHANGE: This change disassembles / removes from the Water Treatment Building the following equipment which is part of the Auxiliary Steam System. two electric auxiliary boilers four auxiliary boiler feed pumps two auxiliary boiler deaerators two chem feed systems blowdown tank four sample coolers and the associated piping, hangers, control panels, electrical and I&C for all equipment. The Aux Steam System, Section 9.5.9.2 of the UFSAR is described to support the turbine butiding vent system, liquid radwaste evaporator, condensate deaeration / heating, RCIC and feed pump testing, main turbine shaft seal steam, and blanketing steam for the MSR's. REASON FOR CHANGE: The Aux Steam system is not being used for any of the above functions. Applicable procedures that are identified on the temporary alteration will be modified or revised to preclude the uso or requirement of the Aux Steam System. There are no plans to use Aux Steam to support any plant system function, therefore, the system can be disassembled and removed from this area to support RF03. SAFETY EVALVATION: The disassembly and removal of the Aux Steam Supply System will not increase the probability of occurrence of an accident previously evaluated since the system presently ir, not operable and has not been in use. Per UFSAR Section 10.4.3, Aux Steam could be used to provide sealing steam, but is not required to be redundant to the gland sealing steam system. The removal of 1 the Aux Steam System will not prevent the loss of condenser vacuum due to the loss of the seal steam generator. Operating without aux steam is within the confines of the present plant operating procedures, therefore, failure of this system will not compromise any safety related system or component and will not prevent safe reactor shutdown. Wherever the aux steam system has an interface with the nuclear steam system, a positive means of separation will be maintained. The Aux Steam System has no safety related functions as defined in Section 3.2. This system is not being used and is not intended to be used to support any plant system. The deletion of the system will not increase the consequences of an accident previously evaluated. NLSATTC2/SNLICFLR - 147 L
r-Attachment to AECM-89/0093 PLS-89-005 Page 2 i The possibility of an accident or malfunction of a different type has not changed with the removal /disabeling of the aux steam system. The aux steam system is not being used and all its intended functions are no longer a requirement for any plant system operation as outlined in the UFSAR. Since the aux steam system is no longer operational and all its intended functions are no longer required for any equipment support, the disassembly of this system will not increase the malfunction nor increase the probtallity of occurrence of a malfunction of equipment important to safety. The basis for any Technical Specification does not consider the use of the aux steam system. The removal / disassembly of the system does not reduce the margin of safety. NLSATTC2/SNLICFLR - 148
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7 if i ? Attcchment to AECM-89/0093
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.SRASN:- PLS-89-006 DOC NOx MWO-90989~ SYSTEM: P41 l 1 DESCRIPTION OF CHANGE: Water samples taken from each SSW basin l have been analyzed in preparation for chemical. cleaning of the E system piping. This activity will pump SSW water from the basin F to improve basin water quality. The water is drawn out of the ,
[ basin.and directed to the outfall. The discharge rate from the basin is limited to less than the available basin makeup capacity, 7 i F REASON FOR CHANGE: Safety Evaluation written to allow use of . temporary pumps to pump down the SSW basin. SAFETY EVALUATION: There is no increase in the probability of f occurrence or in the consequences of an accident or malfunction of p equipment important to safety previously evaluated in the Safety Analysis Report. There are no evident scenarios within the UFSAR which are initiated by not meeting any design requirements of the SSW system. , Long term (30 day) SSW capability was evaluated with SSW at l minimum basin 1cvel. This activity does not. adversely impact the ; L capability of the SSW system to meet it's design requirements. No 'l f c-object or equipment entered the basin that could pass through the protective screen, nor did any object or equipment clog the screen t- such that flow was reduced below design criteria. The surface I area of the equipment is less than the 2.5 percent of the total ( area of the protective screen. Furthermore, all objects and- - D equipment were tethered as to permit quick and simple removal from
.the basin if necessary.
This activity does not' alter the operation of the SSW system. Furthermore, the actions involved to accomplish this activity do l not involve significant hazards which could cause an accident - within the SSW system. There are no seismic or missile concerns involved with performing this activity. All objects can be easily i and quickly removed from the basin should conditions warrant. The equipment auction is maintained above the Technical Specification minimum basin level of 130'3", which corresponds to a level of 2'9" below the basin slab. Therefore this activity 2; does not impact ~ basin minimum level. A periodic check is made to , ensure equipment suction remains above this level. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in , the Safety Analysis Report. At no time during this activity does the SSW system fail to meet its design requirements. The integrity of the protective screen protecting the SSW pumps is maintained throughout the activity. Additionally, the equipment used is not small enough to penetrate the protective screen regardless of whether the pump (s) are running or nut. NLSATTC2/SNLICFLR - 149
p l. Attachment to AECM-89/0093 i PLS-89-006
; Page 2 This activity does not modify or increase the complexity of the
$$W system hence no new failure modes associated with equipment already installed are induced.
- There is no reduction in the margin of safety as defined in the basis for any Technical Specification. If all ob.jects that are placed in the basins are lodged on the protective screen, no increase in pressure across the screen or decrease in flow through the screen is observed. No change in system performance is seen because of the 1
- rge surface area of the protective screen. This activity does not modify or increase the complexity of the SSW
.- system hence no automatic actuation or system logic is bypassed or -
invalidated. 1 NLSATTC2/SNLICFLR - 150 ,
, i s Attachment to AECM-89/0093 SRASN: PLS-89-007 DOC NO: MSTI-1F15-S9-001-0-S SYSTEM: Fil DESCRIPTION OF CHANGE: The purpose of this safety evaluation is to evaluate portions of the testing specified by MSTI 1F15-89-001-0-S " Refueling Platform Test Q1F15E006". The MSTI is written to functionally test the Refueling Platform after completion of DCP 88/0010 Revision 0 and MCP 89/1001 Rev. O. The DCP enhances the operational reliability of the Refueling Platform in order to support the refueling outage schedule. The MCP provides for the ability to " plug-in" Lasertrac Automatic Positioning System.
REASON FOR CHANGE: In order to perform the MSTI, an interlock listed in the UFSAR, and addressed in Technical Specification 3.9.12, must be simulated. The interlock between HFTS and the Refueling Platform will be simulated in order to test the functionality of the Refueling Platform. 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 evaluated in the Safety Analysis Report. None of the existing FSAR accidents including fuel drop accidents in 15.7.4 and 15.7.6 have a probability of occurrence, because no new fuel or irradiated fuel bundles will be moved during the performance of MSTI 1F15-89-001-0-S. The accidents evaluated in the FSAR are concerned with fuel movement. Since no fuel movement will be allowed during performance of the MSTI, there is no increase in the consequences of an accident previously evaluated in the FSAR. The HFTS and the Refueling Platform will use a simulated vertical signal only to satisfy the operational functionality of the Refueling Platform. During performance of this MSTI, no dummy fuel bundles, no irradiated or new fuel bundles will be moved, 1-l therefore simulating the interlock will have no impact, and the l possibility of a new accident is not. postulated. Except for the transfer tube, the affected HFTS equipment is not safety related equipment and failure would not affect any safety related equipment required to prevent or mitigate the consequences
'of any previously evaluated FSAR accident. The Refueling Platform is safety related but only to the extent of it's structural ,
integrity which is not compromised in the performance of this l MSTI. Therefore the probability of occurrence of a malfunction of eauipment important to safety is not increased. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report since no new fuel or irradiated fuel i bundles are being moved during the MSTI. l H NLSATTC2/SNLICFLR - 151
7_ . Attachment to AECM-89/0093 PLS-89-007 Page 2 The basis for Technical Specification 3/4.9.12 is to control personnel access to the potentially high radiation areas (Room 1A525 - the Transfer Tube access area, Auxiliary Building elevation 185'-0") immediately adjacent to the tube and to assure safe operation. The interlocks between HFTS and the Refueling Platform are designed to prevent a fuel bundle from impacting the HFTS carriage. Since no fuel, new or irradiated bundles, will be moved during performance of MSTI IF15-89-001-0-S, the margin of l safety is not reduced for any Technical Specification. I-
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i NLSATTC2/SNLICFLR - 152
z Attachment to AECM-89/0093 l.. SRASN: PLS-89-008- DOC NO: CN-89-097 SYSTEM: R-61 J , DESCRIPTION OF CHANGE: This change connected the public address system installed in the engineering and maintenance building to the plant PA system. ' '~ REASON FOR CHANGE: This change was made to ensure that evacuation alarms and other plant requirements can be heard by personnel in the building. SAFETY EVALUATION: There is no accident previously evaluated in ' the FSAR that is affected by this change. The plant PA system is not connected to any other plant system other than receiving power (120VAC) from UPS panel 1Y83 which is not safety related and located in CAS in the admin building. There is adequate breaker protection in 1Y83 for this increased load. Therefore, the probability of occurrence of accidents remains the same. The consequences of accidents previously evaluated remains the same since this system is not connected to any other plant system except for power and adequate breaker protection is in place. This in-fact provides communications and evacuation alarms and decreases the chances or consequences of accidents previously evaluated. There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. Also, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. No basis for any Technical Specification is based on the public address system. Therefore no margin of safety in any Technical Specification basis changes because of this change. NLSATTC2/SNLICFLR - 153
E l L Attachment to AECM-89/0093 I SRASN: PLS-89-009 000 NO: FCR-89-005 SYSTEM: P41 DESCRIPTION OF CHANGE: This change to FSAR Section 9.2.1.2 allows blowdown of the SSW system coolant as necessary to prevent accumulation of fouling agents. , REASON FOR CHANGE: Blowdown is usually not needed to prevent the
! accumulation of fouling agents since there is not a significant best load on the system.
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 evaluated in the Safety Analysis Report. Blowdown of the SSW b: sins will still take place as necessary. as determined by the Chemistry Sampling Program. This change does not affect the method by which the SSW basins are blown down. No piping, penetrations, equipment, or other system tie-ins will be altered or affected by this change. The safety-related functions of the SSW system are not affected by this change. The ability of the SSW basins to mitigate an accident will not be affected because the method of operation of the systems, and the system design are.not altered by this change. The change does not affect the method of operating the SSW system. Blowdown will still occur in accordance with approved plant procedures. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. There is no reduction in the margin of safety as defined in the basis for any Technical Specification. Blowdown or the frequency l of blowdown is not used as a bases in determining SSW operability. I Blowdown will still occur as needed based on the Chemistry Sampling Program. l l NLSATTC2/SNLICFLR - 154
Attachment to AECM-89/0093 [ SRASN: PLS-89-010 DOC NO: FCR-89-001 SYSTEM: N/A i DESCRIPTION OF CHANGE: The changes to UFSAR Chapters 12 and 13 update these sections reflecting the reorganization of the Radiation Control and Chemistry Departments. Reporting to the Chemistry / Radiation Control Superintendent are the Chemistry
, Superintendent and Radiation Control Superintendent. The Radiation Control Superintendent is supported by the Technical s
Assistant to the Radiation Control Superintendent and Radiation Control Supervisors. The Chemistry Superintendent is supported by the Chemistry Supervisors. REASON FOR CHANGE: This change updates the description of these positions, the qualifications and responsibilities are also updated by this revision. ,
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SAFETY EVALVATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report, The changes are administrative in nature, reflecting organizational changes or providing minor clarifications in the descriptions of Health Physics programs or equipment. All duties performed by Chemistry / Radiation Control personnel will continue to be performed in the event of an accident. The changes in positions and responsibilities will not in any way lessen the effectiveness of the Radiation Protection er Chemistry progracs. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. These changes do not reduce the margin of safety as defined in the basis fcr any Technical Specification. The changes have no effect on the limiting conditions for operation, applicability, action or surveillance requirements as defined in any Technical Specifications. 1 L NLSATTC2/SNLICFLR - 155 ,
i Attachment to AECM-89/0093 i SRASN: PLS-89-011 000 NO: 060P-1C71-V-0002-Rev. 24 SYSTEM: FIS r. DESCRIPTION OF CHANGE: These changes affected the following l sections of Surveillance Procedure 06-0P-1C71-V 0002: - Sections 5.6.11 and 5.6.12: These sections govern the removal of the Refueling Platform (RP) and Fuel Handling Platform (FHP) from the HFTS upender zone (UPZ) following verification of the UPZ > interlock. The existing text requires manual operation using handwheels while the proposed text will accomplish the operation , electrically using the override circuit installed by DCP 88/0010. The maneuvers were evaluated in Safety Evaluation CFR 88/0010R00 and are not included in the scope of this evaluation. Sections 5.8 and 5.9: The proposed changes to these sections are editorial only. No impact on the UFSAR, Tech Specs or Design is created. Section 5.10: The proposed changes to 5.10.1 are , editorial only. , Sections 5.10.2 and 5.10.3 govern the operability verification of the primary and redundant overload cutoff interlocks for the RP Monorail Hoist and Frame Mounted Hoist, respectively. The existing text is replaced in its entirety by directives pertaining specifically to the new, electronic load cells installed per DCP 88/0010. The new load cells and their adequacy from the operability / reliability standpoint were evaluated in Safety , Evaluation CFR 88/0010R00 and are not included in the scope of : this evaluation. Section 5.11: This section governs the operability terification of the Up Travel stops on the Main Hoist, Frame Mounted Hoist and Monorail Hoist. The existing text, including TCH #17, is replaced in its entirety in crder to expand and clarify the required ' directives for all three hoists independently. For the Main Holst, none of the requirements are changed. However, for the Frame Mounted and Monorail Hoists, the surveillance directives are changed to make the actual, demonstrated " maximum up" position consistent with the design intent and documentation including GEK 75573. The original text was intentionally conservative with
- respect to the design, since it required the verification of safe water shielding (7 feet) at the maximum up position. The RP Frame Mounted and Monorail Hoists are not used to itft fuel. Also, the normal-up positions for both of these auxiliary hoists are achieved by redundant limit switches. The proposed changes will demonstrate the ability of both normal-up switches to stop the hoist cable load fitting at 7 or more feet below the water '
surface. The maximum-up setting will be set per design (GEK 75573) at or near the Mechanical jam-stop position.
. NLSATTC2/SNLICFLR - 156
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- Attachment to AECM-89/0093 f-PLS-89-011 Page 2 l Sections 5.12 and 5.13
- These sections are only affected editorially (renumbered as 5.13 and 5.14) due to the addition of a new Section 5.12. This Safety Evaluation is necessary since the changes to Section 5.11 affect the UF$AR. The scope of this evaluation, however, is limited only to those changes which affect the UFSAR and changes related to the. surveillance $ which satisfy the referenced Tech Specs.
