ML20045F055

10CFR50.59 Safety Evaluation for Period 920131-1231, for Grand Gulf Nuclear Station.W/930630 Ltr

ML20045F055
Person / Time
Site:	Grand Gulf
Issue date:	12/31/1992
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GNRO-93-00001, GNRO-93-1, NUDOCS 9307060360
Download: ML20045F055 (200)

Text

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f. Entergy Operations,Inc.

- ENTERGY m*m Pod Geson MS 39150 0 Te:601 437 2300 C. R. Hutchinson m n, ,,

cu.www June 30, 1993 e e c ouom , - m U.S. Nuclear Regulatory Commission Mail Station P1-37 Washington, D.C. 20555 Attention: Document Control Desk

Subject:

Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 Report of 10CFR50.59 Safety Evaluations -

January 1, 1992 through December 31, 1992 GNRO-93/00001 Gentlemen:

I In accordance with the requirements of 10CFR50.59 (b) , Entergy f Operations, Inc. is reporting those changes, tests, and experiments under the requirements of 10CFR50.59 for the period of January 1, 1992 through December 31, 1992. A summary of these changes, tests, and experiments is contained in the attachment. If further information is required, please contact this office.

Yours ruly, o \ ' ,f'i /

b' Q l

CRH/GWR/ams attachment: Table of Contents of 10CFR50.59 Safety Evaluations cci (See Next Page) 060148

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1 9307060360 921231 iE,,i; ,

PDR ADOCK 05000416 E f R PDR lltl

3 June 30, 1993 GNRO-93/00001 Page 2 of 3 cc: Mr. R. H. Bernhard (w/a)

Mr. H. W. Keiser (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L. Thomas (w/o)

Mr. Stewart D. Ebneter (w/a)

Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323 Mr. P. W. O'Connor, Project Manager (w/2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13H3 Washington, D.C. 20555 G9306181 - 2 L

l Attachment to GNRO-93/00001 TABLE OF CONTENTS OF 10CFR50.59 SAFETY EVALUATIONS FOR THE PERIOD JANUARY 1, 1992 THROUGH DECEMBER 31, 1992 SRASN DOCUMENT PAGE NPE-92-001 CR-NPE-91-0009 1 NPE-92-002 NPE-FSAR-91-0064 2 NPE-92-003 NPE-FSAR-91-0051 3 NPE-92-004 CR-NPE-91-0026 4 '

NPE-92-005 DRN #3365 5 NPE-92-006 CR-NPE-91-0027 6 NPE-92-007 CN-91-0246 7 i NPE-92-008 CN-91-0196 8 NPE-92-009 CN-91-0208 9 NPE-92-010 CN-91-0226 10

NPE-92-011 QDR #0160-91 11

NPE-92-012 EER-91-6385 12 i NPE-92-013 EER-91-6487 13 ,

NPE-92-014 CN-92-0102 14 NPE-92-016 CN-92-0022 15 .

NPE-92-017 MCP-91-1157-S00-R01 17 '

NPE-92-018 HNCR 251-89 18 NPE-92-019 MC-OSP64-86058 20 NPE-92-020 DCP-91-0060-S00-R00 21 NPE-92-021 CN-92/0303 23 NPE-92-022 ER GGNS-91/0061, Rev. 0 25 NPE-92-023 QDR 190-92 26 NPE-92-024 MCP-92/1033 27 NPE-92-025 EER-92/6030 29 NPE-92-026 Calc C-A-632, RO, etc. 30 NPE-92-027 Mech. Std. MS-25, R6, App J 32 NPE-92-028 CN-92/0053 34 '

NPE-92-029 CN-92/0023 35 NPE-92-030 CN-92/0295 36 ,

NPE-92-031 MCP-88-1037-R00 37 l NPE-92-032 DCP-87-0005-S00-R01 39 '

40 NPE-92-033 MCP-92-1037-S00-R00  !

NPE-92-034 MCP-91-1073-S00-R00 41 i NPE-92-035 MCP-91-1088-S00-R00 42 NPE-92-037 DCP-91-0072-S00-R00 43 i NPE-92-038 MCP-91-1035-r l'-R00 45 l NPE-92-039 MCP-91-1066-Sus-R00 46 NPE-92-040 MCP-91-1080-S00-R00 48 NPE-92-041 DCP-91-1034-S00-R00 49 NPE-92-042 DCP-91-0030-S00-R00 50 G9306181 - 4 l

Attachment to GNRO-93/00001 k

TABLE OF CONTENTS i OF 10CFR50.59 SAFETY EVALUATIONS FOR THE PERIC."

JANUARY 1, 1992 THROUGH DECEMBER 31, 1992 SRASN DOCUMENT PAGE NPE-92-043 DCP-90-0137-S00-R00 51 NPE-92-044 MCP-90-1036-S00-R00 52 NPE-92-045 DCP-91-0026-S03-R00 53 '

NPE-92-046 MCP-90-1033-S00-R00 54 NPE-92-047 DCP-90-0005-S01-R00 55 NPE-92-048 DCP-89-0089-S00-R00 56 NPE-92-049 DCP-89-0171-S00-R00 58 .

NPE-92-050 DCP-88-0213-S00-R00 60 NPE-92-051 DCP-88-0011-S00-R00 61 NPE-92-052 DCP-88-0087-S00-R00 62 NPE-92-053 DCP-82-0056-S01-R00 63 ,

NPE-92-054 DCP-88-0063-S00-R00 65 NPE-92-055 CN-91-0208 67 1 NPE-92-056 MCP-91-1059-S00-R00 68 NPE-92-057 MCP-91-1130-S00-R00 69 NPE-92-058 DCP-88-0284-S01-R00 71 NPE-92-059 DCP-88-0284-SO4-R00 73 NPE-92-060 DCP-88-0284-SO2-R00 75 NPE-92-061 DCP-88-0064-S00-R00 76 NPE-92-062 MCP-92-1087-500-R00 78 NPE-92-063 MCP-90-1070-S00-R00 79  ;

NPE-92-064 MCP-91-1044-S00-R00 80 -

NPE-92-065 DCP-89-0087-S00-R00 81 '

NPE-92-066 MCP-91-1028-S00-R00 83 .

NPE-92-067 MCP-92-1072-S00-R00 84 l NPE-92-068 MCP-91-1074-S00-R00 85 l

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NPE-92-069 MCP-91-1094-500-R00 86 NPE-92-070 DCP-84-0049-S00-R01 87 NPE-92-071 MCP-91-1084-S01-R00 88  ;

NPE-92-072 DCP-88-0060-S01-R00 90 '

NPE-92-073 MCP-92-1055-S00-R00 92 NPE-92-074 MCP-91-1001-S00-R00 93 j NPE-92-075 MCP-90-1032-S00-R00 94 NPE-92-076 MCP-90-1109-S00-R00 95 NPE-92-077 MCP-90-1059-S00-R00 96 NPE-92-07B DCP-90-0005-S03-R00 97 NPE-92-07S DCP-90-0109-S00-R00 98 NPE-92-080 DCP-87-0087-S00-R00 100 NPE-92-081 DCP-87-0039-S00-R00 102 NPE-92-082 DCP-87-0048-S01-R00 104 NPE-92-083 DCP-87-4023-S00-R00 105 l

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Attachment to GNRO-93/00001  !

I TABLE OF CONTENTS OF i 10CFR50.59 SAFETY EVALUATIONS FOR THE PERIOD JANUARY 1, 1992 THROUGH DECEMBER 31, 1992 )

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SRASN DOCUMENT PAGE i NPE-92-084 DCP-86-4506-S00-R00 107 NPE-92-085 DCP-86-0067-S00-R00 108 NPE-92-086 DCP-84-0189-S00-R01 109 NPE-92-087 DCP-84-0181-S00-R00 110 NPE-92-088 DCP-83-0515-S00-R01 111 )

NPE-92-089 DCP-83-0417-S00-R00 112 NPE-92-090 DCP-83-0417-S00-R00 113 NPE-92-091 MCP-90-1019-S00-R00 114 -

NPE-92-092 DCP-82-5074 .c00-R01 115 NPE-92-093 DCP-83-0170-S00-R00 118 NPE-92-094 DCP-84-0107-S00-R00 119 NPE-92-095 DCP-88-0050-S00-R00 120 NPE-92-096 DCP-84-0249-S00-R00 122  !

NPE-92-097 MCP-90-1011-S00-R00 123 NPE-92-098 DCP-91-0026-S01-R00 124 NPE-92-099 EER-92/6184 125  ;

NLS-92-001 Switchyard Fence 126 .

Modification  :

NLS-92-002 CR-PLS-91-002 127 I NSP-92-001 Cycle 6 Fuel Design & 128 Criticality Analysis l NSP-92-002 Fuel Movement / Shuffle 129 ,

During RF05 NSP-92-003 MNCR-0166-92 131 NSP-92-004 Fuel Related Cycle 6 134 Operational Issues NSP-92-005 Transfer of Environ- 136 mental Surv Program PLS-92-001 TSTI-1C34-92-002-0-N 137 l PLS-92-002 UFSt.R 5.2.2.10 138 PLS-92-003 CR-PLS-92-001 139 PLS-92-004 Temp Alt 92-0011 140 PLS-92-005 W.O. #65005 142 PLS-92-006 02-S-01-4, R24; 144 02-S-01-5, R23 PLS-92-007 Temp Alt 92-0012 146 PLS-92-008 W.O. #67130 148 PLS-92-009 W.O. #28560 150  ;

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l Attachment to GNRO-93/00001 TABLE OF CONTENTS OF 10CFR50.59 SAFETY EVALUATIONS FOR THE PERIOD l JANUARY 1, 1992 THROUGH DECEMBER 31, 1992 SRASN DOCUMENT PAGE i PLS-92-010 06-OP-1000-D-0001, 152 R40, TCN 106 PLS-92-013 UFSAR 11.5.2.3.2 154 PLS-92-014 MWO 68359 155 PLS-92-015 MWO 66729 157 PLS-92-016 05-1-02-VI-2, R17 159 PLS-92-017 OpCon 4 or 5 in 161 i Action c or d PLS-92-018 Tech Spec 3/4.9.7 163 PLS-92-019 Temp Alt 92-0017-O-N 165 PLS-92-020 TSTI-1B33-92-011-0-S 167 PLS-92-021 W.O. #64619 169 PLS-92-022 W.O. #67130 170 PLS-92-023 W.O. #67130 172 PLS-92-024 EER-92/6127 173 -

PLS-92-025 EERR-92/6109 174 PLS-92-026 07-S-14-184, R11,TCN 13 175 PLS-92-027 Plywood Containment Hatch 176 PLS-92-028 W.O. #71304 177 PLS-92-029 TSPS 128, Rev. 0 178 PLS-92-030 EER-92/6127 179 PLS-92-031 MNCR-171-92 180 PLS-92-032 07-S-14-186, R12 181 PLS-92-033 TCN 35 to 04-1-01-G41-1 182

• PLS-92-034 Temp Alt 92-0031 183 i

PLS-92-035 Temp Alt 92-0033 184 PLS-92-036 LDCR Changa 185 No. PLS-92-007 PLS-92-037 EER-89/6229 187 PLS-92-038 MWO 82439 188 PLS-92-039 01-S-06-5, Rev. 22 190  !

PLS-92-040 Site Directive G4.110 191 Rev. 3 PLS-92-041 FSAR CR PLS-92-009 192 PLS-92-042 04-1-03-N19-1 193 PLS-92-043 W.O. #84785 195 PLS-92-044 Transfer of Environ- 196 mental Sury Program PLS-92-045 MAEC-82/0093 197 PLS-92-046 UFSAR CR NL-92-005 198 PLS-92-047 LCTS 15692 199 (AECM-89/0074)

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Attachment to GNRO-93/00001 l

SRASN: NPE-92-001 DOC NO: CR-NPE-91-0009 DESCRIPTION OF CHANGE: A Procurement Engineering Group i

was formed in the Nuclear Plant Engineering (NPE)

Organization. The Group is under the direction of the Manager of Engineering Support. The Group is responsible for making independent engineering and l design basis judgements related to procurement matters I in the Nuclear Plant Engineering Organization and those  ;

associated with the plant. 7 Formation of the Procurement Engineering Group requires  !

revision of the Management and Technical Organizations  :

Section 13.1 of the UFSAR to identify the existence of 3

and functional responsibilities for the Group.

REASON FOR CHANGE: To make Section 13.1 of the UFSAR -

consistent with the present organization configuration of Nuclear Plant Engineering.

SAFETY EVALUATION: The Procurement Engineering Group

will ensure the design basis of the plant by providing

a quality, cost effective, engineering basis for the.  !

procurement of materials required for the operation of

the plant. The addition of the Procurement Engineering ,

l Group into the NPE organization constitutes a purely i

administrative change to the UFSAR and will not have a direct affect on safety or the operation of the plant.

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Attachment to GNRO-93/00001 SRASN: NPE-92-002 DOC NO: NPE-FSAR-91-0064 i DESCRIPTION OF CHANGE: To include a discussion in the UFSAR which addresses the consequences of a missile striking and collapsing / pinching the diesel generator ,

fuel oil storage tank vent line. ,

i REASON FOR CHANGE: To clarify that all essential portions of the diesel generator fuel oil system is .

protected from damage by flying debris carried by l tornadoes and hurricanes.  !