REASON FOR CHANGE: In preparation for RF03, a general revision to Surveillance Procedure 06-0P-1C71-V-0002, " Refueling Interlock Check", is necessary to make editorial changes and process . clarifications, to eliminate conservative inconsistencies between the " design" configuration and the surveillance process, and to streamline the surveillance process in order to support the refueling outage. Spec:fically, Sections 5.6.11, 5.6.12 and 5.8 through 5.13 of the procedure are affected. 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 evaluated in the Safety Analysis Report. No accident evaluated in the UFSAR has probabilities of occurrence associated with performing Refueling Interlock surveillances. Nor does any UFSAR accident identify probabilities of occurrence which are sensitive to the manner or process by which Refueling Interlock surveillances are performed. However, various instrumentation systems important to safety including the Refueling system itself and RC&IS are dependent upon the subject surveillances for assuring reliable and proper operability. In all cases, the revised procedure will delineate a process that remains within the design parameters and configuration of the refueling equipment as defined by design documentation including GEK 75573. In addition, the purpose of the surveillance procedure is to identify and correct problems before they contribute to the probability of a UFSAR accident occurrence. The extent or frequency of the surveillances are not decreased in the revised procedure. No UFSAR evaluated accident has consequences associated with performing refueling interlock surveillances or the manner or process by which such surveillances are performed. In fact, the safety functions of the refueling equipment, interlocks, and other interfacing instrumentation systems (including RC&IS) are intended more for accident prevention (to decrease accident probabilities) than for mitigation of consequences. The consequences of an accident caused by malfunction of the refueling equipment or interlocks tested by the revised procedure would be the same as the accident consequences if tested by the existing procedure. The two major concerns would be spent fuel mishandling and control NLSATTC2/SNLICFLR - 157 yan,y q e . - - ,
, Attachment to AECM-89/0093 i
e PLS-89-011 Page 3 rod withdrawal during Mode 5. For spent fuel mishandling, the consequences described in the fuel handling accidents (UFSAR 15.7.4 and 15.7.6) would apply and would be unaffected. The consequencas of withdrawal of control rods during Mode 5 would be unacceptable. For this reason the UFSAR describes the refueling interlock design features and administrative controls that prohibit control rod withdrawal. The changes to the surveillance procedure does not affect the refueling system design, and the procedural / administrative controls are maintained adequate since the extent and frequency of the affected surveillances are not decreased. In addition, constant control room communications and other Technical Specification requirements are not affected, and the requirement that only the main hoist is used to handle new fuel is unchanged. No equipment which is important to safety in the plant is affected by the proposed procedure revision other than the refueling equipment and interlocks. The purpose of the subject surveillance procedure is to demonstrate operability of the refueling equipment and interlock functions. As described in the UFSAR, the probability of occurrence of a malfunction of the applicable refueling equipment or refueling interlock functions is minimized
=by 1) the design of the applicable refueling system components and structures and by 2) procedural / administrative controls governing the conduct of refueling operations. The proposed changes to the subject surveillance procedure do not violate any design requirement, intent, or specification. Redundancy, diversity, and separation requirements of the original design are maintained.
The adequacy of the administrative controls required to assure low probability of malfunction of refueling equipment and interlocks is maintained since the extent and frequency of the applicable surveillances are not affected. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Only the margins of safety defined for the bases of Tech Spec 3/4.9.6 and 3/4.9.12 apply since affected portions of the revised procedure are required to demonstrate compliance with Tech Specs 3/4.9.6.1 and 3/4.9.12 only. The basis for 3/4.9.6 1s unaffected by the procedure revision becausc acceptable surveillance results of all applicable interlock functions will ensure that only the RP Main Hoist is capable of lifting fuel and that the Main Hoist , capability is sufficiently limited to protect the reactor and its core internals from excessive lifting forces. The basis for 3/4.9.12 is unaffected since the affected surveillance directives will still demonstrate the operability of all required HFTS 1 l NLSATTC2/SNLICFLR - 158 1 l ;
" Attachment te AECM-89/00931 hr. f? PL'S-89-011 p, 'Page 4
- n L- interlocks pF.;r to their use. . The requirement to verify that r Room 1A525 is sealed and is unaffected, and the new provisions to remove the RP and FHP from the upender zone after performance of the UPZ interlock check were' previously evaluated -in Safety Evaluation CFR 88/0010R00-in association with DCP 88/0010. Thus, no Technical Specification margins of safety are reduced.
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i NLSATTC2/SNLICfLR - 159
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y k b, Attachment to AECM-89/0093 la c p L SRASN: PLS-89-013 . DOC NO: 06-0P-1F11-V-0001 % v. 24 SYSTEM: Fil. , DESCRIPTION OF CHANGE: The following summarizes the st.vpe of the changes to Surveillance Procedure 06-0P-1F11-V-0001: E .Section 5.6: This section governs the operability verification of the overload cut-off interlocks for the fuel handling platform
'(FHP) monorail hoist. TCN #4 added Sections 5.6.1b and 5.6.21 to provide administrative controls for removing and reinstalling the-upstop blocks to allow testing of the hoist-interlocks. However, :
proposed changes to Section 5.7 (below) eliminates the need for- ; the TCN #4 changes. Thus, this procedure revision will delete Sections 5.6.lb and 5.6.21. Section 5.7: This section governs the operability verification of the Up Travel stops on the Monorail Hoist. The existing text is replaced in its entirety in order to expand and clarify the , required directives and make the actual, demonstrated " maximum up" position consistent with the design intent and documentation including GEK 75577. . The original text was intentionally conservative with respect to the design, since it required the verification of safe water shielding at the maximum up position. The normal-up position is achieved by redundant limit switches. The proposed changes will demonstrate the ability of both normal-up switches to stop the hoist cable load fitting at 7 or ' more-feet below the water surface. The maximum-up setting will be set per design (GEK 75577) at or-near the Mechanical jam-stop position.. REASON FOR CHANGE: In preparation for RF03, a general revision to Surveillance Procedure 06-0P-1F11-V-0001, " Fuel Handling Platform Interlock Check", is necessary to eliminate TCN #4, to eliminate conservative inconsistencies between the " design" configuration and the surveillance process, and to streamline the surveillance process in order to support the refueling outage. Specifically, Sections 5.6 and 5.8 of the procedure are affected. . SAFETY EVALUATION: There is no increase in the probability of 3
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occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. No accident evaluated in the UFSAR has
. probabilities of occurrence associated with performing Fuel Handling Platform Interlock surveillances. Nor does any UFSAR accident identify probabilities of occurrence which are sensitive to the manner or process by which Fuel Handling Platform Interlock surveillances are performed. The revised procedure will continue to delineate a process that remains within the design parameters and configuration of the fuel handling equipment as defined by design documentation including GEK 75577. In addition, the purpose of the surveillance procedure is to identify and correct problems before they contribute to the probability of a UFSAR accident occurrence. Since the extent or frequency of the NLSATTC2/SNLICFLR - 160
Attachment to AECM-89/0093 PLS-89-013 Page 2 surveillances are not decreased in the revised procedure, the probability of occurrence of-previously evaluated accidents is not increased. No UFSAR evaluated accident has consequences associated with performing fuel handling platform interlock surveillances or the-manner or process by which such surveillances are performed. In
, fact, the safety functions of the fuel handling equipment and ,
interlocks are intended more for accident prevention (to decrease ! accident probabilities) than for mitigation of consequences. The ! consequences of 'an accident caused by malfunction of the fuel s handling equipment or interlocks tested by the revised procedure would be the same as the accident consequences if tested by the ; existing procedure. The major concern would be spent fuel i mishandling for which the consequences described in the fuel handling accident (UFSAR 15.7.4) would apply and would be ,
, , unaffected. Also, the revised procedure does not affect the Fuel Handling Platform design, and the extent and frequency of the affected surveillances a~e not decreased. Thus, the consequences of an accident previously evaluated-in the UFSAR are not j increased. .;
There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The main Fil system accident prevention function is to prohibit unsafe operations of applicable ! fuel handling equipment / components. No new functions are created ;
'by the revision of the subject surveillance procedure, and the !
existing functions deter the fuel handling accident which is -j
-already analyzed in UFSAR 15.7.4. This. accident is' the worst case l accident for mishandling spent fuel, and the cause, sequence of events, and assumptions are unaffected by the subject procedure revision. The design of the Fil system is not changed and the '
procedural / administrative controls are maintained adequate to demonstrate operability of the applicable components and functions. The extent and frequency of the required surveillances 4 are not decreased and the original redundancy, diversity, and I separation requirements of the Fil equipment are maintained. Therefore, creation of a new accident scenario is not postulated. , Malfunction of fuel handling equipment or interlocks which could potentially cause postulated accidents are already considered in the UFSAR. A fuel handling accident as analyzed in UFSAR 15.7.4 would be the worst case result of a spent fuel mishandling " accident caused by Fuel Handling Platform equipment malfunction. The design of the Fil system is not changed and the procedural / administrative controls are maintained adequate to demonstrate operability of the applicable components and functions. The extent and frequency of the required surveillances are not decreased, and no new or different surveillances are incorporated into the revised procedure. Thus, none of the Fil P equipment with safety critical functions different than those NLSATTC2/SNLICFLR - 161
Attachment to'AECM-89/0093' e , PLS-89-013 Page 3
-previously evaluated in'the UFSAR are affected. Also, no other interfacing plant equipment whose-function, intent, or reliability during fuel handling is affected because no design change is involved. Therefore, creation of a new malfunction of equipment scenario is not postulated.
No equipment which is important to safety in the plant is affected by the proposed' procedure revision other than the Fuel Handling Platform and interlocks.- The purpose of the subject surveillance ' procedure is to demonstrate operability of the Fuel Handling Platform interlock functions. As described in the UFSAR, the probability of occurrence of a malfunction of the applicable Fuel . -Handling' Platform equipment or interlock functions is minimized by
- 1) the design of the applicable Fil system components and structures and by 2) procedural / administrative controls governing the conduct of fuel handling operations. The proposed changes to the subject. surveillance procedure do not violate any design requirement, intent, or specification. Redundancy, diversity, and separation requirements of the original design are maintained.
The adequacy of the administrative controls required to assure low probability of malfunction of fuel handling equipment and interlocks is maintainea since the extend and frequency of the applicable surveillances are not affected. The requirement to verify or place the load override switch in the 500 pound position (Section 5.6.1) is unaffected. Also, the Technical Specification requirement for verifying the 500 pound load switch setting during , new fuel movement is unaffected. Thus, the probability of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased. No UFSAR evaluated malfunction of equipment has consequences associated with performing fuel handling platform interlock surveillances or the manner or process by which such surveillances are performed, The safety. functions of the Fuel Handling Platform equipment and interlocks are intended more for decreasing the probability of equipment malfunctions (prevention) than for consequence mitigation. The consequences of malfunction of the fuel handling equipment or-interlocks tested by the revised procedure would be the same as the malfunction consequences if tested by the existing procedure. The major concern is prevention of spent fuel mishandling for which the consequences analyzed in the fuel handling accident (UFSAR 15.7.4) would apply and would be unaffected. The changes to the surveillance procedure does not affect the Fil system design, and the extent .snd frequency of the affected surveillances are not decreased. Thus, the consequences of a malfunction of equipment important to safety previously evaluated in the UFSAR is not increased. NLSATTC2/SNLICFLR - 162
['; p Attachment to AECM-89/0093' 1 PLS-89-013 Page'4 Only the margins of safety defined for the basis of Technical Specification'3/4.9.6 apply since affected portions of the revised procedure are required to demonstrate compliance with Technical Specification 3/4.9.6.3 only. The basis for 3/4.9.6 is unaffected ' by the procedure revision because the Fuel Handling Platform is physically located in the auxiliary building and concerns relative to fuel handling in the RPV or inadvertent engagement of core
' internals by the FHP do not apply. Also, the portion of the procedure which verifies the mail hoist interlocks is not changed.
Thus, no Technical Specification margins of safety are reduced. l I 7 l L-I 1 1 NLSATTC2/SNLICFLR - 163
f . l g . Attachment to AECM-89/0093
- SRASN:_ PLS-89-014 DOC NO: S.O.I. 04.-1-01-E12-1 SYSTEM: E12 l
DESCRIPTION OF CHANGE: This 10CFR50.59 evaluation covers the use of the LPCI flow injection path of RHR~as an alternate method of decay heat removal. Except for the final injection point, the
-flow path is identical to the RHR normal' shutdown cooling mode.
This evaluation assumed a maximum water temperature of 200*F and a , maximum differentia 1' temperature between reactor water and RHR return water of 70*F. Also, the use was-limited to two. injections ( with continuous flows of 7900 gpm'for thirty days or less (8100 gpm is-acceptable for one hour or less). REASON FOR CHANGE: The reason for this change was to evaluate - LPCI flow injection path of RHR as an alternate method of decay heat removal. SAFETY EVALVATION: There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety ; Analysis Report. The following accidents were deemed applicable to this item.
- 1. Failure of RHR Shutdown Cooling The subject mode of operation only entails a change to the final injection path of SDC flow. All other SDC functions are per normal operation. Use of the LPCI injection line
. flow path creates no new failure modes and therefore does not increase the frequency classification of this accident.
- 2. Inadvertent Shutdown Cooling Operation This accident is relative to safety only.if the reactor is not in a shutdown mode'and is operating at or near criticality. -The LPCI alternate SDC mode will only be utilized after the reactor is shut down. The reactor will <
not be returned to criticality with-this mode of SDC in effect. This item is evaluated from the standpoint that the LPCI mode of RHR is considered inoperable when utilizing the normal LPCI injection path for Shutdown Cooling. As discussed in the Nuclear Safety Operational Analysis (FSAR Appendix 15A), the allowable Technical Specifications LCOs for ECCS eouipment are formulated based on evaluations of essential protection sequences such that unacceptable safety results do not occur. Therefore, having the LPCI mode inop per the guidelines of the GGNS Technical Specifications will not increase the consequences of any evaluated accidents. It is also noted that even with LPCI inop, the system could be realigned for this mode of operation if required. Other modes of RHR affected by use of the LPCI flow path are not required for Mode 4 or Mode 5 operations. NLSATTC2/SNLICFLR - 164
- r; '
q f < Attachment to AECM-89/0093- 3 1 PLS-89-014 Page 2 ) These modifications do not change the system configuration or basic operational philosophy in such a way as to cause any accident that has not been rreviously analyzed in the FSAR. Specifically, the potential _for draining the vessel is not increased since normal RHR interlocks and isolation functions remain in effect.
' Operation with flow limited to 7900 gpm insures that for the LPCI thermal sleeves and flow deflectors, flow induced fatigue is minimized while utilizing the LPCI injection path for_ Shutdown Cooling.
There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. During operation in this mode the
' Technical Specifications requirements for redundant methods of decay heat removal (Shutdown Cooling) will be maintained, thereby mitigating the consequences of equipment malfunctions. Use.of the LCPI flow injection path will be used only if shutdown cooling return to feedwater, F053A(B) is not available Ausi loss of the other SDC loop has created an emergency condition. The FSAR has analyzed loss of Shutdown Cooling and RHR Shutdown Cooling .
malfunctions. The use of the LPCI injection flow path has not changed the design function of RHR for Shutdown Cooling to such an extent that would place the system in mode which could lead to possible malfunctions outside the' existing boundary of the FSAR t analysis. As previously stated, operating restrictions on flow and temperature insure no new type malfunctions occur. Sufficient decay heat removal and adequate mixing for accurate temperature indication exists for this mode of RHR Shutdown Cooling. The adequacy of the alternate LPCI mode of Shutdown Cooling has previously been demonstrated. An actual vessel injection is not required to be performed since the demonstration of the design-SDC flow rato of 7,450 GPM has been shown from other operability tests such as suppression pool to suppression pool. Therefore, the margin of safety as established'in basis of the Technical Specification has not been degraded. NLSATTC2/SNLICFLR - 165
y ., Attachmsnt to'AECM-89/0093 L , ] SRASN: PLS-89-015 DOC NO: CR-89-008- SYSTEM: VAR DESCRIPTION OF CHANGE: Chapter 18.1.30 of the UFSAR concerning the leakage reduction program is being changed to add leak detection methods and consolidate information. REASON FOR CHANGE: The Leakage. Reduction Program is to be consolidated with the ISI 10 Year Plan Pressure Testing Program, (with the exception of The Post Accident Sampling System which is implemented by Chemistry's Surveillance Program). In order to accomplish this change the methods for detecting leakage specified. in the UFSAR chapter 18.1.34 are being revised to agree with the r methods used in the Pressure Testing Program. 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 previoosly evaluated in the Safety Analysis Report. -These leak detecting methods are per the ASME - Section XI Pressure Testing Program which satisfies Tech. Spec. 3.4.8 for structural integrity of ASME Code Class 1, 2 and 3 , components. These methods are equivalent to the methods currently in the FSAR. The test will be ran in accordance with approved-
-plant procedures. Therefore the change in leak detection methods '
,does not increase the probability of occurrence of an accident previously evaluated in the FSAR. The consequences of an accident
~
previously evaluated remains the same since the proposed leak detection methods are equivalent to the methods currently _in the-FSAR. The possibilities of an accident of a different type than any evaluated in the FSAR remains the same since the proposed leak detection methods are equivalent to the methods currently in the FSAR. The--probability of occurrence of a malfunction of equipment important to safety previously evaluated remains the same since the proposed-leak detection methods are equivalent to the methods currently in the FSAR. The consequences of a malfunction of equipment important to safety-previously evaluated remains the same_ since the proposed leak detection methods are equivalent to the methods-currently in the FSAR. The possibility of a malfunction of a different type than any evaluated previously - remains the same since the proposed leak detecting methods are equivalent to the methods currently in the FSAR. The basis for any Technical Specification safety margin does not address leak detection n'ethods for the Leakage Reduction Program. The addition of equivalent leak detection methods does not reduce the margin of safety as defined in the basis for any Technical Specification, i NLSATTC2/SNLICFLR - 166
m I, Attachment to AECM-89/0093 x =SRASNL PLS-89-016 DOC N0: CR-f'9-009 ~ SYSTEM: VAR-r DESCRIPTION OF CHANGE: . This evaluation is written to approve the use of the Westinghouse Sensor Response Time Test'(SRTT) System, which utilizes the process noise analysis method, to measure the time response of GGNS process sensors required by Technical Specification to be tested. The following information on the Westinghouse SRTT was taken from the Westinghouse report titled "THE USE OF PROCESS NOISE MEASUREMENTS TO DETERMINE RESPONSE CHARACTERISTICS OF PROTECTION SENSORS IN.U.S. PLANTS" which was
- submitted to the NRC by Byron 1 and Millstone 3 to provide !
justification for the use of this technique for response time testing. The staff reviewed the Westinghouse report which-describes the test method and provides the results of tests conducted at operating reactors from 1977 through 1982 using this technique. The staff concluded that the use of process noise ! measurements will provide an acceptable means to fulfill the
-. requirements for response time testing as specified in each plants Technical Specifications. Plant Staff has reviewed copies of portions of Byron's and Millstone's Technical Specifications including their Bases and determined that even.though these plants are PWRs there is no significant difference in the generic time response requirements imposed by their Technical Specifications and Grand Gulf's.
REASON FOR CHANGE: This evaluation is written to approve the use of the Westinghouse Sensor Response Time Test System. 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 evaluated in the Safety g Analysis Report. Per Sections 7.2.2.1.2.3.1.9, 7.3.2.1.2.3.1.9, 1 7.3.2.2.2.3.1.9 and 7.6.2.7.2 of the GGNS UFSAR (Capability for Sensor Checks IEEE Std. 279-1971) for RPS, ECCS, CRVICS and E0C-RPT, testing may be performed on transmitters one at a time under administrative control. The testing referred to includes
- taking the transmitters. out of service by valving them out and l
injecting a simulated process signal, which is the way time
- responses are done on GGNS transmitters at this time using the dynamic hydraulic ramp test method as described in Section 7.2.1.1.4.8 of the UFSAR. This requires a considerable amount of out of service time for each sensor channel tested. Also Section 7.2.2.1.2.4.3 of the UFSAR (Branch Technical Position EICSB 22) states that the test must be performed in overlapping portions so that an actual reactor scram will not occur as a result of the testing.
NLSATTC2/SHLICFLR - 167
Attachment-to AECM-89/0093
- y'
}
PLS-89-016
'Page 2--
In' comparison the Westinghouse SRTT method utilizing process noise 1 analysis ~is a passive method of measuring response times and does a not require the human involv'ement of valving out and injecting'a. signal into-the process _ transmitter. During SRTT testing, the
' signal analyzer is connected in parallel with tha transmitter outputisignal at-the trip unit in the control room. The
-instrument channel being tested is never out of-service, reflects .
current process conditions and is capable of performing its i intended function should a plant transient occur. The SRTT can test up to four (4) process transmitters at one time which 4 significantly reduces the time necessary to perform required response time testing. This will be administrative 1y controlled
- by procedure to ensure that only one of the redundant transmitters is, tested at a time and that divisional separation is maintained during testing to further reduce any possible impact of testing on the plant.. This does not preclude using the SRTT to measure four .
(4) different process sensors concurrently in the same trip
' channel or division.
1 L' Per Sections 7.2.2.1.2.3.6, 7.3.2.1.2.3.4, 7.3.2.2.2.3.6 and 7.6.2.7.2 of the UFSAR-(Compliance with IEEE Standard 338-1971) the RPS, ECCS, CRVICS and E0C-RPT systems are testable during ' reactor-operation and testing will completely test the sensors through the final actuators in overlapping portions to demonstrate
~the independence of channels and reveal any credible failures.
l: The Westinghouse SRTT, which has already been accepted by the NRC for use at other nuclear plants (e.g. Byron and Millstone) for their response time testing, will provide a more accurate and repeatable response time test because it will not only test the transmitter portion for overlap but also the signal from the l
. control room trip units all the way back to the actual process
- including the sensing lines. This should reduce the' probability of occurrence by providing a more realistic test of the
. process-to-sensor-to-trip unit response time instead of just the sensor's response time. It can also be noted that this test method can detect incorrect wiring and improper instrument valve lineups. Measurements using the hydraulic ramp test method were used to compare the accuracy of measurements taken using the noise analysis technique and that the response times calculated using the noise analysis tended to be more conservative both with respect to ramp tests and vendor specs. This could be partially attributed to the fact that impulse lines are valved out when using the hydraulic ramp technique and are not taken into account.
Since all measurements are done by computer, the impact of human error and the element of inaccuracy due to human interpretation is removed. The data should be much more repeatable for trending to determine sensor degradation provided adequate administrative controls are established by procedure to ensure that the plant is in normal steady state operation close to full power while measurements are being taken and that if a transient should occur during this time the measurements are voidal and repeated when steady state operation is resumed. NLSATTC2/SNLICFLR - 168
e,, rc .. ?, T ', ' Attachment to AECM-89/0093' ' ?! , s , 6 PLS-89-016 : Page 3 i- -The Technical Specification Bases for RPS, ISOLATION ACTUATION. INSTRUMENTATION, ECCS AND E0C-RPT explain that the measurement of
. response time at the ~specified intervals provide assurance that the protective functions of each specified channel will be g completed within the time-limit assumed in the accident analysis.
Therefore,:there is no creation of a possibility for an' accident or malfunction of a different type than any evaluated previously *
-in the Safety Analysis-Report. Also, there is no reduction in the-margin of safety as defined .in the basis for any Technical n
Specification, [ NLSATTC2/SNLICFLR - 169
p, . 1
-Attachment to-AECM-89/00932
?
-SRASN: :PLS-89-017- DOC NO:- ANSI-N13.5-1972 SYSTEM: N/A >
~ DESCRIPTION OF CHANGE: Appendix 3A;of the UFSAR provides a project position statement regarding compliance with Regulatory Guide 8.4, which states that ANSI-N13.5 provides an acceptable ,
basis'for selection and use_of direct-reading and indirect reading pocket dosimeters. Performance of the tests in ANSI 13.5 are not cost-effective if performed by utility personnel. Certificates provided by the manufacturers can be required.as-part of the purchase agreement which will show proof of compliance with ANSI N13.5. The. position statement is changed to read as follows: Pro.iect-Position: Grand Gulf Nuclear Station complies with Regulatory-Guide 8.4, with the following clarifications: (1) The tests required by ANSI N13.5 are not cost-effective if , performed by utility personnel. (2) The vendor (s) supplying dosimeters to Grand Gulf Nuclear Station shall be required to supply certificates of compliance with ANSI N13.5 as part of the procurement
- contract.
REASON FOR CHANGE: This change provides a more cost-effective method of complying with ANSI N13.5. 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 evaluated in the Safety
' Analysis Report. Examination of Chapter 15 of the UFSAR shows no '
o correlation between selection and use of self reading pocket L dosimeters and the accidents evaluated. Selection and use of L - pocket dosimeters has no effect on safety related systems or systems important to safety. No interfaces exist between pocket dosimeter selection /use and plant equipment; therefore, the probability of occurrence or consequences-of a malfunction of equipment important to safety as previously evaluated in the UFSAR is not increased. Because there are no interfaces between pocket dosimeters and plant equipment or systems, the proposed changes do not create the possibility of a malfunction of a different type than any evaluated previously in the UFSAR. The selection and use of pocket dosimeters for L measurement of X- or gamma radiation from sources external to the body does not change any limiting conditions for operation, applicability, actions or surveillance requirements for any Technical Specifications. Therefore, the margin of safety as defined in the bases for any technical specifications is not reduced. NLSATTC2/SNLICFLR - 170
ff 5% t Attachment to AECM-89/0093 t i SRASN:- pLS-89-018 DOC.NO: ANSI-N42.3 SYSTEM: N/A r i F DESCRIPTION OF CHANGE: Grand Gulf.UFSAR Appendix 3A, page p L3A/8.6-1 states the Project Position-is to comply with Regulatory ~ Guide 8.6-1973 which endorses ANSI N42.3/IEEE Standard 309-1970.
.This change requests proposes that this position statement be
. revised as follows:
Project Position: The tests described in ANSI N42.3/IEEE 369-1970 do not provide any information which affects the calibration accuracy of the instruments using Geiger Muller counters which are used by Health Physics at.GGNS. g Therefore, the GGNS position is not to comply. To perform all tests described in ANSI N42.3/IEEE 309-1970 would require approximately 80 man-hours per instrument and would { require testing circuits not actually used at GGNS. I The required tests would.be performed using test circuits which twould not match the actual circuits to which the G-M counters are
,g connected. The instruments used at GGNS to which the d-M counters are connected to are not designed to perform the described tests.
There are 4 types of instruments which use Geiger-Muller counters: scalers,-count rate instruments, dose rate instruments and dose h ' instruments. The instruments with the counters are calibrated following the recommendations of ANSI N323-1978 and the manufacturer. All1 calibrations are performed using National Bureau of Standards traceable sources. For scalers and count rate instruments, the proper. operation of.the-instrument is verified, the response of the counter and the instrument is determined and-the efficiency and correction factor is determined. For dose and dose rate-instruments, the instrmment is adjusted to indicate the i actual exposure or exposure rate to which the counter is exposed. REASON FOR CHANGE: This change was made because the tests
' described.in ANSI N42.3/IEEE 309-1970 do not affect the calibration accuracy of the instruments used at GGNS, 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 evaluated in the Safety Analysis Report. Health Physics portable radiation monitoring b instruments using Geiger Muller counters have no offect on the operation of any plant system. Therefore, a change in any calibration requirement will have no effect on the probability of an occurrence of an accident previously evaluated in the ?dAR.
Health physics portable radiation monitoring instruments provide no interface with any plant systems. Therefore, a change in calibration requirements cannot have any effect on the consequences of an occurrence of an accident previously evaluated in the FSAR, nor can it create the possibility of an accident of a i' different type than any evaluated in the FSAR. NLSATTC2/SNLICFLR - 171
? Attachment to AECM-89/0093 , PLS-89-018-Page 2' Because there is no' interface between portable Health Physics instrumentation and the operation of plant systems,-a change in calibration requirements can neither increase the probability or the consequences of a malfunction of equipment ~important to safety previously evaluatedLin the FSAR, nor could:it create the-possibility of a. malfunction of a different type than any previously evaluated in the FSAR.
llealth Physics portable instrumentation using Geiger Muller .
- c. counters are not addressed in any Technical Specifications. The i calibration of.these instruments cannot change any limiting .
conditions for operation, applicability, actions or surveillance.
~
requirements for any Technical Specification, .Therefore, the margin of safety as defined in the bases-for any Technical ;
. Specifications is not reduced. !
u { l
)
i l-i
.i NLSATTC2/SNLICFLR - 172
r r, , , s-- , Attachment-to AECM-89/0093
- SRASN: 'PLS-89-019 DOC NO: TEMP ALT-89-0003 SYSTEM: FIS
- 1 DESCRIPTION OF CHANGE: The-GE Lasertrac system is a refueling platform automatic positioning system used to reduce refueling-time that is normally consumed by manual operation of the platform- : , by an operator. Operator manual' control of the platform requires expert judgement and skill by the operator in locating.the platform over the fuel assembly and in moving the platform to and :
from the fuel storage and transfer locations, ! t Lasertrac is a computer controlled positioning system with a graphics terminal. The platform "x" and "y" positions are L provided by light transmitting and receiving devices. -The "z"- mast position-is provided by the existing encoder system. The platform "x" direction " laser" is mounted on the bridge and
'uses a reflector that is mounted to the Containment wall to provide accurate position of the' platform. The trolley "y" direction " laser" is mounted on the. trolley and uses a reflector that is mounted at one end of the bridge to provide accurate trolley position. The "z" mast position is taken from the existing encoder system to provide mast position information for system control. All existing bridge, trolley, and_ mast interlocks '
and boundaries must be satisfied for Lasertrac to operate the platform and mast. Any one existing interlock or boundary trip will immediately stop the refueling platform.or mast. When_ installed and calibrated, the Lasertrac system allows the operator to input a move command to any defined fuel location and
~
l the. system will move the bridge and trolley simultaneously at the l- maximum allowed speed to that location via a pre-defined and . L prescribed-shortest distance straight line path. The file move command allows a set of moves to be entered from a file and i executed at the command of the operator. The "z" mast position is controlled to be witi'in all defined boundaries during fuel movement. Lasertrac has its own built-in boundary control system for x, y, & z positions to stop the platform which is in addition , to the existing boundary control . system (Zone Computer) to stop the platform. - l REASON FOR CHANGE: The temporary installation of the Lasertrac system provides an optional operational enhancement by allowing the Lasertrac system to automatically position the platform and mast to any position the operator has commanded. This modification does not bypass or override any existing platform antico111sion, functional, or operational interlocks. In the Semi-Auto or Auto modes of Lasertrac operation, Lasertrac provides additional antico111sion and operational interlocks to the existing system. NLSATTC2/SNLICFLR - 173
g s E Attachment-to AECM-89/0093' c 1 f[ LPLS-89-019 , Page 2-g 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 evaluated:in the Safety
-Analysis. Report. Lasertrac is an addition to the Refueling platform;' therefore, no previously evaluated accident has probabilities of occurrence based on its reliability or function.