SAFETY EVALUATION: The changes in no way change or I alter the design or any of the operating parameters 4 associated with the Standby or HPCS Diesel Generator Systems; therefore, the probability of occurrence or  !

the consequences of an accident previously evaluated in [

the SAR are not increased. Revising the UFSAR to  !

clarify that all essential portions of the diesel generator fuel oil system are protected from flying debris carried by tornadoes and hurricanes and to  :

include a discussion addressing the consequences of a '

fuel oil storage tank vent being struck by a tornadic missile which results in a pinched or collapsed vent in ,

no way affects the availability or capability of the ,

, Standby or HPCS Diesel Generators to perform their i safety function, therefore the changes in no-way i

> increase the probability of occurrence or the consequences of a malfunction of equipment important to i safety previously evaluated in the SAR. No new failure }

modes are being created, the onsite electrical sources j can still sustain a single failure and perform their '

safety function. The changes will not require a change ,

2 to the GGNS Unit 1 Technical Specifications or reduce t the margin of safety as defined in the basis for any

. Technical Specification.

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SRASN: NPE-92-003 DOC NO: NPE-FSAR-91-0051 l I

DESCRIPTION OF CHANGE: To eliminate the discussion in the UFSAR of two unit operation (analysis'of SSW system operating " Mode III") as it pertains to the ultimate heat sink due to the fact that Unit 2 has been i cancelled.

REASON FOR CHANGE: To reflect the inapplicability of two unit operation. -

SAFETY EVALUATION: The change reflects single unit '

operation of the plant (analysis of SSW system operating " Mode IV") as it pertains to the ultimate heat sink (standby service water basins) rather than

for two unit operation (single unit operation was

l previously analyzed and incorporated into the UFSAR).  ;

The change does not modify Unit i equipment. ,

The revision (which reflects single unit operation only) does not require a change to the GGNS Technical  !

Specifications nor does it create an unreviewed safety  :

question. [

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Attachment to GMRO-93/00001 t

SRASN: NPE-92-004 DOC NO: CR-NPE-91-0026 l 1

1 DESCRIPTION OF CHANGE: Add wording in UFSAR concerning ,

when equalizing charging is performed on the station batteries per Technical Specification requirements and ,

procedures.

REASON FOR CHANGE: To discuss when an equalizing f charge is required for Technical Specification  :

requirements and various maintenance procedures. ,

t SAFETY EVALUATION: The UFSAR discusses that the batteries used at Grand Gulf are of the lead calcium type and are designed for use at an optimum float voltage of 132 to 135 volts DC without a periodic t equalizing charge. The 125 volt batteries in use at  !

i Grand Gulf do not require a periodic equalization I charge if the float voltage is maintained at 132 volts I

or 2.20 volts per cell. If the cell voltage or  ;

! specific gravity is found to be lower than the .,

acceptable level during a test or inspection, an  !

equalizing charge will be performed. When the battery i voltage levels are dropped during an 18 month or 60 month discharge test an equalizing charge is required  !

to restore the batteries to the float level voltage. t The operation of the batteries will not change.. This change will not diminish battery capabilities but enhances operation. No unreviewed safety questions  !

have been introduced. -

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Attachment to GERO-93/00001 1

i SRASN: NPE-92-005 DOC NO: DRN #3365 i

DESCRIPTION OF CHANGE: Revise the High Pressure Core f Spray (HPCS) cooling water pump horsepower rating listed in the UFSAR from 125 to 100 to reflect plant i "As-Built" conditions.

i REASON FOR CHANGE: To correct the horsepower rating of  ;

the HPCS cooling water pump.

, SAFETY EVALUATION: The plant electrical drawing and '

UFSAR indicate that the HPCS cooling water pump motor i size is 125 HP. The actual motor size is 100 HP per vendor documents, a field survey and mechanical ,

specifications. Based on mechanical specifications, a  !

100 HP motor was specified for the HPCS cooling water pump. These changes reflect the "As-Built" conditions {

only.  ;

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i Attachment to GNRO-93/00001 ;

SRASN: NPE-92-006 DOC NO: CR-NPE-91-0027 l

DESCRIPTION OF CHANGE: Revise wording in.UFSAR  ;

concerning Divisions I and II diesel generator frequency recovery percentage rating to correct  ;

numerical value. The frequency recovery, as shown in t the UFSAR tabulated results, indicates a 90% numerical value, this value is actually 98%.

REASON FOR CHANGE: In accordance with Branch Technical Position EICSB-2, " Diesel Generator Reliability  !

Qualification Testing", the diesel manufacturer l performed qualification test on the diesel engines.

One of the required test was a margin test which demonstrated the engine generator could successfully start within 10 seconds with a pre-determined load. As i l part of the margin test, the time for the circuit frequency to recover after a start of the diesel engines was recorded. The test used 98% as the reference point for the frequency to recover. The  ;

UFSAR has 90% as its reference point for frequency ,

recovery. The test reports are correct; therefore, the i UFSAR will be revised. j i

SAFETY EVALUATION: The operation of the Division I and {

II diesel generators will not change. The diesel generators serve as an alternate source for the  ;

Engineered Safety Feature (ESP) load.s and have been  ;

evaluated for several qualification tests, including  ;

the margin test. Based on the reports of the margin tests the diesels recovered to 98% rr.ted frequency.

This change does not alter the test iiata, and therefore .

is bounded by existing analysis. No new safety related equipment or changes to existing safety related equipment has been introduced. No new interfaces are -

created with safety related equipment. No new operating modes or capabilities are introduced by this i change. The diesel generators for Division I and II l are configured as shown in the UFSAR and capable of

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supplying non-accident and accident shutdown loads per }

the qualification tests. The Division I and II diesel i generators qualification tests have confirmed that adequate capacity exist for an alternative source as described in the Technical Specifications. ,

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l Attachment to GNRO-93/00001 SRASN: NPE-92-007 DOC NO: CN-91-0246 .

l DESCRIPTION OF CHANGE: The addition of the Security Remote Facility requires a new power feed breaker to feed a new installed camera.

I REASON FOR CHANGE: To update the UFSAR to reflect the new power feed breaker.

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4 SAFETY EVALUATION: The changes will not compromise any '

existing safety related system, structure or component,-

nor will they prevent safe reactor shutdown. The security equipment is a component of the monitoring i system which is non-safety related. {

All cabling and raceway modifications to be performed l will be in accordance with the separation requirements i of Regulatory Guide 1.75.  !

No evaluated accident is affected by any change to the i security boundary. The components of the security  !

system to be changed are not required to mitigate the j consequences of any evaluated accident. No new  ;

interfaces are created and no new failure modes.are introduced. These changes will not introduce an unreviewed safety question.

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SRASN

NPE-92-008 DOC NO: CN-91-0196 l

i DESCRIPTION OF CHANGE: Delete the drywell chilled water minimum flow and pressure switches and remove the

, discussion of these switches from the FSAR.

REASON FOR CHANGE: Due to the. improvements in the GGNS operating and emergency procedures, drywell cooling is J

. required to be re-established very early in the ,

recovery of a drywell isolation. The risk of damage to  :

the drywell chilled water pumps is minimized and minimum flow protection is not required.

SAFETY EVALUATION: The pressure switches perform a pump protection function for the drywell chilled water  ;

i system. Deletion of these switches will not change any normal operating modes of the drywell chilled water- i system. Failure of the drywell chilled water system will not prevent any ESP system from performing its j intended design function. The drywell chilled water  ;

system is not required to mitigate the consequences of I any accident or equipment malfunction. l h

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Attachment to GNRO-93/00001 SRASN: NPE-92-009 DOC NO: CN-91-0208 DESCRIPTION OF CHANGE: Permanently incorporate a temporarily installed Annubar (flow sensing device) in the minimum flow recirculation line for the condensate pumps.

REASON FOR CHANGE: The Annubar is used for flow testing and calibration purposes and will be installed as permanent plant equipment to facilitate future .

system testing. )

l SAFETY EVALUATION: The condensate system serves no safety function and systems analysis has shown that. l failure of this system will not compromise any safety-related system or prevent safe shutdown. Incorporation ,

of the Annubar as permanent plant equipment to  ;

facilitate future system testing will not affect the function or operation of the condensate system. The  !

piping was designed to ANSI B31.1 code requirements and the pipe support has been determined to be structurally

., adequate and acceptable for the additional load of the Annubar. Since the condensate system is not required ,

to affect or support the safe shutdown-of the reactor {

or perform in the operation of reactor safety features, f incorporation of the Annubar as permanent plant equipment will not create an unreviewed safety .

question.

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Attachment to GNRO-93/00001 SRASN: NPE-92-010 DOC NO: CN-91-0226 DESCRIPTION OF CHANGE: The turbine / generator (T/G) maintenance facility will be located on the 166'-0" elevation of the Unit 1 Turbine Building.

I The electrical bus power for the Turbine Building bridge crane will be modified to allow travel of the  !

crane into the T/G maintenance facility. Crane stops for the bridge crane and gantry cranes will be modified ,

to provide the required safety stops for the cranes.

The Gantry Crane on the east side of the T/G i maintenance facility will remain as permanent plant

, equipment.

i REASON FOR CHANGE: The T/G maintenance facility is being constructed to provide a permanent plant facility to perform maintenance on the high and low pressure turbine rotors and the generator rotor and to provide additional laydown areas.

The Gantry Crane on the west side of the T/G maintenance facility will provide additional lifting capability in the facility during outages. This crane will also provide additional lifting capability at the ,

j hatchway in Unit 1. This will help to reduce the

lifting requirements of the turbine bridge crane during .

! outages. i Reinstalling the electrical bus power to the bridge  ;

crane and removing the crane stops allows travel of the

bridge crane through the facility. ,

, SAFETY EVALUATION: The T/G maintenance facility will be established by completing construction of the Unit 2  !

, Turbine Building. The T/G maintenance facility is a

, non-safety related, non-seismic structure and will be

used during outages and operation for maintenance of  :

the HP and LP turbine rotors and the generator rotors l and to provide additional laydown areas. The facility ,

l has been designed to ensure the structural integrity of  ;

the adjacent safety-related, seismic Category I Control '

Building is maintained. All support systers for the )

I facility, which include floor and equipment drains, ,

lighting, fire protection, ventilation, and cranes, are non safety-related and non-seismic. No new area radiation monitors are required for the facility.

The construction of the T/G maintenance facility will not require a change to the GGNS Technical Specifications nor will it create an unreviewed safety question.

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Attachment to GNRO-93/00001 SRASN: NPE-92-011 DOC NO: QDR #0160-91 DESCRIPTION OF CHANGE: Section 5.3.3.1.4.5.2 of the UFSAR states that automated ultrasonic test (UT) examination methods will be utilized in the performance of feedwater nozzle inner radius examinations. Quality

Deficiency Report (QDR) #0160-91 documents the use of manual UT examination on the feedwater nozzle inner radii. The QDR response accepted the use of manual ,

techniques. The UFSAR is being changed to state that automated UT may be used.

i REASON FOR CHANGE: As documented in the QDR, UFSAR i Section 5.3.3.1.4.5.2 states that automated ultrasonic  !

techniques will be used. At GGNS, both manual and  ;

automated methods have been utilized for Reactor Pressure Vessel (RPV) exams. The plant determined that t use of either method resulted in equivalent man-rem exposures. The cost associated with automated UT is i significantly higher. Therefore, the plant elected to '

use manual UT methods in lieu of automated UT. The QDR  :

response establishes that the use of manual UT j examination methods is acceptable and requires a change  ;

to the UFSAR to delete the requirement to use automated j UT examination. By changing "will"-to "may" in Section  !

5.3.3.1.4.5.2, the use of automated methods is made i optional.

SAFETY EVALUATION: NUREG-0619 contains the NRC requirements for performance of feedwater nozzle inner radius examinations. The NUREG does not require ,

automated methods. The initial GGNS decision to utilize automated UT was voluntary. Inclusion of automated UT in the UFSAR was the result of j incorporation of a Question & Response (Q&R) 121.5 and was not the result of an NRC mandate. Automated and manual procedures provide essentially the same

, examination coverage (some minor reduction in coverage 7 i may occur using the automated method due to physical -

limitations associated with the scanner arm configuration). Both methods use similar transducers  !

and both scan from the vessel OD. The ability of the two methods to detect flaws is equivalent. A change to Technical Specifications is not required. There is no unreviewed safety question.

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Attachment to GNRO-93/00001 SRASN: NPE-92-012 DOC NO: EER-91-6385 DESCRIPTION OF CHANGE: This Engineering Evaluation Response requests that temporary lead shielding be attached to certain portions of the Reactor Water Cleanup (RWCU) System to reduce radiation exposure to personnel performing work in this area. The lead shielding will be installed during Operating Modes 4 and 5 only, and must be removed prior to restart.

REASON FOR CHANGE: The temporary lead shielding will be attached to certain portions of the RWCU system in  ;

order to reduce radiation exposure to personnel performing work in this area.

SAFETY EVALUATION: An engineering evaluation shows ,

that with the added weight.of lead shielding and the '

temporary supports installed, the structural integrity of the applicable RWCU piping is maintained even in the  !

unlikely event of an Operating Basis Earthquake (QBE),

Safety Shutdown Earthquake (SSE), Safety Relief Valve discharge induced loads from one valve's subsequent  ;

actuation (SRVONE) , and Seismic Anchor Movements (SAM). ,

All applicable ASME code stress allowables are met.