The intent of Lasertrac ~is to enhance the operational efficiency.- of the Refueling Platform by automating many routine functions. The number of.. fuel maneuvering decisions required =of the Operator is reduced, and Lasertrac's programmable capability allows better '
- pre planning of the required fuel movements. Lasertrac's additional. zone boundaries will be pre-set to add increased protection against fuel collisions with pool walls / gates, the
" Cattle Chute", or other obstructions. These additional
~ boundaries-are enveloped by the existing Zone Computer boundaries
.which are intended for the same purpose. Lasertrac does not affect or invalidate any of the existing Zone Computer functions, nor does it decrease the reliability of the Zone Computer. The Zone Computer will continue to provide all Technical Specification and UFSAR-required interlock interfaces. The required interlocks do not depend on Lasertrac. Thus, the fuel handling accident as-
, ' described in UFSAR 15.7.4 and 15.7.6 continues to be the worst case scenario for mishandling spent fuel with the Refueling l
Platform. Lasertrac does not affect the cause, sequence of L ' events, or assumptions of the fuel mishandling accident. lc Operational limitations of the Refueling Platform using Lasertrac ? is bounded by the existing design parameters and specifications. All Lasertrac equipment'is non-safety related but was designed and installed to satisfy Seismic II/I requirements. The seismic / structural adequacy of the Refueling Platform is not affected by Lasertrac. Also, all required interlocks and boundaries are provided by the existing, unaffected Zone Computer. The Zone Computer cannot be defeated by the use of Lasertrac; thus, failure of Lasertrac would create an operational situation similar to the existing design without Lasertrac. The possibility of any accident being created with Lasertrac installed would be the same as the existing configuration without Lasertrac. Lasertrac is an addition to the Refueling Platform; therefore, no previously evaluated malfunctions of equipment were postulated based on a probability of Lasertrac failure. The possibility of Lasertrac malfunction would not be important to safety since Lasertrac is intended only as an enhancement to operational efficiency. The use of Lasertrac is optional to the operator, but its use reduces the number of fuel maneuvering decisions for the operator. This, coupled with the capability of Lasertrac to be pre programmed (better planning), will reduce the possibility of operator error. Also, the additional (conservative) zone boundaries of Lasertrac are not required to prevent fuel collision with pool gate, walls, or other obstructions. In case of NLSATTC2/SNLICFLR - 174
.__ ~ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ . _ _ _
Attcchment to AECM-89/0093; pLS-69-019' Page 3 Lasertrac malfunction, the existing Zone Computer boundaries will provide this protection. Therefore, there is no creation of a-
-possibility for an accident or malfunction _of a different type than any evaluated previously=in the Safety Analysis Report.
No margins of safety exist based on the use of Lasertrac since it is an addition to the Refueling Platform. ' Margins of safety defined as the bases of the referenced Technical Specifications are not affected since 1) the referenced Technical' Specifications are not affected by the addition of Lasertrac, (2) the LCOs and surveillance requirements of the referenced. Technical Specifications are not dependent on Lasertrac, 3) Lasertrac'does not replace the existing Refueling Platform Zone Computer for which applicable Technical Specifications are dependent, and 4) the required interlock interfaces will continue to be provided by
-the Zone Computer which is not operationally affected by Lasertrac
-and which cannot be bypassed or overridden by Lasertrac.
1 7 HLSATTC2/SNLICFLR - 175 I
kil httschment to AECM-89/00931
=
yf 'J. i [ 3 h
. SRASN: . pLS-89-020- DOC NO:' MWO-F86672 SYSTEM: P41
{ i b ~ DESCRIPTION OF CHANGE: Special maintenance' instructions were written to1 chemically clean SSW B basin. REASON FOR CHANGE:. Chemical cleaning is necessary to ensure the 7 structural integrity of the SSW B loop. Cleaning removes f corrosion products and aids;in arresting the corrosion process. S SAFETY EVALUATION: There is no increase in the probability.of occurrence or in the consequences of an accident or malfunction of L equipment important to-safety previously evaluated in the Safety Analysis Report. Onsite testing performed prior to the chemical-cleaning of SSW piping has verified the capability of.the process in removing fouling present in the SSW system. The results of F this testing have also shown that-the upper bound maximum expected ! base _ metal-loss due to the cleaning process with respect to the j available margins (i.e., structural adequacy and pressure
.I retaining capability).in the SSW pipes is acceptable. .I Additionally, in response to concerns regarding potential j degradation of weld materials and crovice regions, the test 1; i results evaluation confirmed that there was no evidence of .i preferential-corrosive attack in those regions. Performance of this special. instruction, which incorporates the recomended operating. constraints contained in the test results evaluation, will ensure that actual base metal loss will be minimized. Pre-and post-cleaning inspections of selected system component internal surfaces will be performed to further assure system integrity has not been degraded. The system will not be modified such that it will be left in an unanalyzed condition. The SSW j loop to be chemically cleaned will not be required to be operable during the performance of this special instruction. The system will be functionally tested to assure full system design i capability prior to returning the system tc operable status. ,
Performance of this special instruction will ensure that the l operation of SSW equipment during the cleaning process will be i conducted in an approved and controlled manner. l l There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in i the Safety Analysis Report. The SSW loop to be chemically cleaned will not be required to be operable to meet any Technical Specification requirement during the performance of this special - instruction. Due to the divisional and physical separation of the redundant SSW loops the consequences of any single component malfunction of SSW loop equipment important to safety during the performance of this special instruction will not be increased beyond that previously analyzed in the FSAR. Performance of this special instruction will ensure that the operation of the SSW loop to be chemically cleaned will be conducted in an approved and NLSATTC2/SNLICFLR - 176
f7 . N ^ Attachment to AECM-89/0093
%d PLS-89-020 '
Page.2
< s ,.
controlled-manner. .The SSW pump for the loop to be cleaned will not be operated in a manner inconsistent with normal system operating. requirements. As: state'd in the bases for Technical Specification 3/4.7.1 the redundant cooling capacity of these systems (i.e., SSW loops A B, and C), assuming a single failure, is consistent with assumptions used in the. accident conditions within' acceptable limits. A-
-single failure in any one loop will not affect the ability of the other two loops- to perform their intended function. In this instance, it is assumed that one SSW loop may be declared inoperable in Operational Conditions 4 or 5 without adversely f 'affecting the margin ofosafety provided by the remaining operable
- loops. Therefore the margin of safety as defined in the basis for
'any Technical Specification is not reduced by the performance of this special.. instruction.
l-NLSATTC2/SNLICFLR - 177 L
gm
;a Attachment to AECM-89/0093 j
SRASN:: IPLS-89-021 ~- -000 NO: MWO-F84433 SYSTEM: P41
- DESCRIPTION OF- CHANGEi. This evaluation is' valid for temporary ,
submersible pumps mounted on the basin floor in any SSW basin ; E ' quadrant- -It assumes the pumps are connected.to either flexible or rigid discharge piping.. It further assumes that the temporary submersible pumps are not connected to an electrical power source while: Grand Gulf Nuclear Station-is-in Operational Conditions 1, 2,~or 3, and are placed beneath the basin' cooling tower structure.
'These assumptions are controlled by the work instruction issued to perform the activity.
= REASON FOR CHANGE: This change provides for the removal of J k chemical cleaning waste from the SSW basins.
SAFETY-EVALUATION: .There is no increase in the probability of occurrence or in the consequences of an accident or malfunction of s - equipment important to safety previously evaluated in the Safety (- -Analysis Report. .The installation of portable submersible pumps and associated piping and electrical cable in the SSW B basin is considered to be a maintenance activity. Per Technical 6 Specifications 3/4.7.1.1 and 3.7.1.3, both loops are required to be operable in Operational Conditions 1, 2, and 3, while one loop of'SSW may be declared inoperable in Operational Conditions 4 and 5 without entering a Limiting Condition of Operation action statement. The installation of the portable submersible pumps, associated piping, and electrical cable in SSW basins can.be ' performed with the basin at or above minimum Technical i Specifications water level of 130'-3" without declaring the associated SSW loop inoperable for the following reasons:
- 1) During installation the temporary pumps, piping, and cables L
will be tethered or otherwise captivated to prevent !' uncontrolled entry into the basin.
- 2) Should the captivating mechanism fail the equipment being installed would descend in a near vertical trajectory even if the SSW pump was operating due to the extremely low fluid t
velocities present within the basin during normal SSW operation. The force of the resultant impact of the equipment free f alling through the basin fluid and striking the basin floor is not considered to be a significant hazard to the integrity of the basin structure.
- 3) It is improbable that the smallest item to be installed, the pipe bolting material, if dropped, could migrate to the SSW pump protective screen located above the pump suction due to the extremely low fluid velocities present within the basin when the SSW pump is operating. However, should this occur, the pump suction protective screen would prevent entry into the pump mechanism. The total reduction of flow area available through the screen and the resultant increased hydrodynamic loading on the screen would be negligible even if all bolting material migrated to the screen.
NLSATTC2/SNLICFLR - 178
y ,y 1 M_m Attachment to AECM-89/0093 , ps g . PLS-89-021 s Page 2 La
' 4)- 'Any material which may come to rest in the SSW basins can be retrieved by' divers without declaring the SSW system inoperable, r
- 5) The-structural arrangement of the SSW basin prevents direct i lowering of equipment onto the' pump suction protective 97 ,
screen. The minimum horizontal distance between the Unit 1
/
SSW pump centerline and the edge of the pump suction screen _is approximately 16 feet. The minimum horizontal distance Y between the Unit 1 SSW pump centerline and a point directly ,
'below the nearest installation point for temporary ;
submersible pumps beneath the basin cooling tower structure. ' is approximately 33 feet.
- 6) 'The potential.for the temporary equipment to impact the SSW pump casings and protective-screens is negligible. A 2 foot thick concrete curtain wall extends 12 feet downward (121 foot elevation).around the SSW pumphouse perimeter beneath the basin cover slabs. A 3 foot thick horizontal concrete beam extends across the basin (center line elevation 106
-feet) directly beneath the edge of the basin cover slabs.
The pump suction protective screen is mounted 1 foot, 4 inches above the basin floor and a 3 foot high, 6 inch thick, a- concrete fence is located along the north side of the X . protective screen. Considering the existing basin , interferences, a dimensional stackup of the pump and
~,' equipment during and after installation and a review of the possible trajectories demonstrates that Unit 1 SSW pump casings would not be susceptible to impact should the pump, pipe, and cable assembly topple during or after installation.
- 7) The presence of the temporary submersible pumps and associated equipment in the SSW basins will not affect the L '
normal operation of the Unit 1 SSW system. No degradation of Unit 1 SSW system heat removal capability is created by the placement of this equipment in the SSW basins.
- Based on the assessment above the installation of temporary pumps and associated equipment in the SSW basins does not create a L hazard to Unit 1 SSW system components; therefore, the installation of temporary pumps and associated equipment in the SSW basins per this MWO will not increase the probability of any accident previously evaluated in the FSAR.
L L Based on the assessment above the installation of temporary I submersible pumps and associated equipment in the SSW basins does l not create a hazard to Unit 1 SSW system components; therefore, performance of this MWO will not increase the consequences of any l accident previously evaluated in the FSAR. l l NLSATTC2/SNLICFLR - 179 l
3 -y '
; ,(
Aitachment to AECM-89/0093 w PLS-89-021' Page 3 Based on the assessment above the installation of temporary < submersible pumps ard associated equipment in the SSW basins will not modify-the Unit 1 SSW system such that it will be in an unanalyzed condition. . Therefore, performance of this MWO will not !
- create the possibility of an accident of a different type than any '
previously eye.luated in the FSAR. Based on the' assessment above the Unit 1 SSW system functionality
'is not compromised nor is an unanalyzed hazard created by the presence.of temporary pumps and associated equipment in the SSW basins. Performance of.this MWO will not affect the operation of Unit '1. SSW equipment during or af ter installation .of the temporary pumps and associated equipment in the SSW basins. Therefore, the probability of a malfunction of equipment important to-safety as described in the FSAR will not be increased.
There is no creation of a. possibility for an accident or ; malfunction of a different type than any evaluated previously in the Safety Analysis Report. The Unit 1 SSW system functionality is not compromised nor is an unanalyzed hazard created by the presence of temporary pumps and associated equipment in the SSW basins. Performance of this MWO will not affect the operation of
' Unit 1-SSW equipment during or af ter installation of the temporary
' pumps and associated equipment in the SSW basins. Furthermore, dde to the divisional and physical separation of the redundant SSW loops A, B, and C, the consequences of any single component malfunction of Unit 1 SSW equipment important to safety during the performance cf this MWO will not be increased beyond that previously analyzed in the FSAR.
LThe presence of temporary pumps and associated equipment in the SSW basins do not present a hazard to the Unit 1 SSW system components. The Unit 1 SSW system will be operated in a normal manner-during the performance of this MW0. Therefore, the possibility of a malfunction of a different type than that described in the FSAR is not created for Unit 1 SSW system equipment important to safety by the performance of this MWO. The presence of temporary pumps and associated equipment in the < SSW basins do not present a hazard to the Unit 1 SSW system components. No failure of the temporary equipment installed per this MWO will affect any Unit 1 SSW system component. Furthermore, as stated in the bases for Technical Specification 3/4.7.1, the redundant cooling capacity of these systems (i.e., SSW loops A, B, and C), assuming a single failure, is consistent with assumptions used in the accident conditions within acceptable limits. A single failure in any one loop will not affect the ability of any one of the other two loops to perform its intended function or adversely affect the margin of safety provided by the remaining operable loops. Therefore the margin of safety as defined in tne bases for any Technical Specification is not reduced by the installation of temporary pumps and associated equipment in the SSW basins per this MWO. NLSATTC2/SNLICFLR - 180 , I
c, X '
-Attachment to AECM-89/0093
?
-SRASN: PLS-89-022 DOC NO:- MWO-F91-305 SYSTEM: M10' 1
.1 DESCRIPTION OF CHANGE: An engineered platform will be placed over j a small' portion of the suppression pool for use as a laydown area !
for.11ght objects. It will be located between azimuth 251 degrees 1 and 3 minutes and 297 degrees and 41 minutes. l ' P There ere 3 spargers for the SRVs and a 12" RHR 'B' pump test line
.in the suppression pool immediately under the proposed location for the platform. There'are no ECCS suction' points immediately under the platform. The nearest ECCS suction is for HPCS, which
- is at azimuth 312 degrees. This_is approximately 16 feet circumferencially from the nearest edge of the platform.
A safety railing of standard height will be installed at the edges
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of the platform for personnel safety. The railing will have kickboard to prevent objects from falling into the pool. The railing will be drapped with Herculite. The platform is not in the area of the suppression pool make-up dump lines. REASON FOR CHANGE: This change is made to support RF03. 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 evaluated in the Safety Analysis Report. There is no previously evaluated accident in the FSAR affected by this change. The platform will not be directly or indirectly connected to any plant system and cannot affect the operation of any system. NPE calculation No. CC-Q1M21-89021 was performed to confirm the integrity of the structural steel for the platform. The calculation showed that the steel structure would not fall during a design basis earthquake even if it is fully loaded to the design load of'100 lbs/sqft. The safety rail and kickboard at the ends of the platform over the suppression pool
.should contain any-items stored on the platform. Therefore, the probability of items falling from or with the platform during an :
ea.thquake and subsequently clogging an ECCS suction strainer should be very low. There is no ECCS suction strainer directly beneath the proposed location for the platform. The effects of a pool swell-from a SRV blowdown is not a s considerction since the platform will only be in place when the plant is <200 F. Fire retardant material will be used in the construction of the platform thereby reducing the chances of a fire. Only fire retardant wood and fire retardant Herculite will be used. Additionally, there should be frequent personnel traffic near the platform due to its proximity to the drywell equipment hatch. HP personnel will be covering the drywell and the containment 24 hours a day. Should a fire start, the platform can be reached by NLSATTC2/SNLICFLR - 181
g- 7 V
-Attachment to AECM-89/0093 PLS-89-022 Page 2' <+ the fire hose from the station near the 119' containment personnel y ha tc h.- A fire extinguisher is located near the hatch; also.
L Therefore..there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the. Safety Analysis Report. No technical. specification is based on the proposed platform. . The railing-and kickboard at the edges of the platform should minimize the chance of objects
. falling into the suppression pool and clogging an ECCS suction L strainer. Therefore, no margin of safety in any technical Lspecification should be-reached-as a result of installation of the platform, t l'
i i I i NLSATTC2/SNLICFLR - 182
y , = - x 4 Atttchment to AECM-89/0093-h! E 3: SRASN: 'PLS-89-023 -DOC NO: ,FCR-89-007 SYSTEM: B33 h['O DESCRIPTION OF CHANGE: The FSAR states that the sequence of .m . events'during refueling has the Portable Radiation Shield (Cattle l Chute) installed and removed when.both the dryer and separator are installed in the vessel. Because the Cattle Chute was designed to-
~
overhang the vessel flange, it is prudent to install and remove s the Cattle = Chute with the Dryer removed from the vessel, but with , the, separator still bolted in place. This change will minimite or
- elir.inate possible interference between the Cattle chute and dryer when lifting the-dryer from the vessel.
4
-Extensive reviews of heavy load drops on the refuel floor were
. conducted in response to NUREG 0612. This has resulted in a y series of procedures which regulate the handling of loads on the b refuel floor and ensuring that load drops would not result in. <
. unacceptable consequences.
REASON FOR CHANGE: -This change minimizes or eliminates possible interference between the Cattle Chute and Dryer when lifting the Dryer from the' vessel. L < SAFETY EVALUATION:- There is no increase in the probability of g occurrence or in the consequences of an accident or malfunction of k . equipment-important to safety previously evaluated in the. Safety
. Analysis' Report. The UFSAR and the Heavy Loads Final Report L(AECM-82/149) already themselves assumes a drop of the-Cattle Chute. . Because of increased attention to these heavy lifts (HUREG
'0612),-the' lift is controlled by procedure and specifles the I: ; method of handling and rigging and safe load path which meet the requirements of NUREG 0612. Therefore, this change in sequence b :does not increase the probability'of dropping the Cattle Chute.
-This sequence actually eliminates the possibility of a lifting and rigging interference between the Cattle Chute and Dryer when lifting the Dryer from the vessel. Loads which are rigged and handled in accordance with NUREG 0612 are not postulated to drop.
Although the' analysis in the UFSAR for the Cattle Chute drop assumes that both the dryer and separator are in place, subsequent analysis of this drop with only the separator in place (AECM-82/149, Heavy Loads Final Report) has verified the consequences to be bounded by the RPV head drop and dryer drop analysis. Thus, this drop is bounded by UFSAR analysis. Additionally, this report finds that the guidelines of HUREG 0612 are met. The installation and removal process has not been altered, only the sequence in which the Cattle Chute is installed and removed, thus no new types of accidents or failure modes are introduced. The process of installation / removal provides controls which limit / prevent the load in unacceptable areas of 208. Because of the NLSATTC2/SN'LICFLR - 183
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( g, Attachment to AECM-89/0093 L fi PLS-89-023 i Page 2
. potential interference between Dryer and Cattle Chute, this
- s. .
sequence is better in that it eliminated the potential interference. The Cattle Chute is classified as a Class 2/3A load in accordance
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e . with NUREG 0612.. As such, the rigging and handling on the Cattle '
; Chute is controlled by procedure. This change in sequence does not' affect.loadpath, nor handling method of the Cattle Chute.
Therefore, neither the Cattle Chute or safety systems will I experience any increase in malfunction- probability. This change in sequence does not expose any new safety systems to' the potential of a load drop nor does this change increase the height of load lift and increase the impact energy that a drop may have. Thus, no new malfunctions are created by changing the sequence.' However, potential interference.between the Dryer and Cattle Chute are avoided by making this change. Additionally, procedures prohibit the Cattle Chute from being carried directly over the vessel when installing end removing it, .Therefore, there - is- no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. . The final heavy load analysis shows no fuel damage is predicted, if the load were dropped, there are no new safety systems exposed to a load drop, there is no increased potential of a load drop and the load is controlled in accordance with NUREG 0612 guidelines, based upon this there is no decrease in the margin of safety as defined by Technical Specifications. l NLSATTC2/SNLICFLR - 184
w ' r-Attachment to AECM-89/0093-hl r $'= SRASN: PLS-89-024 DOC NO: MWO-91577 SYSTEM: 1E22 DESCRIPTION OF CHANGE: Special instructions to MWO 91577 provide
'a flush path for the LPCI "C" injection line. This flow path utilizes the existing test return mode of RHR "C", establishing a
- .. ; -flow rate of approximately 6500 GPM, opens the 8.PCI "C" injection valve (stroke time approximately 25 seconds), closes the LPCI "C" injection valve, and then secures RHR "C" from the test return . !
mode- . This will result in a flow, rate to the vessel of less than 2500 GPM with an accompanying increase in vessel inventory of about 2500 gallons. REASON FOR CHANGE: This flush is necessary to avoid potential r water visibility problems that could affect the ability to-perform core alterations because of vessel to vessel operation of the Alternate Decay Heat Removal System during RF03. 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 evaluated in the Safety Analysis Report. Performance of this MWO utilizes existing fiow paths for. the. RHR "C" system. The flow to the vessel will result in an increase in vessel inventory, however the special instructions will provide precautionary measures to terminate this E increase prior to reaching a level of 90" to preclude water intrusion into the Main Steam lines. Performance of this MWO will-r4 jeopardize the integrity of the RHR "C" system. The RHR "C" s will merely be operated in a manner consistent with its c design capabilities. 1 D., 3 performance of this MWO the RHR "C" system will be . rendered inoperable under-the provisions of Technical Specification (T.S..) 3/4.5.2. As such sufficient ECCS availability.is ensured to mitigate the consequences of accidents previously evaluated in Chapter 15 of the UFSAR. Since this MWO' will be performed at Cold Shutdown conditions and due to the nature of.the events / accidents described in Chapter 15 of the UFSAR, the consequences of an accident are not greater than previously evaluated in the UFSAR. , There.is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. During performance of this MWO the RHR "C" system is operated under conditions which are allowed by the. system design. No unanalyzed loads or forces are applied to the system components. The RHR "C" system is not subjected to any forces, pressures or other conditions to which it is not designed or that are not expected to occur within its design life. LPCI "C" flow baffle fatigue stress analysis has been previously evaluated for the ADHRS design and at flow rates less than 4500 gpm it was determined that the increase in usage factor was negligible. NLSATTC2/SNLICFLR - 185
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!. E By remaining within' the constraints imposed by the Technical. . >
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* 's , ' margin.of. safety ;is not compromised for any GGNS Technical-. . ;:1 . Spec t fication. .
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Attachment to AECM-89/0093 a, p, e SRASN: PLS-89-025 000 NO: 07-5-14-339 SYSTEM: FIS i DESCRIPTION OF CHANGE: This 10CrR50.5) evaluation covers the use i of a 20 ft. 4-wire rope sling with each cable having a breaking ' strength of 15,000 lbs. (minimum) for use in lif ting the reactor t vessel level servicing platform instead of utilizing the ; dryer /septrator strengback. This alternate rigging method of ! lifting the platform will not contribute to the spread of contamination as with the use of the strongback and will provide : the same function as the strongback with no adverse affect on the ! vessel platform by any rigging induced stresses, t
' i REASON FOR CHANGE: This change reduces the opportunity for the j spread of contamination. ,
v > L' 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 evaluated in the Safety Analysis Report. The probability of occurrence of an accident i caused by the use of this alternate method of lifting the service platform remains the same since this new rigging technique maintains the same design criteria as that presently approved for , use. The new rigging will also conform to NUREG 0612 ,
' requirements.
This alternate method of rigging the reactor vessel service # platform coes not increase the consequences of an accident , previously evaluated in the FSAR because the analysis shows it to . be within the design allowables, thus, not introducing any new ' variables. ! This alternate method of rigging the reactor vessel service f platform does not create the possibility of an accident of a' different type than any evaluated in the FSAR in that no modifications to the service platform or operational philosophy i has been changed.
- The probability of occurrence of a malfunction of equipmont important to safety previously evaluated in the FSAR is unchanged !
since this alternate method of rigging the reactor vessel service 4 L platform has been analyzed to meet the same design criteria as that of the presently approved method. l The increase of consequence of a malfunction of equipment
-important to safety previously evaluated in the FSAR remains unchanged since in either method of rigging, the reactor vessel service platform potentially affects the same equipment.
i The possibility of creating a malfunction of a different type than ; previously evaluated in the FSAR remains unchangeci. This alternate method of rigging the reactor vessel service platform meets the same designed safety factors as the presently approved 7 method stated in the FSAR. i e NLSATTC2/SNLICFLR - 187 4 :
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Attach ent to AECM-89/0093 1 4- 1 .; 9 '
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PLS-89-025 1
- Page 2: l 1
The method of rigging the. reactor vessel service platform is not :
- addressed in the GGNS Technical Specifications, therefore, no 1 change:to the margin of safety as defined in.the basis for the ij
,.. , . Technical Specifications exists,- j t i
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e _. - I' Attachment to AECM-89/0093 SRASN: PLS-89-026 DOC NO: TEMP ALT SYSTEM: R61 DESCRIPTION OF CHANGE: This change moves Public Address (PA)
- receptacle junction boxes (1604) to PA stations with speakers L
(1015) in the following rooms: 1A104, 1A207,-1A205, IA203, 1A208, 1A220, 1A308, 1A320, IA319, and i 1A309. REASON FOR CHANGE: This change greatly improves and enhances plant commur.ication and could prevent an accident. " 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 evaluated in the Safety Analysis Report. Changing the type or adding new PA stations has no effeet on accidents previously evaluated in the FSAR. All existing stations will function as described. The new stations will function as those previously installed and integrate into the system with no offect on the system except to hear in some areas previously void of PA communications. The consequences of an accident remains the same. The PA system will function the same and as designed. This only converts these designated PA junction boxes to PA stations that are installed throughout the plant. The probability of an accident remains the same because the system, it's cabling and it's function remains the same. The equipment installed meets all designs specifications of the existing system. It has no direct interface with any equipment other than the PA system itself. The PA does furnish power and page lines for the evacuation alarm and rotating beacon system. This change will have no effect on this. No equipment important to safety is affected except for the PA and this improves it's capability using it's existing design to give expanded coverage. There is no new or increased loads to UPS, no other system is affected. The PA stations are not used in the basis for any Technical Specification. This alteration will not change this. It will i improve the communications at these new locations during normal and everyday plant conditions. 1 NLSATTC2/SNLICFLR - 189 1
Attachmnt to AECM-89/0093 ; l IL SRASN:' PLS-89-027 000 NO: MWO-E91704 SYSTEM: R61 i , L Investigation under MWO E91704 verified DESCRIPTION OF CHANGE: ! that the wires for Sound Powered Communications to the Containment ! Aux Service platform and 480 VAC to that platform were in the same cable. The power cable wires are inducing noise into the Sound l , Power Communications lines causing a disrupt in communications on ! this loop. This MWO separates the power wires (480VAC) from the Sound Powered Communications wires. l L ! REASON FOR CHANGE: This change is to prevent further disruption ' of the Sound _ Power Communications by the power cable wires. SAFETY EVALUATION: There is no increase in the probability of 3 occurrence or in the consequences of an accident or malfunction of i equipment important to safety previously evaluated in the Safety Analysis Report. There is no ar,cident previously evaluated in the FSAR that is affected by this change. The plant SP system is not connected to any other plant system and requires no power source. The 480VAC will remain in conduit and meet all design intent as ; described in the FSAR. The SP wire will be tie-wrapped to the same conduit but will no longer run inside next to the 480 vac. The consequences of an accident remains unchanged. It eliminates the chance of the 480 lines and the SP lines shorting and is an i improvement to the present configuration. It will also provide , clearer and better communications between the Aux platform and the Control Room lessening the consequences of an accident when this communications is required. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. No basis for any Technical Specification is based on the SP ' system. B 3/4 9.5 states "The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or , core reactivity conditions during movement of fuel in the Reactor ' Pressure Vessel". This change improves communications between the Aux platform in containment and the Control Room. No margin of safety in any Technical Specification changes because of this 1 change. NLSATTC2/SNLICFLR - 190 E---
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Atttchment to AECM-89/0093 SRASN: PLS-89-028 DOC NO: S.P. 06-ME-1M10-0-0002 Rev. 22 SYSTEM: M10 DESCRIPTION OF CHANGE: Revision 22 of the Containment Integrated Leak Rate Test Procedure, Surveillance Procedure 06-ME-1M10-0-0002, specifies several conditions that deviate from the Final Safety Analysis Report (FSAR) and referenced piping and instrumentation drawings (P& ids), which are also contained in the FSAR. These changes and conditions are the following:
- 1. Several containment penetrations which are indicated in FSAR Table 6.2-49 as being tested during the Type A test will not be vented and drained for the ILRT as specified in FSAR Paragraphs 6.2.6.1.c and d. These penetrations include the following:
A. penetrations for systems which are required to,be operable to maintain the plant in a safe condition during the ILRT. 10 CFR 50, Appendix J specifies (Para. III.A.1.(d)) that, " Systems that are required to maintain the plant in a safe condition during the test shall be operable in their normal mode, and need not be vented." The following systems and penetrations, which are specified in FSAR Tablo 6.2-49 as being tested during the ILRT, are not vented or drained because they are required to maintain the plant in a safe condition for the following reasons: (1) Penetration 56 (Condensate makeup to uppee containment pool) remains in service to provide fire protection water to spray nozzles in containment. The penetration and piping in containment are designed to remain water-filled during normal operation and shutdown. Because access to the containment by the Fire Brigade is greatly hampered by the containment being pressurized during the ILRT, the availability of the fire suppression system is critical to containing the spread of a fire. l l (2) Penetration 57 (Fuel pool cooling and cleanup l system dischargo to upper containment pool) is not vented or drained for the ILRT because it cannot be vented or drained on the Auxiliary Building side without interrupting Fuel Pool Cooling water flow to the Auxiliary Building Spent Fuel Pool. Maintaining cooling water flow to the Auxiliary Building Spent Fuel Pool decreases the probability of allowing the spent fuel to overheat due to lack of adequate cooling water. NLSATTC2/SNLICFLR - 191
Att chment to AECM-89/0093
- t. l PLS-89-028 page 2 (3) Penetrations 105A, 106A, B & E, 107B, D & E and t
108A (Containment and Drywell hydrogen analyzer sample and return) remain in service to provide sampling capability to detect hydrogen in containment during the ILRT. Although the hydrogen [
- i. analyzers are not required to be operable during !
L the ILRT, increasing the pressure in containment * !- increases.the density of oxygen in the air, and any * [ fires which might start in containment would burn i
._ faster because of the increased oxygeri density. By i maintaining the hydrogen analyzer sampling penetrations in service, the ability to detect significant quantities of hydrogen in the containment is maximized, thereby decreasing the probability of occurrence of a buildep of hydrogen 11 to an explosive level in containment.