Therefore, the system operability in Operating Modes 4 and 5 is not affected by the addition of the temporary lead shielding.

4 i 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, design parameters, or operational parameters of the affected system after the temporary lead shielding and temporary supports are removed. '

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Attachment to GNRO-93/00001 SRASN: NPE-92-013 DOC NO: EER-91/6487 l DESCRIPTION OF CHANGE: This Engineering Evaluation Response will provide the design requirements for ,

temporary removal of eight snubbers and one spring hanger to allow performance of ISI weld inspection on ,

the recirculation system piping. This inspection is to ,

be performed during RF05 during a cold shutdown (Operating Mode 5), while the reactor head is off.

REASON FOR CHANGE: The performance of an ISI weld inspection on the recirculation system piping, Loop A, requires that eight snubbers and one spring hanger be .

temporarily disassembled to allow inspection access to  ;

pipe welds during RF05.

SAFETY EVALUATION: The modifications made by this EER I will not affect the system function, operation, or i performance in any way. The capability of the system i

to perform its safety function is not affected. .

Temporary removal of eight snubbers and one spring hanger will not adversely affect the structural integrity of the associated piping. The piping and [

pipe support designs meet ASME Section III requirements i and are qualified as Seismic Category I. Therefore, the piping and pipe supports will function in their intended manner, t l The reanalysis of the piping systems has shown that all ASME Section III code allowables have been met and therefore, the probability of a piping failure has not increased. The modification will not introduce any new  ;

postulated piping failures and the existing hazards ,

evaluations are not affected.

No new system interfaces with any equipment have been ,

created and no existing interfaces have been adversely l affected. No new failure modes for the system or any

equipment have been created. By remaining within the

same allowables specified by the applicable codes as stipulated for piping, supports, and supporting structures, the margins of safety provided by these allowables are not affected. ,

1 No revision is required to Technical Specification 3/4.7.4 since individual snubbers are not identified in the technical specification. No revision is required  ;

i to Technical Specification 3/4.4.8 since the structural integrity of the piping system is maintained in t accordance with ASME Section XI. Therefore, the

. changes made per this EER will not create an unreviewed ;

j safety question.

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Attachment to GNRO-93/00001 i SRASN: NPE-92-014 DOC NO: CN-92-0102 ,

DESCRIPTION OF CHANGE: The objective of this design change is to resolve the concerns regarding the relief valve reset pressure on the Division I and II Standby' Diesel Generator starting air storage tanks. The objective will be accomplished by replacing the valves with those specified by Energy Services Group, the diesel supplier.

REASON FOR CHANGE: To ensure that adequate starting air pressure is maintained in the event of a discharge or an inadvertent lifting of a relief valve.

SAFETY EVALUATION: Calculations have shown that the >

new valves and additional piping will have no adverse impact to the operation or function of the system and '

will not affect the structural integrity of the starting air storage tank. The replacement valves meet ASME Section VIII and Seismic Category I requirements; the piping and pipe supports were designed to ASME Section III and Seismic Category I requirements and all ,

will function in their intended manner. The new relief  :

, valve will reset at 10% or less than the set pressure. l The set pressure of the relief valve is 275 psig. The <

minimum reset pressure of the valves is 247.5 psig, which is higher than 195 psig required per the '

Technical Specification and the lockout setpoint. This design change will enhance the reliability of the i diesel generators by ensuring the relief valve reset i pressure will maintain the required air pressure in the starting air storage tanks. Therefore, this change will have no adverse effect on the ability of the

l. diesel generators to start when required. No existing. ,

system interfaces are affected by this change and no new interfaces are created.

l This change will not require a change to the Technical Specifications and will not create an unreviewed safety question.

[14]

Attachment to GNRO-93/00001 _

SRASN: NPE-92-016 DOC NO: CN-92-0022 DESCRIPTION OF CHANGE: A modification provided design documentation for replacement of obsolete three inch ball valves in the condenser tube cleaning system (CTCS) with ball valves from a different manufacturer.

This change provides the design documentation for replacing four gate valves with ball valves as installed by the modification change.

The system affected by this change is the Circulating

] Water (CW) System N71, which includes the CTCS.

REASON FOR CHANGE: The function of the four gate valves is isolation. The current valves are not performing their design function because the design of a gate valve allows small suspended particles to interfere with the seating surfaces resulting in leakage past the seat.

SAFETY EVALUATION: The CW system is not addressed in the GGNS Technical Specifications; however, the condenser vacuum setpoint and radioactive liquid effluent monitoring of the CW system blowdown are addressed. The evaluated changes do not alter or affect the condenser vacuum low setpoint as addressed in the Technical Specifications for main steam line isolation or radioactive liquid effluent monitoring as addressed in GGNS Technical Specifications. The changes will not otherwise affect plant operation as described by the Technical Specifications.

An increase in reactor pressure as a result of a turbine trip is evaluated in SAR Chapter 15. Loss of the CW system may result in a turbine trip through the loss of condenser vacuum. Implementation of the evaluated changes will not adversely affect the CW supply to the condenser and will therefore not adversely affect condenser vacuum. All of the described changes will maintain all controls normally depended upon to respond to changes in CW system operation.

The changes will not alter or affect the operability of exirting safety related equipment. In addition, failure of the CW system will not compromise any safety related systems or prevent safe shutdown.

Turbine Building flooding by a gross failure of the CW piping has been evaluated in the SAR. No additional modes of failure are created by implementation of the described changes. Therefore, the existing evaluations are considered bounding for the system.

[15]

i Attachment to GWRO-93/00001 NPE-92-016 Page 2 Tne Technical Specifications do not contain any margins of safety for the operation or design of the CW system.

Implementation of the described changes will not affect or prevent safe shutdown of the reactor vessel.  ;

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Attachment to GNRO-93/00001 SRASN: NPE-92-017 DOC NO: MCP-91-1157-S00-R01 DESCRIPTION OF CHANGE: Minor Change Package (MCP)-91/1157 provides design documentation for the elimination of the Cooling Tower De-icing Ring Isolation Valves. The de-icing ring supports indicate no visible deterioration and therefore the de-icing ring and its supports will be abandoned in place.

REASON FOR CHANGE: The condition of the de-icing ring isolation valves has deteriorated over the years to a point of inoperability. Material Non-conformance l Report (MNCR) 0058-89 documented this condition and prompted an evaluation of whether the de-icing subsystem was required. An engineering report concluded that the de-icing subsystem was not necessary and recommended its decommission.

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SAFETY EVALUATION: The Circulating Water (CW) System is not addressed in the GGNS Technical Specifications; however, the condenser vacuum setpoint is addressed.

The evaluated change does not alter or affect the condenser vacuum low setpoint as addressed in Technical Specification 3/4.3.2 for main steam line isolation.

The change will not otherwise affect plant operation as described by the Technical Specifications. An increase in reactor pressure as a result of a turbine trip is evaluated in SAR Chapter 15. Loss of the CW system may result in a turbine trip through the loss of condenser vacuum. Implementation of the evaluated change will not adversely affect the CW supply to the condenser and will therefore not adversely affect condenser vacuum.

The change will not alter or affect the operability of existing safety related equipment. In addition, a CW system analysis has shown that failure of the CW system will not compromise any safety related systems or prevent safe shutdown. The CW system is described as safety class "other". The CW system serves no safety related function. The change will not compromise any safety related system or prevent safe shutdown since no new interface with equipment important to safety is created nor is such equipment prevented from operating as designed. Turbine Building flooding by a gross failure of the CW piping has been evaluated in the SAR.

No additional modes of failure are created by implementation of the described change. Therefore, the existing evaluations are considered bounding for the system. The Technical Specifications do not contain any margins of safety for the operation or design of the CW system. Implementation of the described change will not affect or prevent safe shutdown of the reactor.

[17)

httachment to GNRO-93/00001 .

SRASN: NPE-92-018 DOC NO: MNCR 251/89 DESCRIPTION OF CHANGE: Assess the ability of the Control Room Heating, Ventilation and Air Conditioning System to meet established design criteria considering the potential design deficiency described below with the implementation of appropriate administrative controls for operator action.

A deficiency document was written to address potential problems with the operation of the Control Room A/C Units following a LOP /LOCA.

REASON FOR CHANGE: To maintain operation of the Control Room air-conditioning units after a LOP /LOCA in order to maintain the approved temperature in the Control Room.

SAFETY EVALUATION: Implementation of the disposition for the deficiency document will not increase the probability of occurrence of an accident previously evaluated in the FSAR. The disposition specifies the criteria for establishing Standby Service Water flow rates to the Control Room A/C units, following a LOP /LOCA, to ensure that the units are capable of removing the design heat load from the Control Room envelope. Failure to regulate the flow of SSW to the condensers on the Control Room A/C units following a LOP /LOCA will result in degraded unit performance (i.e., reduced unit capacity), however this will not impact the system failure analysis or the accident analyses presented in the FSAR.

Operator action will be required to throttle the SSW flow; however, personnel doses will not exceed those specified in the FSAR and will be well within the guidelines of General Design Criteria 19, Appendix A of 10CFR50.

Implementation of the disposition will not increase the consequences of an accident previously evaluated in the FSAR. The disposition provides the guidelines which are to be used to establish the SSW flow rates to the Control Room A/C units following a LOP /LOCA to ensure I that the units are capable of maintaining the Control Room envelope at the design conditions of 72 F and 50%

relative humidity.

[18]

Attachment to GNRO-93/00001 i

NPE-92-018 Page 2 Implementation of the disposition will not increase the probability of a malfunction of equipment important to safety previously evaluated in the FSAR. The safety function of the Control Room HVAC System is to maintain a habitable environment and ensure the operability of components in the Control Room. Establishment of mitigating administrative controls which require that the SSW flow rate to the Control Room A/C units following a LOP /LOCA be regulated as specified in the disposition will ensure that the units are capable of -

performing their safety related function. Therefore, the probability of a malfunction of equipment important to safety previously evaluated in the FSAR has not been increased.

Implementation of the actions specified in the disposition will not increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR. As long as the criteria established for the flow of SSW to the units in the disposition are met, the consequences of the malfunction of this equipment are within those previously evaluated in the FSAR.

Implementation of the disposition will not create the possibility of a malfunction of equipment important to safety different than previously evaluated in the FSAR.

No new fai]ure modes are being introduced.

Implementation of the disposition will not reduce the margin of safety as defined in the basis for any technical specifications. Technical Specification 3/4.7.2 addresses the Control Room Emergency Filtration System; however, the Emergency Filtration System is in no way impacted by the disposition. Establishment of mitigating administrative controls assures that the Control Room HVAC system is capable of removing the design heat load from the Control Room following a LOP /LOCA; therefore, the margin of safety will not be reduced.

[19]

Attachment to GNRO-93/00001 SRASN: NPE-92-019 DOC NO: MC-OSP64-86058 DESCRIPTION OF CHANGE: This revision of the Combustible Heat Load Calculation (HLC) is conducted to account for the addition of combustible materials into Fire Zone OC507 and the removal from and addition of combustibles to OC704. The combustibles added to OC507 are office furniture. The combustibles removed from OC704 are temporary plywood flooring and those added are office furniture, above and below floor cabling, a raised fiberglass floor system and a video projection

]} unit.

REASON FOR CHANGE: Maintenance of the Fire Protection Program requires, in part, the accounting and control of insitu and transient combustibles allowed within the '

plant, specifically as they impact the Fire Hazards Analysis (FRA). The purpose of this effort is to demonstrate that the fire protection measures in place are adequate to mitigate the effects of a fire so as not to advarsely impact the ability to achieve and maintain safe shutdown of the plant.

Also this revision of the HLC assures an accurate description of the plant combustible loading and postulated fire durations for each fire zone within the plant. Therefore, this change substantiates that the addition of the subject combustibles does not exceed the design limitations of the existing fire protection systems in the affected fire zones.

SAFETY EVALUATION: This change will not affect technical specifications or the bases for any technical specifications because Operating License Amendment No. 82 relocated the Fire Protection Technical Specifications to UFSAR Appendix 16A. This change will not increase the probability for occurrence of accidents or a malfunction of equipment important to safety previously evaluated in the SAR because no new ignition sources have been added and the combustibles added are similar to those previously evaluated. This change will not increase the consequences of an accident or malfunction of equipment important to safety from that previously evaluated in the SAR because the postulated fire will not spread or adversely affect equipment or components beyond that previously evaluated in the SAR. This change will not create the possibility for an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the SAR because the fire type and its ability to propagate beyond the fire area of origination has not changed from that previously evaluated.

{20)

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Attachment.to GNRO-93/00001 SRASN: NPE-92-020 DOC NO: DCP-91-0060-S00-R00 DESCRIPTION OF CHANGE: Design Change Package (DCP) 91/0060 provides design documentation for the ,

following:

a. Installation of replacement fill and spacers to maintain the thermal performance of the cooling tower while minimizing the propensity of the fill ,

system to foul. j

b. Installation of replacement drift eliminators and a support system to eliminate the problens due to. '

the distribution pipe couplings. l

c. Installation of structural members to increase the I

load carrying capability of the stirrup and lintel support system.