(4) Penetrations 113 through 120 (Suppression pool level instruments) remain in service to provide suppression pool level indication to the control room. The ability to detect the change in level would allow the control room personnel to take corrective action before any unexpected increase or decrease in pool level reached a level that could damage safety-related equipment or cause an accident. In accordance with the requirements of ANSI /ANS 56.8 -
~
1987 (Section 3.2.1.5), the leakage for these penetrations is determined by local leak rate testing methods and the overell penetration leakage, calculated in accordance with minimum pathway leakage methodology, is added to the 95% upper confidence limit (UCL) of the ; calculated Type A test leakage rate as a penalty.
-B. Penetrations for systems that are designed to be operable under post-accident ccnditions. 10 CFR 50, Appendix J specifies (Para. III.A.1.(d)) that, " Systems that are normally filled with water and operating under post-accident conditions, such as the containment heat removal system, need not be vented." Similar words can be found in FSAR Section 6.2.6.1.f. The penetrations listed below are penetrations for safety-related emergency core cooling systems (ECCS) which are designed for and expected to operate under post-accident conditions to maintain the reactor vessel flooded and the nuclear fuel cooled.
(1) Penetration 18 - RHR to RPV head spray (2) Penetration 20 - RHR Loop A Low Pressure Core Injection NLSATTC2/SNLICFLR - 192
(~n s Attcchment t3 AECM-89/0093- f i PLS-B9-028 Page 3 , (3) Ponetration 21 - RHR Loop B Low Pressure Core Injection (4). Penetration 22 - RHR Loop C Low Pressure Core- ! Injection j [ - (5) Penetration 26 - HPCS pump discharge to RPV (6) Penetration 31 - LPCS pump discharge to RPV [
'In accordance with the provisions of ANSI /ANS 56.8 - ;
1987 (Section 3.2.1.5), the leakage for these i penetrations is determined by local leak rate testing ! E methods and the overall penetration leakage, calculated ,
,in accordance with minimum pathway leakage methodology, .:
is added to the 951b upper confidence limit (UCL) of the l calculated Type A test leakage rate as a penalty. : C. The main steam lines (Penetrations 5, 6, 7 and 8) may- , l remain isolated during the ILRT due to outage ! scheduling. Because the reactor vessel is scheduled to l
-be filled to the top of the refueling pool at the t scheduled window for the ILRT, either the main steam lines would have to be water-filled during the ILRT or ,I the main steam line plugs would have to remain in place ,
L with seals inflated during the ILRT. -Per discussions L with Region II it.would be better to have the main steam 'i lines water-filled and add the local leak rate tess leakage to the.ILRT result as a penalty rather than to ~! have a pressurized air or nitrogen bottle in containment , during the ILRT to keep the main stcam line plug seals [ inflated. 2.. The Type A test is conducted in accordance with the acceptance criteria of'Bechtel Topical Report BN-TOP-1, Rev. 1, " Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants," an' alternate testing method approved by the Nuclear , Regulatory Commission (NRC) and which is in compliance with the provisions of ANSI N45.4. The Atomic Energy Commission's . (AEC's) evaluation of the topical report concludes that > BN-TOP-1" is acceptable for referencing in Safety Analysis , Reports," and the AEC's cover letter concludes that BN-TOP-1 is " acceptable by reference in applications for construction- 'l permits and operating licenses." Bechtel Topical Report BN-TOP-1 is currently the only NRC-approved method for ! performing Type A tests with durations less than the 24 hours specified in ANSI N45.4. BN-TOP-1 requires a minimum t L duration of 8 hours for the Type A test. . h l NLSATTC2/SNLICFLR - 193 e + - , _
U
-Att chment to AECM-89/0093 j l ^
PLS-89-028. l Page 4. i 1 ;
.3. The Type A test is not conducted using the methods specified ;
in ANSI Standard N274, " Containment System Leakage Testing Requirements," as specified in FSAR Section 6.2.b.1. ANSI W N274 was a draft standard which was superseded by ANSI /ANS 56.8 - 1981 and is now obsolete. .The Type A test will be l conducted in accordance with the methods, proceduresi and i requirements of Bechtel Topical Report BN-TOP-1-(as discussed ; D in Item 2 above), ANS H45.4 - 1972, " Leakage-Rate Testing of ,! h' Containment Structures for Nuclear Reactors" (as required by i GGNS Technical Specification 4.6.1.2) and ANSI /ANS 56.8 - ! g 1987, " Containment System Leakage Testing Requirements,". , i; 4.- The quantities and types of sensors to be used are not in '! [L accordance with the descriptions contained in FSAR Table i J 6.2-49, and the placement of the temperature and dewpoint . sensors will not be in the locations specified in P&ID . q", . M-1111A (FSAR Figure 6.2-76). Also, the numbers assigned to j the sensors in the revision do not match those in FSAR Table'6.2-48, although they do match those in P&ID M-1111A. p The sensors used in the ILRT are placed in locations p calculated to represent actual atmospheric air conditions in b the containment so that the total. weight of dry air in the containment can be determined as closely as practical. .The
,_. sensors may be moved depending on the heat loads in various P parts of the containment and the movement may be based on the results of previous tests, actual temperature measurements in the containment, or experience with ILRTs at other nuclear L plants. The latest sensor locations were calculated by Bechtel power Corp. using the results from the 1985 ILRT at GGNS and Bechtel's experience with ILRTs at other plants.
7 i
- 5. The criteria for determining that the containment air is stabilized are different from those given in FSAR paragraph 6.2.6.1.b. The criteria given in the FSAR specify the following:
Containment atmosphere is considered stabilized when the average rate of change of air temperature (weighted average) over the last hour does not deviate by more than 0.5 F/hr from the average rate of change of air temperature over the last three hours." ANSI /ANS 56.8 - 1987 (para. 5.3.1.3) specifies that the containment air is considered stabilized when "the latest rate of change of the containment atmosphere volume weighted absolute drybulb temperature, averaged over the last hour does not deviate by more than 0.5 *F/hr (0.3 *C/hr) from the j l !. i L
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- NLSATTC2/SNLICFLR - 194 I
h A
h Attachment to AECM-89/0093 PLS-89-028 $ Page 5 average rate of change of the containment atmosphere volume weighted absolute drybulb temperature averaged over the last . fmLt hours."
- 6. The range of the imposed leakage rate during the verification test is different from the range specified in FSAR Table 6.2-47, Item D, which specifies that the superimposed leakage rate be 50% to 100% of L GGNS Technical Specification i 4.6.l.2.c.3 requires that. the superimposed leakage rate be 75% to 125% of L,.
REASON FOR CHANGE: For most of the conditions, the FSAR has not been updated to incorporate past changes in the plant, standards, procedures and methods and, therefore, needs to be revised. This evaluation makes these needed changes. SAFETY EVALUATION: Chances 1.A and 1.B: 10 CFR 50, Appendix J, specifically allows l penetrations such as these to remain unvented and water-filled, as > noted under Descriotion. In addition, the leakage for these penetraticas is determined by local leak rate testing methods in accordance with Appendix J, and the leakage is added back to the Type A test results as a penalty using minimum pathway leakage j methodology. All of these penetrations remain in, or are restored to, their normal operational lineup before the plant is started up. Since these practices are in accordance with regulatory requirements-and represent a conservative assessment of containment leakage rates, leaving these penetrations unvented and water-filled during the ILRT will not increase the probability of , occurrence of an accident previously analyzed in the FSAR. Chance 1.C: As indicated above, this change complies with the intent of 10 CFR 50, Appendix J. In addition, the leakage for ; these penetrations is determined by local leak rate testing methods in accordance with Appendix J, and the leakage is added back to the Type A test results as a penalty using minimum pathway leakage methodology. All of these penetrations remain in, or are , restored to, their normal operational lineup before the plant is started up. Since these practices are in accordance with > regulatory requirements and represent a conservative assessment of containment leakage rates, leavirg these penetrations unvented and water-filled during the ILRT will not increase the probability of occurrence of an accident previously. analyzed in the FSAR. NLSATTC2/SNLICFLR - 195 .
<e Attach:ent to AECM 89/0093- f i PLS-89-028 Page 6 ! Chance 2: Bechtel Topical Report BN-TOP-1 provides a conservative l assessment of containment leakage since additional conservatism was written into the procedures in order to justify the shorter ,
-minimum test duration. Specific details of the conservatism are i discussed in BN-TOP-1. BN-TOP-1 has been approved by the NRC and ;
is an accepted conservative methodology for performing ILRTs. It is a valid, conservative test which, therefore, will not increase ; the probability of occurrence of an accident previously analyzed in the FSAR. Chance 3: ANSI /ANS 56.8 - 1987, which is a replacement standard ' for ANSI N274,-is a generally recognized industry standard for , performing leakage testing on primary containment buildings at nuclear plants. It was developed by recognized experts in the , field, including representatives from the NRC. The methodology and requirements in ANSI /ANS 56.8 - 1987 comply with current
. requirements and interpretations of 10 CFR 50, Appendix J, and ANSI N45.4 - 1972. Containment integrity is assured by performing -
the ILRT in accordance with all applicable requirements of Appendix J and other standards approved by the NRC. Chance 4: The instrumentation to be installed in the containment
- at GGNS to measure temperature, dew point, pressure and flow have specifications that are in line with generally accepted standards (e.g., ANSI /ANS 56.8 - 1987) of accuracy, range, and repeatability for instrumentation used in containment integrated leak rate tests :
and are checked for acceptability using the Instrumentation Selection Guide calculations in Appendix 0 of ANSI /ANS 56.8 - 1987. Therefore, the use of instrumentation not specifically as described in the FSAR will not introduce errors which would allow the containment building to be subjected to pressures or temperatures not previously analyzed in the FSAR. In addition, the integrated leakage rate calculated with these instruments will be in accordance with the accuracies generally accepted by the NRC and the nuclear power industry, and the possibility of accepting a leakage rate that is greater than the allowable. leakage of Technical Specification 3.6.1.2.a is minimized. The assignment of numbers to the ILRT instrumentation is editorial and administrative in nature and has no effect on the technic:1 basis for any components, system or accident analyzed in the FSAR. Chanae 5: The only significant difference between the stabilization criteria in the FSAR and those in ANSI /ANS 56.8 - 1987 is the length of time over which the temperature must be averaged. The FSAR contains approximately the same wording as the
-NLSATTC2/SNLICFLR - 196
Attachment to AECM-89/0093 j i L PLS-89-028 Page 7 ANSI /ANS 56.8 paragraph, except that the temperature change is averaged over 12Mr hours, instead of three hours. ; c ; i Averaging ou r four hours gives greater confidence in the stability of the containment air than averaging over three hours. Since four hours is likely to be a harder length of tinie to average over than three hours, this is conservative. Chanae 6: The purpose of the verification test is to verify the [ ability of the instrumentation and calculations to accurately , determine the overall leakage rate during the Type A test. The ; i ' exact range of the superimposed leakage rate is not significant. Since the verification test leakage rate complies with the
-requirements of the GGNS Technical Specifications, containment integrity is assured. -
Therefore, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. In addition, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated
.previously in the Safety Analysis Report. Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification, o
l NLSATTC2/SNLICFLR - 197 l
\
T[ , Attachment to AECM-89/0093 [ L r SRASN: PLS-89-029 DOC NO: TEMP ALT 89-0002 SYSTEM: R28 ; DESCRIPTION OF CHANGE: This Temporary Alteration expands the [ scope of equipment removal to support the Aux Steam System ' disassembly as previously described in Safety Evaluation #005/89. : This expanded scope of work includes the removal of florescent lighting and receptacles (R28), the public address system (R61) suspended from the structural steel platform at Elev. 142'-0" in . the Water Treatment Bldg, and also, the removal of the structural i steel platform. REASON FOR CHANGE: To facilitate the removal of the Aux Steam boilers and the deaerators from the Water Treatment Building, v SAFETY EVALUATION: The removal of the lighting components, PA i: system components and structural steel platform do not increase , the probability of occurrences of an accident previously evaluated. These components were installed primarily to support the Aux Steam System in the Water Treatment Bldg, since the Aux Steam System in this area is being deleted per Temporary Alteration #89-0002 Ref. Safety Evaluation #005/89, these support components also must be deleted. ' An increase in the consequences of an accident previously evaluated will not be affected since the lighting and the structural steel platform are used to support the Aux Steam System. Since the Aux Steam System has been deleted in the Water Treatment Bldg these components are no longer required. The public address system components deleted will not affect nor degrade communications. In the event of an area emergency, additional public address. components are located within 40 feet of the component being deleted. With the removal of the lighting components, PA system components and structural steel platform, the possibility of an accident of a ' different type, as evaluated in the UFSAR, has not changed. The function of these components are in support of the Aux Steam System within the Water Treatment Bldg only, which is being , deleted, and do not interfere with other plant equipment. The structural steel platform is an independent structure and does not affect the integrity of the Water Treatment Bldg seismic loading. The probability of occurrence of a malfunction of equipment important the safety will not be prevalent because the removal of the lighting components, the public address system components and the structural steel platform in the Water Treatment Bldg were used for the support of the Aux Steam System which has been deleted per Temporary Alteration #89-0002 Ref. Safety Evaluation >
#005/89, and this system was not important to safety.
Because the lighting components, public address system components and structural steel platform used to support the Aux Steam System NLSATTC2/SNLICFLR - 198
7 Attachment to AECM-89/0093 PLS-89-029 Page 2 operation in the Water Treatment Bldg are being deleted, an
- ncrease of any. equipment to malfunction, important to safety, will not be affected.
The possibility of a malfunction of a different type has not changed with the removal of the lighting components, public address system components and structural steel platform since their use is to support the Aux Steam System and this system is. being deleted by Temporary Alteration #89-0002. The margin of safety is not reduced for the basis of any Technical L Specification by the removal of the lighting components, public address system components and structural steel platform, because its use is no support the Aux Steam System which is being deleted by Temporary Alteration #89-0002.
-1 1
l 5-I ' i i NLSATTC2/SNLICFLR - 199 1 i
Attcchment to AECM-89/0093
, lSRASN: 1 PLS-89-031 DOC NO: MSTI-1P64-88-001-0-S SYSTEM: P64 .
' DESCRIPTION OF CHANGE: This design installs an Automan II-C .
control panel (N1P64D006D) in Fire 2one OC402 to provide remote [ . manual actuation of the main halon bank protecting the BOP i Computer and Control Panel Room. This DCP 85/4051, Rev. 1 installs a pressure switch PSN589 located in pneumatic actuation piping. The pressure switch provides.the required auxiliary trip i functions (i.e., closing of fire dampers, tripping of HVAC fan + motors) upon manual actuation of the halon fire suppression y system. ( REASON FOR CHANGE: Modification Special Test Instruction (MSTI) l 1P64-08-001-0-S purpose is to perform a functional test for DCP !
, 85/4051, Rev. l'and a flow check to ensure the header, piping and nozzles to OC403 are not blocked. :
SAFETY EVALUATION: Section 9.5.1.2.2.5 of the UFSAR provides an explanation of system operation for the Halon 1301 System. ( Performing MSTI 1P64-88-001-0-S will not increase the probability .; of occurrence of any failure mode listed in the UFSAR. Precaution ' 4.1 of MSTI 88-001-0-S indic'ates that with the system inoperable, !