REASON FOR CHANGE: The tower performance has declined to the point that the cold water returning to the condensers is 6.5 F hotter at design conditions than it  ;

would have been when the tower was newly rebuilt.

Approximately 17 MWe loss of plant output has been attributed to the fouling of the fill. t SAFETY EVALUATION: The Circulating Water (CW) System ,

is not addressed in the GGNS Technical Specifications; however, the condenser vacuum setpoint is addressed.

None of the evaluated changes alters or affects the- '

condenser vacuum low setpoint as addressed in Technical Specification 3/4.3.2 for main steam line isolation.

The changes will not otherwise affect plant operation as described by the Technical _Sspecifications.

An increase in reactor pressure as a result of a turbine trip is evaluated in SAR Chapter 15. Loss of the CW system may result in a turbine trip through the  ;

loss of condenser vacuum. Implementation of the evaluated changes will not adversely affect the CW supply to the condenser and will therefore not adversely affect condenser vacuum. All of the described changes will maintain all controls normally depended upon to respond to changes in CW system operation. The instrumentation will only provide local indication and inputs to the data acquisition system and will not be relied on to respond to any changes in CW system operation.

[21]

Attachment to GNRO-93/00001 NPE-92-020 Page 2 The changes will not alter or affect the operability of existing safety related equipment. In addition, a CW -

system analysis has shown that failure of the CW system will not compromise any safety related systems.or prevent safe shutdown. The CW system is described as ,

2 safety class "other". ,

DCP-91/0060, the additional instrumentation, the sample point, and the removal of the acid dilution and  !

dispersal troughs will not alter the design, function or operation of any equipment important to safety as evaluated in the SAR. The CW system serves no safety

• related function. The changes will not compromise any safety related system or prevent safe shutdown since no new interface with equipment important to safety is created nor is such equipment prevented from operating as designed.

Turbine Building flooding by a gross failure of the CW '

piping has been evaluated in the SAR. No additional modes of failure are created by implementation of the described changes. Therefore, the existing evaluations are considered bounding for the system.

The Technical Specifications do not contain any margins of safety for the operation or design of the CW system.

Implementation of the described changes will not affect or prevent safe shutdown of the reactor vessel.

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Attachment to GNRO-93/00001 SRASN: NPE-92-021 DOC NO: CN-92/0303 DESCR.IPTION OF CHANGE: The Leak Detection System deferential temperature recorder E31R611 currently has 12 thermocouple inputs that represent 12 area differential temperatures. Each subtraction results from a pair of thermocouples connected in series at a terminal block inside the panel where the recorder is mounted. Forty eight thermocouple wires are reduced to 24 which are then connected to the recorder. This subtraction method is not accurate because of the non-linerarity of the thermocouple curve. This problem can -

i be corrected by replacing the four Erasable Progranmable Read Only Memory (EPROM) chips, wiring the 24 thermocouples directly to the recorder and allowing the recorder to perform a software subtraction of the thermocouples. The point assignment of. the differential temperature channels will have to be changed from 1 thru 12 to 25 thru 36.

REASON FOR CHANGE: The recorder E31R611 cannot meet the specified accuracy as a result of the non-linearity of the thermocouple curve.

SAFETY EVALUATION: Delta T recorder E31R611 monitors various areas of the plant for leak detection purposes.

The recorder has six contact outputs which power control room annunciators when a setpoint is exceeded.

This alerts the operators to potential steam leaks.

This recorder is non-safety related, however, and performs no active safety related function. It is not required for Regulatory Guide 1.97 indication and no credit is taken in the UFSAR for operator actions based on information taken from the recorder or the associated annunciators. The recorder is adequately isolated from the Class 1E bus.

The changes will not compromise any existing safety related system, structure or component nor will they prevent safe reactor shutdown. No evaluated accident is predicated by a failure of the affected recorder.

This design change will be an improvement in terms of reliability and monitoring capability. The changes will not compromise any existing safety related system, structure or component nor will they prevent safe reactor shutdown. No evaluated accident is predicated by a failure of the affected recorder. This design change will be an improvement in terms of reliability and monitoring capability. The changes will not compromise any existing safety related system,

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Attachment to GNRO-93/00001 NPE-92-021 Page 2 )

l structure or component. The failure of the recorder .

will not initiate any evaluated transient or accident.

The E31 (Leak Detection) system operation and function l will not change. The recorder is not required to i mitigate the consequences of any evaluated transient or '

accident. No new interfaces are created and no new failure modes are introduced. This change will therefore not introduce an unreviewed safety question.

The recorder is not currently addressed in the  !

Technical Specification and this change will not  !'

require that it be added to the Technical

. Specification.

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I SRASN: NPE-92-022 DOC NO: ER GGNS-91/0061, i Rev. 0 DESCRIPTION OF CHANGE: Engineering Report GGNS-91/0061 I documents and defines the control room envelope and its design basis for the Grand Gulf Nuclear Station. The  ;

definition of the envelope and its design basis  ;

included in the report differ from the current  ;

description in the FSAR in that the report includes the HVAC chases, OC404A and OC405A, but does not include the Unit II upper control panel room in the envelope.  ;

This is acceptable because the total free volume t included in the control room envelope only changes from i approximately 253,000 cubic feet to 252,804 cubic feet. I This small change in control room volume has no significant effect on either the performance of the l Control Room HVAC system or any analysis supporting i control room habitability. Also, the number of people '

the Control Room HVAC System is capable of sustaining in the emergency mode of operation is increased from 12 to 50. As shown by Calculation MC-QSZ51-87068, Rev. 0, this is within the capacity of the Control Room HVAC System.

i REASON FOR CHANGE: Engineering Report GGNS-91/0061, Rev. O, " Define Control Room Envelope", was prepared to  !

define and document the requirements of the Grand Gulf Nuclear Station Control Room envelope. The various >

requirements for the envelope previously existed but 1

were not in one location. The primary purpose of this ,

engineering report is to gather together, in a central  ;

location, these various requirements. ,

t i i SAFETY EVALUATION: This safety evaluation addresses  ;

the impact of Engineering Report GGNS-91/0061, Rev. 0  :

and Calculation MC-QSZ51-91152, Rev. O on the safe >

, operation and accident mitigation capabilities of the t J

Grand Gulf Nuclear Station. This evaluation concludes I that the subject engineering report succinctly  :

i summarizes the existing data relative to the control  ;

room envelope. It does not affect any operational i characteristics of any system. It is therefore i concluded that this does not constitute an unreviewed safety question.

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i Attachment to GNRO-93/00001 I

t SRASN: NPE-92-023 DOC NO: QDR 190-92 ,

i DESCRIPTION OF CHANGE: Quality Deficiency Report (QDR) 190-92 documented a deficiency in the Siemens j Power Corporation (SPC) procedure used to perform the i Feedwater Controller Failure Analysis (FWCF). As a  !

result of the deficiency, SPC reanalyzed the FWCF event .

for the Cycle 5 and 6 reloads. The results of the  ;

analyses indicated a small increase in the calculated  !

delta-CPR (Critical Power Ratio) . The Cycle 6 i technical specification operating limits bounded the  ;

increase in delta-CPR in all cases. l 1

REASON FOR CHANGE: QDR 190-92 documented a deficiency [

i in the SPC procedure used to perform the Feedwater  !

Controller Failure Analysis (FWCF). As a result of the i deficiency, SPC reanalyzed the FWCF event for the [

Cycle 5 and 6 reloads. +

SAFETY EVALUATION: The results of the FWCF analysis indicated a slightly higher delta-CPR than was I previously analyzed. These delta-CPR values remain l within the approved Technical Specification operating limits. The analysis utilized previously approved codes and methodologies. The FWCF analysis is no more +

i limiting than the previously analyzed generator load l reject analysis and therefore no potential impact on i i the previously evaluated safety of the plant exists. [

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attachment to GNRO-93/00001 i

l SRASN: NPE-92-024 DOC NO: MCP-92/1033 l l

DESCRIPTION OF CHANGE: Minor Change Package f (MCP) 92/1033 replaces the existing asbestos packing  :

witn graphite packing on certain motor operated valves I (MOVs) in the Reactor Core Isolation Cooling (RCIC), {

Nuclear Boiler, and Reactor Water Cleanup (.RWCU) ,

systems (Valve Nos. Q1E51F063, F064, Q1B21F005, l l Q1G33F001, F004, F100, F101, F102, F106, F250, and l t

F251). Also, the conventional packing system on the i RCIC and RWCU MOVs is replaced with a live-loaded i i packing system. The live-loaded packing system is the l same as the conventional packing system with the  !

exception that Belleville type washers are installed'on i the packing gland bolts to help the packing resist i leakage. This represents a change to the statement in UFSAR 12.1.1.3 that live-loaded valve packings are not  ;

used to reduce the leakage of contaminated coolant from the primary system at Grand Gulf. l" t

F REASON FOR CHANGE: The identified valves have exhibited excessive packing leakage and have frequently i required backseating to reduce leakage. Regulatory  :

1 Guide 8.8 recommends the use of live-loaded valve  !

packings and bellows seals to reduce leakage of  ;

l contaminated primary system coolant. Electric Power ,

! Research Institute (EPRI) NP-5697 reports that " ...

flexible graphite valve stem packing has demonstrated r superior stem seal performance, particularly when ,

combined with live-loading systems that maintain  !

adequate packing gland pressure" and recommends the use [

, of improved packing systems.  :

i  !

t SAFETY EVALUATION: The change will not require  ;

revision of any Technical Specification nor reduce the -

margin of safety as defined in the basis for any Technical Specification. There will be no deviation from the maximum isolation time requirements of Technical Specification Table 3.6.4-1. Since an l

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improved performance for live-loaded graphite valve packing over conventional asbestos packing has been  ;

demonstrated by testing (EPRI NP-5697), the change will l not increase the probability of an accident previously evaluated in the SAR. No new failure modes will be i' introduced. Therefore, the change will not increase the consequences of an accident previously evaluated in i the SAR. The probability of a malfunction of equipment j important to safety previously evaluated in the SAR  ;

will not be increased since the new packing system is  !

equivalent to, or better, than the conventional system. >

Packing failure of either system would not prevent the  ;

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Attachment to GNRO-93/00001 i

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NPE-92-024 Page 2 i

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valve from performing its safety function so the change ,

does not increase the consequences of a malfunction of I equipment important to safety previously evaluated in  ;

the SAR. The release of contaminated primary system  !

coolant is the only credible failure result of the change. This has been evaluated, so the change does t not create the possibility for an accident of a  !

different type than any previously evaluated in the SAR. Because the.new packing system does not create any new failure modes not shared with the existing i conventional packing system, and the bolting / packing-retaining materials are the same, the change will not create the possibility for a malfunction of equipment important to safety of a different type than any -

previously evaluated in the SAR. ,

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Attachment to GNRO-93/00001 I

SRASN: NPE-92-025 DOC NO: EER-92/6030 l

i DESCRIPTION OF CHANGE: Engineering Evaluation Report l (EER) 92/6030 requested the reperformance of the-stress calculations for the Main Steam Relief Valve (;MSRV) i inlet and outlet flange taking the drilling of the stud  ;

holes for helicoil inserts into consideration. This EER requires main steam relief valves B21F041, B21F047, and B21F051 to be evaluated considering the ASME code i 4 case inquiry response. .

REASON FOR CHANGE: During removal of the Safety Relief Valves (SRVs) from the plant for maintenance, problems are being encountered with the bolting materials for i the inlet and outlet flange. Some bolts cannot be i removed and are having to be cut off to support SRV removal due to possible damage with the SRV flange threads. The cut off bolts must be removed to adequately determine any damage.

SAFETY EVALUATION: Stress Calculation MC-Q1B21-92034,  ;

Rev. O shows that Inconel Helical Coil Insert Threads

• are appropriate for the repair of flange threads of Dikkers MSRVs B21F041, F047, and F051. These meet the i structural and functional compliance criteria defined .

in ASME Code 1983 Edition. The Inconel X-750 Helicoil l inserts are to be installed (inspection, repair and  !

rework) in accordance with the instruction provided in

, Plant Operation Manual 07-S-14-357, Rev. 3. The disposition instructions provided in Material Non-conformance Report (MNCR) 111-85 are still in effect and should be observed when removing, installing, and  ;

} performing maintenance on the SRVs.

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Attachment to GNRO-93/00001 SRASN: NPE-92-026 DOC NO: Calculations C-A-632, i Rev. O, etc.

DESCRIPTION OF CHANGE: Calculations for Probable ,

Maximum Precipitation (PMP) floodwater elevations in the plant area have been revised.

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. REASON FOR CHANGE: Calculations for PMP floodwater elevations in the plant area have been updated to incorporate current site conditions, including the removal of some impediments to flow, and addition of the protected area paving.