, a continuous fire watch with backup fire suppression equipment for j the unprotected areas is to be established within one hour (i.e., -
Technical Specification 3.7.6.4.a). Therefore, the probability of occurrence of an accident previously evaluated in the FSAR will " not be increased. - The basic function of the equipment will remain the same. Neither the sequence of events nor the consequences of failure of the
- Halon fire suppression system for the Computer and Control Panel. '[
Room will be changed by performing MSTI 1P64-88-001-0-S. Technical Specification requirement 3.7.6.4 is established during this test. { MSTI IP64-88-001-0-S helps ensure that the design is installed and functions correctly. Performance of the system is consistent with original design and vendor recommendations. The operation of
.' safety related equipment will not be affected by the performance of MSTI 1P64-88-001-0-S because Technical Specification requirements are incorporated.
i Operation of equipment important to safety as previously evaluated in the PSAR will not be affected. The design change has no new 't interface with equipment important to safety. When this functional test for DCP verification is performed, Technical Specification fire watch requirements are established. All system affected by DCP-85/4051 and MSTI 1P64-88-001-0-S remains bounded by previous safety analysis reviews. No system components are expected to operate outside of design or Technical Specification limits. Technical Specification requirements are , followed when performing this test. NLSATTC2/SNLICFLR - 200 p-
r, ,
-0 Attcchment'_to AECM-89/0093
-n .
,0 i PLS-89-031
',' Page 2 !
)
)
DCP-85/4051 is non-safety related and no new malfunction for ! equipment important to safety will be created. The functional 5 test MSTI 1P64-88-001-0-S will not create the possibility of a j
-malfunction of equipment important to safety., Technical -j Specification requirements are included in MSTI 1P64-88-001-0-S. j I Performance of functional test for DCP-85/4051 ensures correct operation. The design meets the applicable fire protection j standards and system specifications.- This test in'no way reduces- '
the margin of safety as defined in the basis for any Technical ; Specifications. i I L-t i. r l-5 i f I b k r i r
~NLSATTC2/SNLICFLR - 201 t
. . - . < v
yp. Attcchment to AECM-89/0093 i [ p E p' SRASN: . NSP-89-001 DOC NO: .ANF-1.3 Reload Fuel SYSTEM: N/A E DESCRIPTION OF CHANGE: Advanced Nuclear Fuels Corporation has
. designed and is fabricating reload fuel for GGNS-1 Cycle 4 (reload fuel batch ANF-1.3). The fuel is scheduled to begin arriving on site in January 1989. As Cycle 4 is scheduled to begin on .
[ , approximately May 1, 1989 the ANF-1.3 fuel will be stored in the 1 [ new fuel vault and/or spent fuel pool until the start of refueling h outage 3 (RF03). The storage and handling activities comprise:
, a) The storage of ANF-1.3 fresh reload fuel in the new fuel
[ vault, . i k b). The storage of ANF-1.3 fresh reload fuel in the spent fuel L pool ~and, p c) The movement of new fuel to either of these temporary storage locations.
- The reload fuel and the storage and handling activities are
!. described further below. The reload fuel of batch ANF-1.3-is I composed of two different fuel bundle types. These include the conventional 0x8 design (272 bundles for batch ANF-1.3) and 4 6
. bundles of a 9x9-5 design. The Bx8 bundle design has an average >
enrichment of 3.61 weight percent (w/o) U-235 in the enriched zone. The enriched zone contains eight burnable poison rods with .; 4.0 w/o Gd203 in the top 24 inches and eight burnable poison rods with 5.0 w/o Gd203 in the bottom 114 inches. The 9x9-5 bundle : design contains an average enrichment of 3.47 w/o in the enriched ; zone. The 9x9-5 bundle enriched zone contains eight burnable poison rods with 5.0 w/o Gd203 in the top 24 inches and eight
- burnable poison rods with 6.0 w/o Gd203 in the bottom 114 inches. *
-Both bundle designs feature 6 inches of natural Uranium at the top ;
I and bottom of the bundles. i Storage of ANF-1.3 Fresh Reload Fuel in the New Fuel Vault
.Section 9.1.1 of the GGNS UFSAR describes the design features and safety evaluation of the new fuel storage vault. This section of the UFSAR was reviewed for possible impact resulting from the storage of ANF-1.3 reload fuel. The new fuel storage vault racks ;
were designed with spacing between fuel bundles such that K-effective remains below 0.95 under both normal and abnormal . conditions. The UFSAR does not contain the detailed basis for, or the assumptions used in, the supporting analysis performed by GE. t This information was provided separately. The acceptance criterion used to ensure that the design basis is maintained in the storage racks is that the maximum exposure dependent reactivity of a fuel bundle (K-infinity) must not exceed 1.31, calculated for an infinite array of uncontrolled fuel in the ' reactor core geometry at 65 degrees Celsius. The K-infinity determined for establishing this basis is independent of enrichment and specifles a maximum allowable lattice infinite multiplication factor as calculated by standard GE lattico physics , methods. This multiplication factor includes calculational biases and uncertainties.
--NLSATTC2/SNLICFLR - 202
Attcchment t3 AECM-89/0093 p E .. b NSp-89-001 [ Page 2 e The maximum exposure dependent reactivity for reload fuel batch ANF-1.3 has been calculated by MSU System Services Inc. (SSI).
. The SSI analysis considered both the Bx8 and 9x9-5 designs to be ; used for Cycle 4 and was Ferformed for the same conditions as j those described in LT-58, dated 3/17/83, from B. J. Erbes, GE, to J. G. Cesare, SERI, " Grand Gulf Fuel Storage Racks".
The SSI lattice physics methods have been benchmarked against the GE methodology for various fuel designs. Based upon this b ' benchmark, a calculational bias and uncertainty (954 probability at the 95% confidence level) was established between the GE and [( SSI lattice physics methods.- The SSI lattice physics methods were applied to the ANF-1,3 fuel designs. These results were adjusted for the methodology bias and uncertainty previously established. L The SSI-calculated maximum in-core reactivity for the Bx8 and i= 9x9-5 designs was calculated to be 1.25 (K-infinity). This result includes the methodology bias and uncertainties previously described and,'therefore, is directly comparable to the L GE-established acceptance. criterion. This value is below the GE acceptance criterion of 1.31 (K-infinity). Therefore, the ANF-1.3 reload fuel is bounded by the GE licensing analysis and the ability to maintain the K-effective in the new fuel vault below 0.95 is ensured. t Storage of ANF-1.3 Fresh Reload Fuel in the Spent Fuel pool Section 9.1.2 of the GGNS UFSAR describes the design features and safety evaluation of-the spent fuel storage racks. The licensing basis described in the UFSAR is based on a criticality safety analysis of a uniformly enriched 8x8 fuel bundle containing 3.5 weight percent (w/o) U-235. Since the reload fuel batch ANF-1.3 design contains higher enriched (3.61 w/o U-235) 8x8 fuel and a new 9x9-5 fuel design (4 bundles), the criticality safety analyses were revised and submitted to the NRC-(Letter AECM-88/0206, and Letter AECM-88/0228). The NRC recently approved a revision to the licensing basis for the spent fuel storage racks to allow storage of the Cycle 4 8x8 reload fuel and the 9x9-5 LTAs (Letter, L. L. Kitner, USHRC, to W. T. Cottle, SER1, " Grand Gulf Nuclear Station, Unit 1 - Criticality Analysis For Cycle 4 Fuel). Movement of an ANP-1.3 Fresh Reload Fuel Bundle Section 9.1.4 of the GGNS UFSAR describes the fuel handling system which will be used during the processing of reload batch ANF-1.3 prior to the initiation of RF03. The applicability of the UFSAR description to ANF-1.3 fuel is based on a comparison of the physical parameters of the ANF-1.3 fuel bundles to those described in UFSAR Chapter 4 or Current Cycle Safety Analysis (CCSA) report ANF-87-67, " Grand Gulf Unit 1 Cycle 3 Reload Analysis". This NLSATTC2/SNLICFLR - 203 s
Attcchment to AECM-89/0093 t NSP-89-001 ; Page 3 ; comparison (provided below in tabular form) shows that the ANF and i GE fuel bundles have similar geometries and weights. Puel .Puel Channeled Bundle Bundle Design Vendor Weight Side Height 8x8 GE 699 lbs. 5.215 in 176.000 in
- 8x8 ANF 690 lbs. 5.188 in 176.044 in i 9x9-5 ANF 690 lbs 5.188 in 176.014 in The limiting event for a fuel handling accident is described in UFSAR Section 15.7.4 as the drop of a channeled irradiated spent fuel assembly onto stored irradiated spent fuel bundles. These irradiated bundlea provide the source term for the radiological consequences evaluation.
A key parameter which affects the radiological consequences of this event is the number of fuel rod failures resulting from the drop. UFSAR Section 15.7.4.3 describes this calculation and the results. A total of 101 rod failures were calculated; 62 rods would be from the dropped bundle and 39 rods from the stored bundle. Based on the similarity of the fuel designs outlined , above and the calculational methods used in UFSAR Section 15.7.4.3, an SSI analysis has determined that the drop of an ANF-1.3 fresh 8x8 fuel assembly will result in the failure of 101 rods; the drop of an ANF-1.3 fresh 9x9-5 LTA will result in the failure of 115 rods. In both the fresh 8x8 and the fresh 9x9-5 assembly drop accidents, however, only 39 of the failed rods will be from the stored, irradiated spent fuel bundles as compared with 101 failed irradiated rods determined by the UFSAR accident analyses. , Therefore, the UFSAR evaluation, which addresses the radiological consequences of the drop of a spent fuel assembly onto stored spent fuel bundles, bounds the radiological consequences of dropping a fresh ANF-1.3 fuel bundle. REASON FOR CHANGE: This safety evaluation evaluated the proposed action: a) The storage of fresh reload ANF-1.3 fuel in the new fuel l- vault. b) The storage of fresh reload ANF-1.3 fuel in the spent fuel pool. c) The movement of a new fuel bundle to either of the temporary storage locations described in a) and b). l NLSATTC2/SNLICFLR - 204
Att:chment to AECM-89/0093 NSp-89-001 Page 4 j 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 evaluated in the Safety l Analysis Report.
. Confirmatory analyses have been performed to show that the ANF-1.3 reload fuel is compatible with and similar to the reload fuel 4
stored in the New Fuel Vault during previous reload activities. Therefore, the precursors to any accident previously evaluated will not be affected. The NRC has approved a revision to the licensing basis for storage of the ANF-1.3 reload fuel in the spent fuel storage racks. Because of the similarity of the ANF-1.3 reload fuel to the reload fuel stored in the spent fuel pool during previous reloads, the storage of the new fuel types in the spent fuel storage racks will not affect the precursore to any accident previously evaluated. Confirmatory analyses have been performed to show that the ANF-1.3 reload fuel bundles have weights and geometries similar to those of the GE fuel bundles on which the analyses described in the SAR are based. The momentum and kinetic energy effects of dropping a fresh fuel bundle for ANF-1.3 reload fuel are similar to those for previous reload fuel types. The number of failures of irradiated rods caused by dropping an ANF-1,3 reload fuel bundle as determined by the analyses, will not exceed the failures caused by dropping a fuel bundle of the type used in the FSAR analyses. Therefore, the precursors to any accident previously evaluated will not be affected. Confirmatory analyses have shown that the reactivity of the ANF-1,3 reload fuel in the New Fuel Vault is within the acceptance i criteria established for previous reloads for new fuel. Therefore, as for previous reloads, the occurrence of inadvertent criticality is precluded for ANF-1,3 reload fuel. The NRC has approved a revision to the licensing basis for storage of the ANF-1.3 reload fuel in the spent fuel storage racks. The analyses performed in support of the revised basis show that the maximum reactivity of the racks when loaded with ANF-1.3 reload L fuel is within the acceptance criteria for spent fuel pool criticality for the storage of new fuel as determined for previous l reloads. The equipment to be used for the on-site storage and handling of the new fuel bundles is similar to that used for previous reloads; no additional 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 I will not be affected. NLSATTC2/SNLICFLR - 205
F Attrchment to AECM-89/0093 l s. l NSP-89-001 [ Page 5 The equipment to be used for the onsite storage and handling of the new fuel bundles (Reference 8) is similar to that used for previous reloads. The fuel handling and storage equipment will L not be subjected to operational conditions different from tnose during previous reloads; changes to the equipment protection features will not be required. The consequences of malfunction of equipment important to safety are bounded by the consequences of the Puel Handling Accident evaluated in the FSAR. Therefore, there is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously l' in the Safety Analysis Report. Analyses have been performed to determine the reactivity for the ANF-1.3 reload fuel. The acceptance criterion stated in the FSAR for K-effective in the new fuel vault is 0.95. The corresponding licensing basis value for maximum in-core reactivity is 1.31 (K-infinity) as determined for previous reloads (Reference 3).
.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-core reactivity was calculated (in Reference 1) to be 1.25 (K-infinity). This value is below the acceptance criterion of 1.31 established for previous reloads. The NRC has approved a revision to the licensing basis for the storage of the ANF-1.3 reload fuel in the spent fuel pool storage racks. The maximum reactivity for the storage of fresh 8x8 fuel as stated in the NRC safety evaluation is 0.936 (K-effective); the corresponding value for fresh 9x9-5 fuel is 0.9197. The K-effective for both fuel types is therefore shown to be below the acceptance criterion of K-effective = 0.95. The acceptance criterion remains unchanged from that for previous reloads. The radiological consequences of a Bundle Drop Accident depend l primarily on the number of failures of irradiated rods. Analyses have been performed to determine the maximum number of rod l failures of irradiated rods consequent upon dropping a fresh fuel i bundle on stored, irradiated fuel bundles. The results show that 39 of the failed rods will be from the irradiated fuel bundles. The UFSAR accident analyses show that a maximum of 101 irradiated rods would fail in the limiting situation when a spent fuel bundle is dropped onto stored spent fuel bundles. Therefore, the failure of 101 irradiated rods continues to be the limiting event for determining the radiological consequences of the fuel bundle drop accident. Therefore, performing the activities in connection with onsite storage of new fuel for Cycle 4 will not result in a reduction in the margin of safety as defined in the basis for any technical specifications. NLSATTC2/SNLICFLR - 206 l ____.__ ____ a
Attcchment to AECM-89/0093 SRASN: NSP-89-002 DOC NO: Cycle 3 Operation SYSTEM: N/A DESCRIPTION OF CHANGEt This change extended Cycle 3 operation by approximately 200 mwd /MTU to 18,800 mwd /MTU. The GGNS Cycle 3 core average exposure is extended beyond that assumed in the i Cycle 3 safety analysis performed by Advanced Nuclear Fuels (ANF).
, REASON FOR CHANGE: This change allowed for the continued operation of GGNS until RF03 was initiated.
i 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 evaluated in the Safety Analysis Report. Confirmatory evaluations have shown that the l accident analyses described in the UFSAR (CCSA) continue to remain applicable for the extension of Cycle 3. This activity does not affect the precursors to any accident previously evaluated. This activity does not require changes in the manner in which plant systems are operated and does not affect overall system performance such that the probability of occurrence of an accident previously evaluated in the FSAR is increased. The proposed cycle exposures remain bounded by the approved fuel ; 1 mechanical design limit exposure; therefore, the fuel clad boundary performance is unaffected. The impact of a change in the exposure assumptions for accidents previously evaluated have been assessed. With the restriction that the exposure INER be raintained less than 1.0, the consequences of accidents previously evaluated in the UFSAR (CCSA) are bounding with respect to the cycle extension. This change has no direct or indirect impact on equipment important to safety. No equipment is altered by the cycle extension. The original design criteria is not impacted. The proposed change does not affect the reliability of structures, systems or components, system protection features, system reliability performance or challenges to systems or equipment. The precursors to any malfunction of equipment important to safety is not affected. l The extension of Cycle 3 does not subject the equipment to operational conditions different from those experienced previously during Cycle 3; changes to the equipment protection features are not required. The consequences of malfunction of equipment important to safety are bounded by the consequences of accidents described in the UFSAR. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Operation of the reactor during the cycle extension is similar to and consistent with operation throughout Cycle 3. The extension does not require any activities NLSATTC2/SNLICFLR - 207
7 Attcchment t3 AECM-89/0093 L NSP-89-002 Page 2 L different from those conducted throughout cycle 3; no new L operational modes are required; no plant modifications are i required., Besed on the similarity of plant operation during the cycle extension to that experienced throughout Cycle 3, no new equipment is required; no new activities are required; no i modifications to the existing equipment are required; no changes in operational setpoints are required. Accidents analyses applicable throughout cycle 3 operation continue to be applicable for the extension to Cycle 3. An evaluation of the impact of the cycle extension on the safety analysis results described in the UFSAR (CCSA) was conducted by ANF. An additional evaluation was performed by SSI Nuclear I Engineering Services. These evaluations determined that:
- 1. The increased exposure could result in an increase in delta CPR of loss than 0.01. Analyses performed for a similar plant cycle extension demonstrated a maximum increase in delta CPR of 0.003.
- 2. The current Cycle 3 safety analysen support a CPR operating limit of 1.17.
The current CPR operating limit (1.10) is thus bounding for the impact of increased cycle exposure; there is no reduction in the margin of safety. In addition, the cycle 3 core is beyond full power capability. As exposure continues to increase, power and therefore rod line are reduced. Therefore, the CPR margin to the operating limit during the cycle extension will increase. The extension to Cycle 3 does not require a change in system configuration or allow operation such that any Technical Specification Limiting Conditions for Operation (LCOs) or surveillance requirements must be revised. The activity does not bypass or invalidate automatic activation features required by the Technical Specifications. Therefore, an extension of GGNS Cycle 3 to a core average exposure of 18,800 mwd /MTU does not result in a reduction in the margin of safety as defined in the basis for any technical specifications. NLSATTC2/SHLICFLR - 208
f n Atttchment t3 AECM-89/0093 lp', f SRASN: NSp-89-003l DOC NO: ~ 28 ANF-1.1 8x8 Fuel Assemblies SYSTEM: N/A DESCRIPTION OF CHANGE: During the third refueling outage (RF03) [' at Grand Gulf Nuclear Station Unit 1 (GGNS-1), SERI will replace
-28 ANF-1.1 8x8 fuel assemblies and the remaining 248 GE initial core Bx8 fuel assemblies with 272 new, unirradiated ANF-1.3 8x8 , fuel assemblies and 4 new, unirradiated ANF 9x9-5 Lead Test Assemblies (LTAs). The new fuel assemblies are neutronically, thermal-hydraulically, and mechanically similar to the ANF fuel assemblies introduced into the GGNS-1 core during previous reloads. The significant-differences from previous reloads are that.the new ANF 8x8 fuel assemblies have a slightly higher bundle 7 average enrichment _(3.37 weight percent (w/o) U-235 as compared b with 3.01 w/o).and that 4 ANF 9x9-5 LTAs'are being introduced into h the core during RF03. All ANF fuel assemblies were designed and
[ built specifically for the GGNS-1 reload core. I. ' p~ - REASON FOR CHANGE: To allow the following usctions which are not , described in the UFSAR. l'-
- 1. Refueling operations, as allowed by the GGNS-1 Operating
' License, with ANF-1.3 fuel.
- 2. Operations in connection with RF03 for Cycle 4 fuel in Modes 4, 5, and * , as allowed by the GGNS-1 Operating License, with an all-ANF core comprising ANF 8x8 fuel and ANF-'
9x9-5 LTAs. 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 evaluated in the Safety Analysis Report. Confirmatory evaluations have shown that the ? accident analyses described in the UFSAR, and which were applicablo to previous reloads, continue to remain applicable for RF03 and continue to bound the accidents applicable to RF03. Therefore, the precursors to any accident previously evaluated , will not be affected. Confirmatory analyses have been performed to show that the ANF-1.3 reload fuel bundles have weights and geometries similar to those of the GE fuel bundles on which the analyses described in the UFGAR are based. The momentum and kinetic energy effects of dropping a fresh fuel bundle for ANF-1.3 reload fuel are similar to those for previous reload types. Therefore, the precursors to any accident previously evaluated will not be affected. The number of failures of. irradiated rods caused by dropping an irradiated fuel bundle, as determined by the analyses, is bounded by the number of failures caused by dropping a fuel bundle of the type used in the UFSAR analyses; the dose rates resulting from the activity releases determined by ANT have been shown to be less than the dose rates resulting from the activity releases stated in the FSAR. The radiological consequences resulting from the bundle drop are bounded by the consequences described in the UFSAR. NLSATTC2/SNLICFLR - 209 y
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F . Atttchment to AECM-89/0093 I l . NSP-89-003 f page 2 Calculations to determine that adequate shutdown margin exists during fuel shuffling have been performed. Restrictions applicable to fuel shuffle activities have been provided for inclusion in the appropriate procedures in a manner similar to previous reloads. Therefore, the precursors to any accident previously evaluated will not be affected. The reload fuel which will be handled during RF03 As similar to and compatible with the reload' fuel which was handled for previous reloads. The equipment required to be used during RF03 is similar to that used for previous reloads; no additional loads will be imposed on any equipment; no increase in frequency of operation of the equipment will result; changes to the equipment protection features will not be required. The precursors to any malfunction ; of equipment important to safety will not be affected. There is no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. The fresh Cycle 4 fuel is similar to and compatible with the fuel inserted into the core during previous-reloads. The design of the fresh Cycle 4 fuel does not require any activities different from those associated with previous reloads; no new operational modes are required; no plant modifications are required; no new equipment will be required; no new activities are required; no changes in operational setpoints are required. There is no reduction in the margin of safety as defined in the , basis for any Technical Specification. Accident analyses applicable to previous reloads during operational modes 4, 5, and
- either remain applicable for RF03, or are cycle-specific (i.e., '
the fuel handling accidents). Accident analyses from the latter category show that the acceptance criteria applicable to previous - reloads continue to be adequately satisfied. The radiological consequences of a Bundle Drop Accident depend primarily on the number of failures of irradiated rods. Analyses have been performed to determine the maximum number of failures of irradiated rods consequent upon dropping an irradiated fuel bundle on stored, irradiated fuel bundles. The results of the ANF analyses shc,w that a maximum of 82 irradiated rods will fail. The UFSAR accident analyses show that a maximum of 101 irradiated rods would fail in the limiting situation when a spent fuel bundle is , dropped onto stored spent fuel bundles. The dose rates calculated based on the ANF-determined activity releases of the Iodine, Xenon, and Krypton isotopes are less than the dose rates calculated based on the activity releases stated in the UFSAR. Therefore, the failure of 101 irradiated rods (and the corresponding release of fission products) continues to be the limiting event for determining the radiclogical consequences of the fuel bundle drop accident. NLSATTC2/SNLICFLR - 210
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Attcchment to AECM-89/0093' ! SRASN: NLS-09-001 DOC NO: SOM Directive No. 1.101, SYSTEM: N/A i Page 4 of 5; Attachment 1, ( page 3 and 4 of 5 p [ DESCRIPTION OF CHANGE: This change reassigns Transportation and i. Telecomunications functions within Corporate Services and ; transfers Support Services within Nuclear Support to Corporate [ Services. This planned reorganization in Corporate Services and Nuclear j Support [' o Reassigns Telecommunications within Corporate Services under the newly created position of Manager Information Services. Creates new position of Telecommunications Supervisor. o Reassigns Transportation within Corporate Services under the newly created position of Manager Administrative Services. Creates new position of Supervisor Facilities Management. ; o Transfers Support Services within Nuclear Support to . Corporate Services under the newly created position of Manager Administrative Services. o Changes title of Manager, Contracts and Support Services to Manager, Contracts. The change will affect UFSAR pages 13.1-19, 13.1-20, 13.1-24 and Figure 13.1-la. REASON FOR CHANGE: The reason for the change is to more , effectively organize transportation and telecommunications functions and consolidate administrative support functions within Corporate Services. SAFETY EVALUATION: There is no increase in the probability of occurrence or in the consequences of e.n accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. With this reorganization all the duties, responsibilities and comitments being perfonned in the existing organizational structures will continue to be performed. As the staffs within Transportation, Telecommunications and Support Services are the same as before, all the required minimum qualifications will continue to be met. The reorganization will have no affect on plant design or operations; therefore, there will be no increase in probability of occurrence or consequences of accidents previously evaluated in the FSAR; nor will there be any increase in probability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated in the FSAR; nor will there be created the possibility of an j i accident or malfunction of equipment important to safety different I than any previously evaluated in the FSAR. Therefore, there is no creation of a possibility for an accident or malfunction of a l different type than any evaluated previously in the Safety Analysis Report. NLSATTC2/SNLICFLR - 212 1
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> The BASES section of the Technical Specifications provide general- ;
requirements applicable to each of.the Limiting conditions for-
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- the; justification for. Safety Limits and Limiting Safety System ,
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Settings. '.None~of these BASES. depend on.the organizational ; structure of Corporate Services or Nuclear Support;~thus the margin of safety as defined in the bases for any Technical j w , specification,is not reduced.- . V, ', _ :
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< Atttchment 2 to AECM-89/0093 ,
10CFR50.59 Evaluated Changes to'the UFSAR (Rev. 4) But Not Previously Reported to the NRC l . DOCUMENT l' ,FSAR l l 9 l' -EVALUATED SECTIONS FSAR IMPACT l
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l l Provide discussion of seismic qualification l
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l UFSAR-CR-88-072l 3.10.1.4.1.7
;l: l, TABLE 3.10-1 .l of thermowells. l
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-l* . -l l< proposed l-TABLES: 13.1-1;l This change consolidated plant Licensing andl l Changes.-to l l: 13.1.3;l Nuclear Licensing. l
. l, Chap. :.13 l FIGURE 13.1-2 l l
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. l Proposed . ' [ 13.1.1.1.2.5; This change relocates Nuclear Licensing froml
-l Changes to l 13.1~.2.2.11;- l the General Office to GGNS asite. The l l'UFSAR l 13.1.2.2.13 l Discipline Support Supervisor, Document l ll -l l. Control' Support Supervisor and Records l l l l Management Supervisor report to the Support l
'l. l l Systems Superintendent. Information Systems l
-l. l l are not the responsibility of the Computer l.
l: l 'l Services Superintendent. l 1
-l- l l l 1, I I l l : Proposed . l 13.1.1.2.1.1 l This document describes the changes to the l' l-Changes to SERIl l SERI corporate organizations which report tol g l Corporate l l the Vice President, Accounting and Treasurer l' l Organization l l the Vice president, Human Resources and l slc l l Administration and the Vice President, Legall l- ~l l l and External Affairs. l
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i l . l l l ~ FSA'l Change l 12.3.1.2 l This reflects the removal of the wire cage l
.l .l FIGURES: 1.2-7;l barriers at the heads'of stairwells 1A36, l l l 1.2-8; 6.5-7; l 1A37, and 1A38. l l l 12.3-27 l l
'l' l l l 1 l l FSAR Change l 9.1.4.2.3.8' l This change removes the description and use l l l l of the Jib Crane. l l .I '
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<l-Change to UFSARl Appendix 9C l This change provides information for the l l Analysis of l l Analysis of a Safe Shutdown in the event of l l Safe Shutdown l l a Major Fire Accident. l l-for Major Fire l l l
-l Accident l l l l' I l l l
1 M9 11702/SNLICFLR - 1 9- _ i
4 - g .c . 3-. :Atthchment 2 to AECM-89/0093 i, . kb 1 . , i b , ,l(-1 DOCUMENT .l FSAR . -l l I p( ll' EVALUATED 'i SECTIONS: PSAR IMPACT ._ l
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l p-- NPEFSAR89/ (l'5.4.2; 5.4.7; _ l RHR LPCI A and B injection piping will be -l- ; j '0038) -l FIGURES: 15.2.15. 'l used to return cooled reactor water to the l
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"_ :lLCR-NL "9-015l.. TABLE 8.2-1; ll Thin change updates and corrects 'l !' l. : l: TABLF. 8.2-la;"
l' transmission line. outages data for.the'11FKVl ' w._ '.l' l Section 8.22.1;-~ l and.500KV lines that supply offsite. _l P l" l FIGURES: ' 8.2-1;J l electrical power of Grand Gulf Unit 1.- l F l 1 8.2-2 = l _- L" l t l .. l . l l(CR-09/0035- l TABLE'3.10-l' .l This change deletes some of the subject I g -l' .l. l components from TABLE'3.10-1. j ._
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. Proposed l 13.1.1'.1.1; l+This change realigns the reporting of the l Reporting. l 13.1.1.1.1.6;n l Quality Programs organization'from the Vice l ,
lJLine Change ll 13 .1. '1.1.1. 6. 3 ; l ' President' Nuclear Engirmering and Support tol !
-l for Quality l 13.1.1.1.1.6.3.1; l the Vice President, Nuclear Operations. l '
g Programa' ,l 13.1.1.1.1.6.3.2; l l E Organization l, 13.1.1.2.1.2; i l q
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l' l-13.1.1.2.1.2.2.5; l l p fl' l ' 13.1.1.'2.1. 2. 2. 5.1 ; l l .{ l- . l - 13.1'.1. 2.1. 2. 2. 5. 2 ; l l ,
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I I W, l? Creation of l 13 .1.1.1.~ 1. 6 ; . l This change' creates the position of .l l Director of.l 13.1.1.1.1.6.2; I Director of Operations Support. l
.'l' Operations _ l 13.1.1.2.1.3; j l l Support ~
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. I I I l CR-NL-89-009l 13.1.1.2.2.1; 'l This change: l Ll' l 13.1.1.1.2.1.2.2.2;l o Crestes a new position of Director, Plant l
- l" l 13.1.1.2.1.2.3 l Projects and Support. l .
l 1 l l l: l 'l o Changes the title of Manager, Special l
'l - l l Projects ' o Manager, Project Manager. l l' l 1 l
.l l l o Separaten the position of Manager, l
- j. .l l Emergency Preparedness and Executive l l l l Assistant to the Vice President of Nuclearl l .l l Operations. l
'l - l l l M9111702/SNLICFLR - 2 m
Attachment 2 to AECH-89/0093 l . DOCUMENT . l- FSAR l l
.l EVALUATED l SECTIONS I FSAR IMPACT l-1 l l l L l l CR-NL-89-009l- l o Changes the reporting line of.the Manager,l l (Continued) l l- Plant-Modification and Construction and- l
.l~ l l the Site Controller. l b, l l l' l l l 1 l-
=l CR-NL-89-10 l-13.1.1.2 l This change creates the position of l l l FIGURES: 13.1.1; l Executive Vice President and Chief Operating l
- l. l 13.1.la l' Officer. l l l l l
'M9111702/SNLICFLR - 3
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