SAFETY EVALUATION: Calculations for PMP water surface elevations are performed in a step-wise manner starting  ;

downstream and proceeding upstream. Since the effects  ;

of obstructions to flow diminish as the calculation progressed upstream, changes which could affect l 4 intermediate results may have little or no impact upstream. The UFSAR currently lists the results of intermediate calculations. No flood protection commitments are tied to these calculated water levels ,

1 obtained from intermediate calculations. By deleting j intermediate results unnecessary UFSAR revision can be avoided. Thus, deletion of intermediate results and

• changes to UFSAR 2.4 for clarification do not: alter calculational methods or results for PMP water levels near the power block; relax 6" freeboard requirements on PMP flood barriers, or change the level of flood protection provided by those barriers. -

1. A minimum of 6" of freeboard exists on all PMP l i flood barriers;  !

2. For areas west of the power block, the PMP water surface elevation versus duration curve is bounded by a previously analyzed curve;

3. For areas east of the power block, the PMP water surface elevation versus duration curve results in leakage volumes and flow rates which will not invalidate the conclusions reached in the original in-leakage analysis.

The probability of malfunction of equipment inside the power block has not increased since (1) PMP water  ;

levels and durations west of the power block are ,

bounded by existing analyses; (2) PMP water levels and l durations east of the power block result in leakage volumes and flow rates which will not invalidate the conclusions reached in the original in-leakage analysis.

[30]

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Attachment to GNRO-93/00001 NPE-92-026 Page 2 Malfunction of the door seals has already been ,

considered, as has the malfunction of safety related equipment inside the power block due to leakage of floodwater into the plant. The effects of PMP flooding are bounded by existing analyses, and at least 6" of freeboard exists on all PMP barriers.

Revision of these calculations will not affect structural considerations such as roof ponding and will impose insignificant hydrostatic loads on plant area exterior walls. Sections of the UFSAR which have been revised do not delete flood protection commitments or ,

alter flood water elevation calculational methods.

l UFSAR Table 1.4-32 contained results from intermediate

, calculations-pertaining to Culvert No. 1 and other plant area culverts. No flood protection or calculation commitments are tied to these intermediate results. Deletion of this inforeation (including Table 2.4-32) and changes made to UFSAR Section 2.4 for clarification do not: alter calculational methods or  !

results for PMP water levels near the power block; relax 6" freeboard lequirements on PMP flood barriers; or change the level of flood protection provided by those barriers.

Therefore, the margin of safety for Technical l Specification 3/4.7.10, will not be reduced.

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Attachment to GNRO-93/00001 SRASN: NPE-92-027 DOC NO: MECH. STD. MS-25, REV. 6, APPENDIX J DESCRIPTION OF CHANGE: Mechanical Standard MS-25, Rev.

6, Appendix J contains the values for Maximum Expected Differential Pressure (MEDP) for various safety-related ,

valves which were used as inputs into calculations for required valve thrust calculations. In some cases, I these MEDP values differ from MEDP values listed in the i UFSAR.

Appendix J also contains valve limiting component stress allowable thrust (LCSAT) values for Generic Letter 89-10 program Motor Operated Valves (MOVs). l REASON FOR CHANGE: NRC Generic Letter 89-10 '

(Safety-Related Motor-Operated Valve Testing and  ;

Surveillance) discusses the NRC assessments of the t i reliability of all safety-related MOVs based on extrapolations of then currently available results of testing performed in response to Inspection &

Enforcement Bulletin (IEB) 85-03. As a result of these assessments, the NRC determined that failure of safety-related and " position changeable" valves would occur i much more often than had previously been estimated.

Generic Letter (GL) 89-10 therefore provided a number

' of recommended actions for licensees to perform ~for r improving the reliability of the applicable MOVs. Due to increases in the conservatisms in the methodology for calculating required valve thrust values (notably the requirement for the use of a valve factor of .5 i rather than .3 for gate valves), a more realistic MEDP value must be used rather than the original bounding -

, (i.e., conservative) valve MEDPs as stated in the original design specifications and/or UFSAR. The MEDP-values were determined by Nuclear Plant Engineering i (NPE) calculations for compliance with the requirements of NRC Generic Letter 89-10 (Safety-Related Motor- ,

1 Operated Valve Testing and Surveillance). These calculations evaluated each MOV on a case-by-case basis for accident or operator mispositioning scenarios identified as " worst-case" by engineering review.  :

i Calculations have also been performed to determine the valve stem thrust that will maintain components of the valves within the stress allowable values for the components.

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Attachment to GNRO-93/00001 NPE-92-027  !

Page 2 3

SAFETY EVALUATION: Use of the MEDP values contained in

, MS-25, Rev. 6, Appendix J that differ from the values identified in the UFSAR will have no adverse effect on plant safety. The MEDP values in MS-25, Rev. 6, Appendix J are all based upon conservative individual  !

calculations and therefore represent conservative bounding values. Stress allowable limits in MS-25, Rev. 6 have been demonstrated by calculation to be acceptable for all valve components.

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Attachment to GNRO-93/00001 4

SRASN: NPE-92-028 DOC NO: CN-92/0053 DESCRIPTION OF CHANGE: This change was issued to I replace the two pen Bailey model 771 (modified)  !

recorders B21R615A,B (Reactor Pressure Vessel Fuel j 4

Zone / Shutdown Level Recorders) with three pen Bailey  :

model 771 (standard) recorders. This change will j

, instead replace the Bailey 771 recorders and the shelf t units in which they are mounted with L&N model 136 ,

i recorders. The Bailey 771 recorders require 24 Vdc power. The existing B21R615A,B recorders receive Engineered Safety Feature (ESF) power from 120 Vac  ;

Uninterruptible Power Source (UPS) power panel fuse 08-1Y89-15 (08-1Y84-15) via a 24 Vdc power supply. The L&N 136 recorders require 120 Vac power. The  ;

replacement B21R615A,B recorders will receive ESF power  !

from 120 Vac UPS power panel fuse 08-1Y89-06 (08-1Y84-06). The load on each of the new circuits is assumed to be increased by 41 Va. This includes an increase of 26.4 Va for the new L&N 136 recorders and a

, 14.5 Va increase as a result of MCP-91/1090, i

REASON FOR CHANGE: MCP-91/1024 was issued because it 1

was thought that the modified 771 recorder was not  !

available. CN-92/0053 was issued because the lead time  !

for the standard Bailey 771 recorder is excessive (26 [

weeks) and it is believed that Bailey is about to

, discontinue this model. ,

SAFETY EVALUATION: The Reactor Pressure Vessel (RPV)  !

Fuel Zone / Shutdown Level Recorders B21R615A,B are i' required for Regulatory Guide 1.97 Category 1 Control

' Room indication. The function of the recorder loops has not been changed. The ranges of the recorders have 7 not been changed. The functions of B21 (Nuclear Boiler  !

System) and other plant systems are not changed. The ,

changes of this CN will not compromise any existing safety related system, structure or component nor will they prevent safe reactor shutdown. This design change has been approved by the NPE Human Factors Reviewer.

The new recorders are Class IE and seismically qualified. The 41 Va increase to the 744 Va load on i

the two new circuits is almost negligible. The 20 amp fuses in these circuits are still acceptable. Also, the assumed loading of the inverters 1Y87 (1Y88) in the >

. Div I (Div II) battery load calculations is not i exceeded. This change will therefore not introduce an unreviewed safety question.

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i SRASN: NPE-92-029 DOC NO: CN-92/0023 to ,

MCP 88/1051 l DESCRIPTION OF CHANGE: An equipment drain filter and j

, floor drain filter are utilized in the liquid radwaste  ;

systems at GGNS for the treatment of equipment and  !

floor drain wastes. These filters are precoated with a filter aid material via the radwaste precoat pump and l precoat addition tank. The modifications provide for  ;

the addition of an air operator on a plug valve in the i precoat supply line and the addition of an air actuated plug valve in the precoat return line to the precoat

addition tank. This change also provides for the t installation of needle valves to regulate the

opening / closing rate of valves. t i

Finally, this change will modify the Modicon controller  !

logic so that the equipment drain filter and floor ,

drain filter valves will not open at the beginning of Step 8 in the filtration cycle unless the respective body feed pump is running. The time delay for these

, valves to remain open will be changed from 25 seconds 1 to 5 seconds.

REASON FOR CHANGE: To provide a means of eliminating [

the precoat recirculation supply and return line  !

drainage and the consequential hydraulic transient upon l precoat pump start. Also, this change prevents the  !

fast closure of certain radwaste process valves.  !

The Modicon logic change allows for the compensation for the decrease in the Radwaste System volume j attributed to the operation of the body feed pumps and i the elimination of frequent makeup from the condensate  ;

. and refueling water storage and transfer system.

SAFETY EVALUATION: Updated Final Safety Analysis

! Report (UFSAR) Section 3.2 classifies the radwaste ,

systems and all their components as "Other" meaning i that loss of system function would not affect safe

shutdown of the plant. Per UFSAR Table 3.2-1, the l

} radwaste systems are considered non-Q, non-safety  !

related, non-seismic, and NRC Quality Group D. The i modifications made by this change will not impose a  ;

change to these criteria for the liquid and solid ,

! radwaste systems. Furthermore, the postulated worst {

l case failures (radwaste tank rupture and piping leaks) j analyzed in UFSAR Sections 15.7.2 (release to  ;

atmosphere) and 15.7.3 (release to ground water) l

envelope the occurrence and consequences of postulated. i i accidents due to any failure associated with these modifications.

[35]

Attachment to GNRO-93/00001 J

SRASN: NPE-92-030 DOC NO: CN-92/0295  !

. l DESCRIPTION OF CHANGE: The objective of this change is to cap several condensate cleanup post strainer flush  !

lines. The post strainer outlet flush lines will be ,

cut and capped just downstream of the appropriate ball i valves. The abandoned post strainer outlet flush lines  ;

to the outlet header will be removed and the header  !

will be capped at each flush line branch connection.

, The post strainer inlet flush lines will be cut and capped just upstream of the appropriate ball valves.  :

The 4-inch inlet header will be capped at each flush line branch connection. Also, the reach rods for 2 valves will be removed.

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REASON FOR CHANGE: Leakage past valves in the post  ;

, strainer flush lines contributes significantly to  :

radwaste. After being rebuilt, the ball valves leak by  :

the seats at a total rate of approximately 17 gpm.

During the transfer of resin, a significant amount of  :

water is added due to this leakage which results in an  !

escalated volume of radwaste to be processed. Leakage  !

through the ball valves also tends to pressurize the i resin separation and cation regeneration tank consequently hindering the performance of ultrasonic resin cleaning. ,

SAFETY EVALUATION: The UFSAR states that the i condensate system provides no safety function. The  !

system analysis has shown that a failure of the system j will not compromise any safety-related systems or i prevent safe shutdown. According to the UFSAR, the 4

condensate cleanup system is provided with an automatic 3~

bypass to maintain condensate flow in the event a high

differential pressure is realized across the condensate  ;

cleanup system. This design change does not affect any t

technical specification and does not involve an i unreviewed safety question. Furthermore, eliminating }

an ineffective on-line flushing of the post strainers l will not increase the consequences or the probability t of occurrence of the loss of feedwater transient i analyzed in the UFSAR, or any other accident or l 4 transient analyzed in the UFSAR.

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Attachment to GNRO-93/00001 SRASN: NPE-92-031 DOC NO: MCP-88-1037-R00 l

DESCRIPTION OF CHANGE: The Interim Modification and Engineering (M&E) facility is being constructed as part of the overall_" Site Master Plan". The M&E facility  !

will be provided with automatic sprinkler protection  ;

installed by authorized contractors. l t

REASON FOR CHANGE: To provide the M&E facility with i automatic sprinkler protection. l l

, SAFETY EVALUATION: The changes accomplished by this

• modification are consistent with the requirements of .

the original fire protection system design criteria for i materials and construction. Sufficient valving is  :

i provided to limit the effects of a single failure on i the overall fire protection system. No special systems ,

interaction requirements, such as seismic concerns, are applicable to this change. No fire protection system j components shall be required to operate in excess of design limits. Therefore, implementation or performance of the action described in the evaluated document will not increase the probability of .

occurrence of an accident previously evaluated in the i i

UFSAR. l

The UFSAR provides an analysis of safe shutdown in the event of a major fire demonstrating that a single i exposure fire cannot impair redundant safe shutdown- i related components. Connection of the M&E facility ,

fire suppression water system to the plant fire water i system will not affect the postulated frequency,  !

sequence of events and/or systems operation as  !

discussed in the UFSAR analysis. Therefore, the r consequences of an accident previously evaluated in the UFSAR will not be increased. >

i, The piping and attendant components installed comply with pertinent design specifications for material and i construction practices. Valves can effectively isolate j the changes accomplished by the subject document from all fire water suppression systems important to the protection of safety-related systems and/or components.

Consequently, single failure criteria is satisfied.

This change does not impose hydraulic demands on the  :

plant fire water system in excess of existing l

, requirements, does not delete or modify system j protection features, and does not reduce system j reliability. Therefore, the probability of a i malfunction of equipment important to safety previously '

evaluated in the UFSAR will not be increased.

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l Attachment to GNRO-93/00001 l NPE-92-031

, Page 2 The potential failure modes associated with this design >

change are identical to the failure modes postulated ,

I for the remainder of the fire protection underground in

• the original system design. Adequate section control

capabilities presently exist to limit the impact of postulated failures associated.with the new section of

underground piping.

This change does not degrade the ability of the fire protection system to perform its intended function, ,

does not introduce new or different failure criteria i and does not adversely affect or invalidate existing analyses for postulated design basis fires. Therefore,  !

no margins of safety as defined in the bases for any  :

Technical Specification are reduced.

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Attachment to GNRO-93/00001 l

SRASN: NPE-92-032 DOC NO: DCP-87-0005-S00-R01 DESCRIPTION OF CHANGE: A new raised floor and interior partition walls will be erected to provide an atmosphere more conducive to efficient unit operation.

This change provides instructions to improve the overall aesthetics of the Control Room.

REASON FOR CHANGE: Various human engineering deficiencies in the Control Room which were

contributing to inefficient unit operation were I

corrected. This change also provides instructions to improve the overall aesthetics of the Control Room.

SAFETY EVALUATION: The new partition walls and raised [ '

floor will be located in the Control Room to isolate the main controls area and to delineate a hierarchy of  !

command in the Control Room. Due to their location, ,

the walls and raised floor have been seismically l designed to preclude any II/I hazards. These walls and floor are not boundaries of Fire Zone OC503 and are i therefore.not required to be fire rated. However, all material, including insulation, wall coverings, and ,

, floor coverings are non-combustible. These additions are not required to be boundaries of the Control Room '

envelope. Therefore, the probability of occurrence of 3 an accident previously evaluated in the UFSAR will not be increased.

l The new Control Room partition walls and flooring to be f

installed serves to promote more ef ficient unit  :

operation.

The new additions have been verified to be structurally j

} adequate to withstand the loadings associated with j Control Room walls and floors. The boundaries of the Control Room envelope and Fire Zone OC503 remain as defined in the UFSAR and the GGNS Fire Hazards Analysis .

(FRA), respectively; therefore, the walls and floors  !

are not required to be airtight or fire rated. There are no changes being made to any items which mitigate l the consequences of an accident. i These changes will not impact any of the performance  !

characteristics of the Control Room Heating Ventilation i and Air Conditioning (HVAC) System. This floor and  !

t these walls are not boundaries of Fire Zone OC503 and '

are therefore not required to be fire rated.

Therefore, this design change does not impact Technical Specifications 3/4.7.2 or 3/4.7.7.

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[39]

Attachment to GNRO-93/00001

)

1 SRASN: NPE-92-033 DOC NO: MCP-92/1037, Rev. 0- l DESCRIPTION OF CHANGE: This change added 3/4" pressure

(

equalizing lines from between the seats drain to the upstream (suppression pool) side of flexible-wedge gate i suppression pool isolation valves 1E12F004A(B).

REASON FOR CHANGE: This change is being implemented to prevent future occurrence of pressure locked valves.

SAFETY EVALUATION: The only purpose of the installed ,

3/4" vent line is to perform the passive _ function of providing ventilation from the valve internals to prevent pressure locking. The primary containment integrity aspects of these valves remains unchanged by this modification. Credit will be taken for the downstream valve disc to perform the sealing function to isolate the RHR system in the case of excessive leakage. These valves will be tested for leakage in a ,

manner consistent with the FSAR and Technical  !

I Specification requirements. The RHR system will be pressurized with water to a minimum of 11.5 psig (peak l containment accident pressure) with the system totally ,

isolated from primary containment. A leakage rate for the entire system will then be determined and compared to an acceptance limit not to exceed 1 gpm multiplied l by the number of valves in the subsystem tested. l Testing the downstream (outboard) disc in the reverse direction is conservative to testing in the accident -

direction from the containment side because accident -

l conditions tend to seat the disc whereas testing c conditions reduce the seating force, hence tending to unseat the disc. Consequently, this modification does ,

not result in testing which differs from that currently described in the FSAR. It should be noted that the valve internals up to the outboard disc will now be 4

considered part of the containment. Thus, bypassing of  !

a single disc will have no effect on containment  !

integrity and should not be construed as such. This design change will have no affect on the operability of the gate valves or associated RHR system. The design .

has been evaluated against the applicable design  !

criteria, installation, and operational requirements, and all necessary requirements and commitments are met.

Although these valves do not receive a containment  !

isolation signal, the ability of 1E12F004A/B to isolate as required by Technical Specifications is unaffected by this modification. This modification is intended to provide added assurance that the valves will open if ,

required._ Since the isolation function discussed in the Technical Specifications remains unchanged by this l design, all margins of safety remain unaffected.

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Attachment to GNRO-93/00001 SRASN: NPE-92-034 DOC NO: MCP-91/1073, Rev. 0 '

DESCRIPTION OF CHANGE: Replace high pressure core spray (HPCS) diesel generator engine start air pressure l control valve components with different model number  ;

components. .

REASON FOR CHANGE: The presently installed components-  ;

are either obsolete or prone to leak excessively. ,

SAFETY EVALUATION: The installation of'different model number start air pressure regulating valve components (pilot and slave regulators) will provide equivalent replacements for obsolete components no longer manufactured or components prone to leak excessively. .

The replacement components will not increase accident i or malfunction probabilities or consequences. The replacement components will not create any risk of a different type of accident or malfunction and does not reduce a margin of safety as described in the Technical 7 Specificetions bases. The proposed action does not  ;

involve an unreviewed safety question.

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Attachment to GNRO-93/00001 ,

SRASN: NPE-92-035 DOC NO: MCP-91/1088, Rev. 0 )

DESCRIPTION OF CHANGE: Replace the existing air

• operated door latch cylinders on the containment and drywell airlock doors with "like-for-like" new cylinders.  ;

REASON FOR CHANGE: The existing air operated door latch cylinders have become pitted and worn out to a point where the tubing leak test, which is performed each outage and when any modifications are made to the safety related pneumatic air system, cannot be easily passed. The existing cylinders are obsolete and can no longer be obtained for maintenance changeouts.

SAFETY EVALUATION: The replacement of the existing worn out and obsolete air operated door latch cylinders on the containment and drywell airlocks with a new model manufactured by the original vendor will be a "like-for-like" replacement. The new air cylinders will have the same fit, form and function of the old air cylinders and will be qualified to the requirements of the original design specification including pressure and environmental. The containment and drywell airlock safety related pneumatic air system will still perform- ,

its intended function of allowing ingress and egress from the containment and drywell while maintaining isolation / separation of the containment and drywell during normal operating and accident conditions. All of the GGNS Technical Specification Basis and Surveillances will remain unchanged with this modification. This modification will not adversely affect any previously evaluated safety related '

equipment or accidents and will not increase the possible radiation doses which have been previously evaluated.

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Attachment to GNRO-93/00001 SRASN: NPE-92-037 DOC NO: DCP-91/0072 J l

4 DESCRIPTION OF CHANGE: This evaluation jackage will change the fuses that protect the battery feeder circuits 72-11A01 and 11B01. The existing fuses for the battery feeder circuits may not coordinate with the feeder circuits of the battery chargers. The fuses for the battery feeder circuits will provide coordination with all feeder circuits on the 125V DC busses. Also, .

the trip devices (EC-1) for circuit breakers 72-11A01, i 11A05, 11B01, and 11B05 shall be disabled, thereby making the circuit breakers manual disconnect switches.

REASON FOR CHANGE: The existing fuses for the battery feeder circuit and the distribution panel feeder circuits were installed to allow GGNS to eliminate maintenance on the 125V DC bus circuit breakers (72-11A01, 11A05, 11B01, and 11B05). Tnis action item was written pertaining to the failure of GE model AK circuit breakers. The problem was due to mechanical binding which causes the trip coil to remain energized, thus destroying itself. GE attributed the failures to circuit breaker mis-adjustment or lubrication problems.

GE recommended changing lubricants and increasing t maintenance to prevent the problem. However, these circuit breakers are mounted directly to the busses and cannot be detached unless the DC busses are de-energized, resulting in the loss of vital DC loads.

• The resultant change was made to allow the fuses to provide the overcurrent protection function and use the affected circuit breakers as manual disconnect switches only. Subsequent inspection findings determined that the fuse coordination with other DC distribution system overcurrent devices were not conclusive. Also, a -

commitment was made to disable the circuit breaker trip devices to enhance overall reliability. Thus, this change will provide the coordination and protection required by the DC distribution system and will disable the circuit breaker trip device to enhance reliability.

SAFETY EVALUATION: Calculation EC-Q1L21-88003, Rev. 2 provides the support for this change to implement the i modifications previously described. Because the new fuses have greater capacity, coordination between the fuses and the other protective devices shall be achieved. The fuses shall also provide sufficient protection against cable thermal damage for their

, feeder cables on busses 11DA and 11DB. The disabling i of the trip devices (EC-1), for circuit breakers 72-11A01, 11A05, 11B01 and 11B05 will preclude spurious ,

trips of these breakers.

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E Attachment to GNRO-93/00001 l NPE-92-037 Page 2 The installation of the new fuses will not change the -

design function of the distribution centers and independence of the distribution system is maintained by Regulatory Guide 1.75 separation.  !

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Attachment to'GNRO-93/00001 SRASN: NPE-92-038 DOC NO: MCP-91/1035, Rev. 0 DESCRIPTION OF CHANGE: This change replaces the body-to-bonnet bolting and the stem of High Pressure Core Spray (HPCS) motor operated valve with bolts and stem manufactured from higher strength materials than those presently installed.

REASON FOR CHANGE: To prevent any of the valve components from being over-stressed under any operating conditions, including accident conditions.

SAFETY EVALUATION: The valve stem and body-to-bonnet bolting replacement to higher strength materials increases the maximum allowable stem thrust in order to accommodate a higher torque switch setting for the HPCS l valve. The valve logic remains unchanged; therefore, its required operability as an active component as part of the FDCS remains unchanged. In addition, the valve' n.ssive function as a pressure retaining ,

compo- of the HPCS also remains unchanged since the '

pres .aining modification to the valve involves repla wer strength body-to-bonnet bolting with '

highel .gth bolting which will increase the integri of.the bolted joint.

The active functional requirements of the valve do not

• affect accidents evaluated in the UFSAR pertaining to the High Pressure Core Spray (HPCS) system. The only

• event directly caused by the HPCS is inadvertent HPCS injection; however, this is postulated as a result of

• unintended manual pump start which this change has no affect upon. The passive, pressure retaining function ,

of the valve remains consistent with pre-modification >

conditions. Therefore, there is no increase in the ,

probability of accidents nor increase in consequences  ;

of any accident previously evaluated in the UFSAR.

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Reliability of the valve to actively open.and close and passively function as a pressure retaining component are only enhanced by this modification. Therefore, since the valve itself is no more likely to fail, the HPCS system and equipment depending upon HPCS injection are also no more likely to fail.

Based on the safety evaluation, no Technical Specification changes nor unreviewed safety questions. I are identified as a result of this modification.

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[45]

Attachment to GNRO-93/00001 SRASN: NPE-92-039 DOC NO: MCP-91/1066 DESCRIPTION OF CHANGE: The emergency start lockout setpoint on the Divisions 1 & 2 diesel generator starting air systems is being changed from 160 psig to 120 psig.

REASON FOR CHANGE: To ensure the diesel generator will start upon receipt of a Loss of Coolant Accident (LOCA) initiation signal when the starting air receivers are at the Technical Specification minimum allowable pressure.

SAFETY EVALUATION: No Technical Specification change is required due to this modification. The diesel generators are designed to automatically start and achieve rated frequency and load within 10 seconds after receiving a start signal. Furthermore, the diesel generator starting air system is designed to provide at least five starts without recharging the receivers. Finally, the engine is_ designed to operate for seven days with control air pressure supplied by the starting air receivers without recharging the receivers. These design features are maintained and are not adversely affected by this change. Lowering the emergency start lockout setpoint will not increase the probability or consequences of an accident previously evaluated in the SAR. This is because the safety function of the Divisions 1 & 2 diesel generators to start and achieve rated frequency and load within 10 seconds is not affected. No new failure modes are introduced by this change, therefore the probability of occurrence of a malfunction of equipment important to safety is not increased. The consequences of a malfunction of equipment important to safety is not increased. The consequences of a malfunction of equipment important to safety previously evaluated in the SAR are not increased by this change. This is because lowering the emergency start lockout setpoint will not prevent any safety system from performing its accident mitigation function, and the effects of any malfunction of equipment important to safety would be unchanged. This change does not create the possibility for an accident of a different type than any previously evaluated because no change is being made to emergency diesel generator accident response and operation.

[46]

Attachment to GNRO-93/00001 NPE-92-039 Page 2 Also, no new failure modes are introduced by this change in setpoint. The possibility of a malfunction of equipment important to safety different than any previously evaluated in the SAR are not created because there is no change to the design of the start logic and ,

operation of the emergency diesel generators other than lowering the setpoint at which an automatic start is terminated on a LOCA signal. Finally, the margin of safety as defined in the basis for any technical specification is not reduced since the safety function of the Divisions 1 & 2 diesel generators to start and 1 achieve rated frequency and load within 10 seconds is not affectr,d. Also, at least five starts are provided a without recharging the receivers. Finally, the engine will operate for seven days with control air pressure supplied by the starting air receivers without recharging the receivers.

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Attachment to GNRO-93/00001 l

1 SRASN: NPE-92-040 DOC NO: MCP-91/1080, Rev. O i

DESCRIPTION OF CHANGE: Reroute air bubbler tubing for instruments 1N19-LT-N086A(B) (C) (condenser neck expansion joint seal water level transmitter) .

REASON FOR CHANGE: The present level transmitter bubbler path is shared by pipe also used by a float type level switch [1N19-LSL-N022A(B) (C) ] . The shared pipe results in a level switch operation at a level different than the calibrated setpoint.

SAFETY EVALUATION: The rerouting of the bubbler tubing will isolate instruments 1N19-LT-N086A(B) (C) and 1N19 -LSL-N022A (B) (C) from each other, permitting both instruments to perform to their original design requirements. The rerouted tubing will not increase accident or malfunction probabilities or consequences.

The replacement components will not create any risk of a different type of accident or malfunction and does not reduce a margin of safety as described in the Technical Specification bases. The proposed action does not involve an unreviewed safety question.

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~ Attachment to GNRO-93/00001 l l

SRASN: NPE-92-041 DOC NO: DCP-91/1034, Rev. 0-DESCRIPTION OF CHANGE: This change eliminates the potential interface between the Fire Protection Water ,

System and the Construction Water System by removal of a short section of pipe. The Unit 1 Fire Protection >

Water System will not be affected by this design change.

REASON FOR CHANGE: To remove the valved interface between the Fire Protection Water System and Construction Water System.

t SAFETY EVALUATION: This change removes a section of incomplete Unit 2 fire protection piping.

This change does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously ,

evaluated in the SAR because: 1) This change does not l affect fire protection systems or components required for Unit 1; 2) The piping modifications shall be accomplished in accordance with the specifications and standards originally specified for the Fire Protection  ;

System which ensures the integrity of the affected piping; 3) No structural, seismic, missile or other j hazards are introduced.

This change does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the SAR because: 1)

Elimination of the valved interface between the Unit 1 '

Fire Protection System and the incomplete Unit 2 Fire Protection System will not introduce new or different failure modes; 2) The integrity of the Unit 1 Fire Protection System and the Construction Water System is not adversely affected.

This change does not reduce the margin of safety as  !

defined in the basis for any technical specification because the Unit 2 Turbine Building does not contain .,

Unit 1 safety-related equipment and the proposed change  :

will not adversely affect the ability of the Fire 3 Suppression System to perform its design bases function.

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I Attachment to GNRO-93/00001 i

SRASN: NPE-92-042 DOC NO: DCP-91/0030, Rev. O DESCRIPTION OF CHANGE: This change removes the prelube '

system on the radial well pumps. The purpose of the prelube system is to lubricate the rubber bearings on the pump shaft that are above the water column prior to J the pump starting. Once the pump is started, the pump supplies its own lube water up through the pump column to lubricate those bearings.

REASON FOR CHANGE: The prelube system is a high maintenance item and the location of the radial wells allows for limited access during high river levels.

Rebuilding the radial well pumps is planned every 12 to ,

18 months, on a permanent rotating schedule, depending ,

on plant conditions and river level. During pump '

rebuilding, all of the bearings will be replaced so that any bearing that might experience unusual wear as  !

a result of the pump starts will be replaced.

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SAFETY EVALUATION: The radial well pumps are normally I in continuous operation during which time their bearings are self-lubricated. The only time the prelube system performs any function is during infrequent radial well pump starts. The time for the l pump to lubricate the line shaft bearings is about 10 s seconds. This should not attribute to excessive wear .

on the bearings that are above the water level. An increase in bearing failure rate as a result of this change is not anticipated. The preluhe system only affects the radial well pumps and does not affect any system or component important to safety. The Standby Service Water (SSW) basins, which receive their makeup water from Plant Service Water (PSW) and are maintained j in accordance with Technical Specification 3.7.1.3 are i sized to provide post-IOCA cooling for 30 days without external makeup. All other systems which are normally supplied by PSW and are required to operate during an  :

accident are supplied by SSW during an emergency. ,

The design change will not affect any present technical specifications nor does it impose any new technical l specification requirements. The removal of the prelube system from the radial well pumps will not compromise i

any safety-related system or component and will not prevent safe reactor shutdown. The change will not prevent any safety related component or system from .

performing its intended function as previously analyzed nor will it result in an increase in the consequences of any accident.

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[50] j

Attachment to GNRO-93/00001 SRASN: NPE-92-043 DOC NO: DCP-90/0137 DESCRIPTION OF CHANGE: This change provides the design and installation of new security fencing, intrusion detection systems and closed circuit TV (CCTV) systems.

Additionally, this change provides design and installation for the relocation of two fire protection hydrants and for the addition of a new fire protection hydrant.

REASON FOR CHANGE: To upgrade the south perimeter camera coverage and intrusion detection. To enhance ~

fire protection.

SAFETY EVALUATION: All subject equipment is non-safety related. All cabling and raceway modifications to be performed will be in accordance with the separation requirements of Regulatory Guide 1.75 and where required, seismic supports have been provided to preclude the creation of any seismic concerns.

No accident is currently postulated in the UFSAR as a result of the fire mains. In addition, the relocation of existing hydrants and the addition of a hydrant within the protected area will not cause the fire water supply to be unable to meet the anticipated fire water demand. The excavation /backfilling activities will be performed in accordance with the original specifications and design criteria. Final site grading requirements have been provided to ensure probable maximum precipitation (PMP) drainage is not adversely affected.

This change will not increase the probability or the consequences of any accident evaluated in the UFSAR, does not create the possibility of a new accident or malfunction, and does not reduce any margin of safety defined in any Technical Specifications.

[51]

Attachment to GNRO-93/00001 1

SRASN: NPE-92-044 DOC NO: MCP-90/1036, Rev. 0 '

4 DESCRIPTION OF CHANGE: The RCIC Turbine Exhaust Drain Pot Level Annunciator will be modified such that a time delay will be provided to allow for normal operation of i I

RCIC Steam Exhaust Drain Pot Isolation Valve E51-F005 with an acceptable leakage rate through RCIC Steam

< Supply Valve E51-F045 without annunciation in the main .

Control Room. Alarm window L620 will be utilized and renamed as "RCIC TURBINE EXHAUST LINE DRAIN  :

MALFUNCTION".

REASON FOR CHANGE: Based on plant experience and t engineering design, the RCIC turbine exhaust drain pot i level annunciation requires modification. At the present time, due to the leakage of valve E51-F045, whenever the drain pot enters the draindown cycle, the  !

alarm is activated. If the leakage through F045 meets l

or exceeds the drain rate, the possibility exists that l the water will back up into the turbine exhaust line or  !

F005 will be continuously open. Both of these conditions are undesirable. A time-delayed alarm will l allow nornal operation of the drain pot to occur and i normal operation of valve F005 without operator  ;

assistance.

SAFETY EVALUATION: These changes will not compromise i any existing safety related system, structure or i component nor will they prevent safe reactor shutdown.

j No evaluated accident is predicted by a failure of the affected relays or annunciator window. The failure of i

any added relays will not initiate any evaluated i transient or accident. The E51 (RCIC) system operation j and function will not change due to addition of the Class 1E relays and annunciator logic changes.

The relays or the annunciator are not required to ,

, mitigate the consequences of any evaluated transient or accident. No new unbounded interfaces are created and

! no new failure modas are introduced. These changes will therefore not introduce an unreviewed safety ,

question. The annunciator window is not currently (

, addressed in the Technical Specifications and these  !

! changes will not require that they be added to the  :

I Grand Gulf Technical Specifications. Existing RCIC i Technical Specifications are not impacted by this

change.  !

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i Attachment to GNRO-93/00001  !

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i SRASN: NPE-92-045 DOC NO: DCP-91/0026-3,  !

Rev. 0 l

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DESCRIPTION OF CHANGE: This change deletes the high  !

negative D/P alarm portion of the Fuel Handling Area l Pressure High-Low Annunciator 1T42-PDA-L608. Also, the t Fuel Handling Area Pressure Signals Difference High [

Annunciator (1T42-PDAH-L609) will remain operable to j alert the operator of problems in one of the signals i which input to the L608 annunciator. This change also t

installs a time delay relay in the alarm circuit to .

eliminate any spurious alarms due to the temporary  !

decrease in the negative D/P. ,

. REASON FOR CHANGE: To prevent the spurious alarming of ,

the Fuel Handling Area Pressure High-Low and Fuel i

Handling Area Pressure Signal Difference High annunciators.  ;

SAFETY EVALUATION: This change deletes a portion of I

{ the input to and adds a 2 minute time delay to the l remaining portion of an annunciator in the Fuel Handing i Area Vent System (T42) which is not required for system operation. This annunciator is not required to support j the safe shutdown of the reactor or to perform in the l operation of reactor safety features. These changes do {

l not prevent any equipment relied upon to mitigate the l consequences of any evaluated transient or accident L from performing its safety function. No existing i design functions for equipment important to safety are i affected by this design change. These changes do not ,

J prevent any equipment relied upon to mitigate the  !

consequences of a malfunction of equipment important to .

safety from performing its safety function. Therefore,  !

the consequences of a malfunction of equipment  !

important to safety previously evaluated in the UFSAR l is not increased. This change is not required for safe shutdown or accident mitigation as evaluated in the }

UFSAR. This annunciator is not essential in monitoring i the plant for compliance with any Technical Specification.  ;

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Attachment to GNRO-93/00001 I l

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SRASN: NPE-92-046 DOC NO: MCP 90/1033, Rev. 0 ,

DESCRIPTION OF CHANGE: Presently-position switches  !

1E30-ZS-N005 and 1E30-ZS-N006 monitor the position of '

the upper containment fuel pool / transfer canal gate and -

upper fuel pool / reactor cavity gate respectively. ,

Should the reactor mode switch be placed in the RUN position and either of the above gates installed, I annunciator E30-ZA-L605, " Fuel Pool /XFER Gate in Place" will alarm. Also computer points E30-N005 and/or  !

E30-N006 will alarm any time these gates are installed.  !

, I This change eliminates the inadvertent alarming of ,

annunciator " Fuel Pool /XFER Gate in Place". This j change eliminates the function of position switches  :

1E30-ZS-N005 and 1E30-ZS-N006 and therefore deletes annunciator " Fuel Pool /XFER Gate in Place" and computer l points E30-N005 and E30-N006. 1

, REASON FOR CHANGE: An engineering evaluation l identified annunciator " Fuel Pool /XFER Gate in Place" ,

4 as falsely alarming. The cause was identified to be  ;

i the gate position switch contacts shorting, as a result i of being located under water, causing this annunciator  !

to inadvertently alarm. l SAFETY EVALUATION: Existing administration procedures  !

! prohibit operation in mode 1, 2, or 3 with the upper containment pool gates installed. These procedures  !

verify that the upper containment pool gates are in ,

their stored position upon startup of the plant, will l

! not allow installation of these gates during reactor mode 1, 2, or 3 and verify at least once per 31 days [

that these gates are in their stored position. ,

Administration procedures allow installation of these i gates only during reactor modes 4 or 5. The plant

) enters modes 4 or 5 on a planned basis (refueling) less I

frequently than once a year. UFSAR Section 7.5.1.3

states in part that " operations occurring once a year e or.less frequently which could impair engineered safety J i

d feature system performance, are controlled by

administrative procedures". Administration procedures j i

are in place to adequately control the pos. tion of the  ;

upper containment fuel pool / transfer canal gate and  !'

1 upper containment fuel pool / reactor cavity gate and

• therefore deletion of annunciator " Fuel Pool /XFER Gate 4 in Position" and computer points E30-N005 and E30-N006  !

will not require a change to the Technical  !

l Specifications or create an unreviewed safety question. l 1

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[54]

i Attachment to GNRO-93/00001 l l

SRASN: NPE-92-047 DOC NO: DCP-90/0005-1,  ;

Rev. O i I

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DESCRIPTION OF CHANGE: This change provides the design  ;

requirements for all the necessary air accumulator  !

supports, piping modifications, valve replacements,  !

removal of Q1B21F039 stop check valves and removal of ,

one relief valve on each set of Automatic  ;

Depressurization System (ADS) accumulators associated [

with the replacement of existing carbon steel ADS and non-ADS accumulators with stainless steel accumulators.  ;

i REASON FOR CHANGE: This change provides a permanent l solution to the degradation of coatings inside the ADS

, air accumulators by replacing the carbon steel ,

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accumulators with stainless steel accumulators.

I SAFETY EVALUATION: The change does not result in any ,

operational or functional change to the affected system. The size and location of the ADS accumulators will not change. The material for accumulators, associated piping and drain valves will be changed from carbon steel to stainless steel in order to provide i

. improved corrosion resistance. Replacement of check i valves on the accumulator inlet piping will not affect ,

the design function of these valves. Removal of stop t check valves which are currently locked closed will not  ;

affect the function of the system. Calculations have  !

been performed to show that removal of one out of two '

relief valves associated with each pair of accumulators .

will not adversely affect the pressure relief capacity. .

All piping and support modifications have been designed i in accordance with ASME Section III or ANSI B31.1 code '

requirements as applicable. Therefore, this change  !

will not require a change to the Technical I Specifications and will not create an unreviewed safety .

! question.  !

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Attachment to GNRO-93/00001 l

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l SRASN: NPE-92-048 DOC NO: DCP-89/0089, Rev. O l 1

DESCRIPTION OF CHANGE: The 36 shroud head stud (SHS)  !

assemblies will be replaced with 20 shroud head stud assembly modifications (SHSAMs). Eight SHSAMs were l installed during RF04 and an additional 12 will be installed during RF05. Sixteen bolts and studs will be permanently removed during RF05. Also, upper retainer assemblies will be removed if outage schedule permits. '

If the retainers are not removed existing inspections will continue. This modification does not require a new test or experiment, and does not affect the SAR. L Installation procedures for the SHS will be revised to j reflect the new SHSAM design. ,

i REASON FOR CHANGE: The original SHSs experienced wear  ;

due to impingement by feedwater flow on the bolt shaft.

l This wear can render the locking bolt collar assembly  !

inoperable and weaken the bolt shaft. The replacement  ;

SHSAMs will be located out of the direc* feedwater (

flow, thus negating continued wear and lost parts j concerns. Also, upper retainer assemblies will be t removed as outage schedule permits to prevent possible  !

l long term degradation and loose parts concerns.

l SAFETY EVALUATION: The SHSs are not safety related [

4 components. The installation of 20 SHSAMs and removal l 1 of the remaining 16 studs and bolts does not affect the i

pressure vessel boundary integrity. Engineering  !

analyses have demonstrated that only 16 SHSs are

necessary to satisfy the design specification and the l ASME Section III, Subsection NG code allowables, with a l l maximum of two' consecutive non-functioning studs.  !

Analyses have demonstrated that there will be no  ;

significant shroud head flange distortion (or leakage)  !

if up to two studs in-a-row are removed. Upper retainer assemblies will serve no function after the 1

! modification and their removal will prevent loose parts i from occurring. The proposed change does not increase i I the probability of occurrence or the consequences of an e I

accident or malfunction of equipment important to safety previously evaluated in the SAR because the SHS  ;

assemblies are not safety related, do not perform any  !

safety related function, and are not considered in the t

initiation or mitigation of any of the accidents and/or  !

l transients already considered in the UFSAR. Bypass  !

i leakage (through 16 drain holes in the grid flange below the SHS inserts) will result in an increase in t core inlet enthalpy of approximately 0.04%. For the  ;

current cycle reload analyses, this slight increase in  ;

core enthalpy will have no affect on the calculated i

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1 Attachment to GNRO-93/00001 I

NPE-92-048 f Page 2 l

critical power ratio (CPR)'for the limiting event  !

analysis due to conservatism in the inputs for the fuel analysis. No significant change to normal or accident ,

operation will result. The proposed change does not r create the possibility of an accident or malfunction of a different type than any evaluated previously in the  !

SAR because analyses have demonstrated that a l sufficient number of SHSAMs will be installed to  ;

satisfy the design specification and ASME Section III, l 1

Subsection NG allowables and prevent flange distortion.

Therefore, the assembly of the shroud head and steam separator to the flange will not be affected and will not affect any other safety related function or component. In addition, the proposed modification will reduce the potential for lost parts, thus reducing the l possibility of occurrence of any additional equipment failure. The proposed change does not reduce the margin of safety as defined in the basis for any 2

Technical Specification because the replacement of the i SHS assemblies does not impact the allowable safety .

limits addressed in the Technical Specifications bases.  !

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Attachment' to GNRO-93/00001 i

i SRASN: NPE-92-049 DOC NO: DCP-89/0171-0,. l Rev. O DESCRIPTION OF CHANGE: Relief Valves N1G36F084/F085/  ;

F087 in the Reactor Water Cleanup (RWCU) Filter / l Demineralizer (G36) System have welded end connections.  !

A set of flanges will be installed in the inlet and outlet piping of these relief valves.

As part of the piping analysis for the modification, an  :

effort was made to reduce the number of snubbers in the affected piping, which also includes portions of the Reactor Water Cleanup (G33) System. This change provides the design for the removal of snubbers that .

could be deleted from the piping arrangements. (

(

In some cases, minor compensating modifications in  ;

remaining pipe supports were found necessary to

• naintain the system below allowable stress limits and these are also included in this change.

REASON FOR CHANGE: This modification to install flanges in the piping will decrease the time required to remove and replace the valves for calibration and testing.

Removal of snubbers is desirable since there is a i

personnel time and dose commitment associated with any necessary maintenance or repair of these components.

The reliability of the system is also improved by the decreased chance of snubber drag or lock during thermal expansion.

SAFETY EVALUATION: The addition of flanges will not adversely affect the structural integrity of the associated piping. The flanges installed meet ANSI B31.1 code requirements. The function of the G33/G36 systems is unaffected by the installation of flanges in the relief valve inlet and discharge piping, 4

or the removal of snubbers from the system. The affected portions of the G33/G36 systems serve no safety function except for drywell isolation.

1 The piping affected is ANSI B31.1 and non-safety l related except for a line which serves a drywell  !

. isolation function at penetration 366 and is designed ,

per ASME Section III. This line and the connecting  !

ANSI B31.1 piping have been analyzed as Seismic Category I. The remaining ANSI B31.1 piping designs  !

i meet ANSI B31.1 requirements and are designed as  ;

Seismic Category II/I.

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Attachment to GNRO-93/00001 i

NPE-92-049 Page 2 Removal of snubbers is made possible by the reanalysis of the piping system using the criteria of ASME Code Case N411. Code Case N411 allows the use of higher piping system damping values in the stress analysis and, as a consequence, the need for the pipe motion-limiting function of snubbers is decreased. Code Case l N411 was not available during the original design of i GGNS piping systems and has been subsequently adopted  !

into the UFSAR and other design specifications to allow 4 for its use.

This design change will not affect the function, operation, or performance of any other system. The effects of pipe support load changes on structural  :

attachments have been evaluated and are acceptable.

The piping inside the RWCU heat exchanger room has been analyzed such that no new pipe break locations are .

required to be postulated. Therefore, this change will not create an unreviewed safety question. .

No changes to the Technical Specifications are required  ;

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as a result of this change.

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Attachment to'GNRO-93/00001 ,

SRASN: NPE-92-050 DOC NO: DCP-88/0213,-Rev. O ;

DESCRIPTION OF CHANGE: This change relocates or  ;

deletes various control room annunciators. These  !

annunciators are being relocated within the control room only.

The relocation will involve disconnection and reconnection of contact inputs to the appropriate annunciator logic cabinet, the inputs to computer, and-installation of relays for contact multiplication where re-transmission relay is not available in the annunciator logic cabinet. This change also includes design modification for deletion of annunciators B33-  !

XA-L616 and B33-XA-L617 that resulted due to deletion of auto loop operation of the flow control valves.

4 REASON FOR CHANGE: Detailed Control Room Design Review  !

(DCRDR) was performed to resolve discrepancies identified in Human Engineering Discrepancy Report HED #821 per NUREG-0700. The purpose of this change is to resolve the remaining portion of the human engineering discrepancies documented in HED #821.

SAFETY EVALUATION: This change involves rearrangement- ,

of several annunciator window locations within the i control room. The plant annunciator system does not  !

perform any safety function nor does it directly affect any equipment that performs a safety function, as i physical separation and isolators are utilized by the annunciator system to preclude adverse intersection 1 with Class 1E equipment. This change does not alter the operational design of the annunciator system other than specific window locations, which are being relocated per DCRDR recommendations. Deletion of i functions associated with annunciators B33-XA-L616 and B33-XA-L617 has been evaluated. No other plant systems are adversely impacted by this change. Therefore, this l change does not. introduce any unreviewed safety question. I F

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Attachment to GNRO-93/00001 SRASN: NPE-92-051 DOC NO: DCP-88/0011, Rev. O DESCRIPTION OF CHANGE: The main transformer cooler units control logic will be redesigned to relieve the operators from the responsibility of de-energizing the cooler units. The operators will still maintain the responsibility of energizing the cooler units prior to re-energizing the transformers. The following design objectives will be met:

1. Allow the lead cooler group to be energized prior j to the energization of the transformer. j

2. Remove the lead cooler group from service whenever

, the unit goes off line. ,

3. Maintain existing control logic for the secondary cooler group.  ;

REASON FOR CHANGE: The energization and de- l energization of the main step-up transformers lead cooler units is presently controlled procedurally.  !

During unit scrams, the de-energization of these. cooler l units presents an additional burden to the operators, ,

and could be overlooked. This would result in the i operation of these cooler units while the transformers  !

> are de-energized, potentially leading to static ,

(streaming) electrification within transformer oil and i resultant catastrophic failure of the transformer i units.  !

SAFETY EVALUATION: This change does not represent a ,

i change in information, operation, function or ability i to perform a function as presently described in the l

UFSAR. The functional changes will enhance overall ,

transformer reliability by reducing the potential  ;

occurrence of static electrification. These changes l will also reduce required operator response during  !

post-trip periods and will therefore potentially i enhance overall plant operation. The changes do not  !

! affect any equipment required to perform safety functions in response to plant accidents and transients f as previously evaluated. No new accident or transient  !

initiating event contributors are introduced by these ,

changes. Therefore, this changes does not introduce any unreviewed safety questions. l I f s

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Attachment to GNRO-93/00001 SRASN: NPE-92-052 DOC NO: DCP-88/0087, Rev. 0

)

DESCRIPTION OF CHANGE: A tritium tracer is added to i the primary water to allow detection of primary water (

, leakage within the generator. This function is j performed by a tritium flushing tank that is inserted  !

in the primary water circuit as a recirculation line  !

around the leakage water return pump. The Chemistry l department takes samples of the primary water to determine tritium concentration. Whenever an addition of water to the Primary Water (N43) system is required, tritium must be added to maintain a proper concentration. At present, tritium is added by  ;

isolating the tritium flushing tank and removing the  !

top from the tank.  !

) This change will add vent and injection piping i

(including isolation valves) connections on top of the ,

tritium flushing tank. Also, this change will add j drain piping connection (including isolation valve) to  :

the tank inlet piping between two isolation (N43FA53 & l 3

N43FA54) valves, which is routed to the chemical  !

radwaste (CHRW).  !

REASON FOR CHANGE: This change provides removal of water from the tank by opening the valve in the drain i piping and addition of tritium to the tank by opening  ;

the valve in the injection piping. This modification i will eliminate removal of the top from the tank. Also,  !

this change will minimize personnel exposure to the water containing the radioactive isotope (tritium).  !

1  !

l SAFETY EVALUATION: The tritium flushing tank and ,

associated components are part of the Generator Cooling i Water system (N43). This system is non-safety related and non-seismic. The system is not required for safe shutdown of the plant. This change to the tritium ,

flushing tank does not affect any other safety-related L i

systems. The change shall be in accordance with ,

l ANSI B31-1, Power Piping Code, 1973 Edition through l Winter 1974 Addenda. The N43 system is not explicitly  ;

described in the SAR. There are no unreviewed safety  ;

l questions.  ;

l  !

[62]  !

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Attachment to GNRO-93/00001 l I

l SRASN: NPE-92-053 DOC NO: DCP-82/0056-1, Rev. O i DESCRIPTION OF CHANGE: The objective of this change is to resolve the concerns regarding the relief valve reset pressure on the Division I and II standby diesel generator starting air storage tanks. The objective will be accomplished by replacing the valves with those specified by Energy Services Group, the diesel

! supplier. The change will include all piping and support modifications necessary for the replacement of the relief valves of the starting air storage tanks of ,

the Division I and II emergency diesel generators. The l diesel generator reliability will be enhanced because the valves will ensure that adequate starting air  ;

pressure is maintained in the event of a discharge or i 1 an inadvertent lifting of a relief valve.  :

i k REASON FOR CHANGE: The relief valves used on the  ;

starting air storage tanks of the Division I and II '

emergency diesel generators have blowdown reset pressures of 40-60% of the set pressure. Reset pressures of 40-60% of the 275 psig set pressure result i in tank pressures which are below the 160 psig j emergency start signal lockout. Energy Services Group  ;

has identified this problem as a potential defect in -

accordance with 10CFR21, and has specified a suitable replacement valve for use on the starting air storage [

, tanks. This change replaces the currently installed i valves with the type recommended by Energy Services i Group.

SAFETY EVALUATION: Calculations have shown that the  ;

! new valves and additional piping will have no adverse  ;

impact on the operation or function of the system and t will not affect the structural integrity of the starting air storage tank. The replacement valves, piping and pipe supports were designed to ASME {

Section III and Seismic Category I requirements and  !

will function in their intended manner. The new relief i valve will reset at 10% or less than the set pressure. l The set pressure of the relief valve is 275 psig. The  ;

minimum reset pressure of the valves is 247.5 psig, which is higher than 195 psig required per Surveillance Requirement 4. 8.1.1.2 (a) (7) of the Technical Specification. This change will enhance the .

reliability of the diesel generators by ensuring the

, relief valve reset pressure will maintain the required air pressure in the starting air storage tanks.

Therefore, this change will have no adverse affect on ,

the ability of the diesel generators to start when >

required. No existing system interfaces are affected by this change and no new interfaces are created.

[63]

Attachment to GWRO-93/00001' :l j

NPE-92-053 Page 2 1

This change will not require a change to the Technical I Specifications and will not create an unreviewed safety i question. l t

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I Attachment to GNRO-93/00001 l i

SRASN: NPE-92-054 DOC,NO: DCP-88/0063, Rev. O i

DESCRIPTION OF CHANGE: This change provides the design ,

for the removal of snubbers from the following piping  :

arrangements: l l

1. Reactor Water Cleanup (RWCU) Pump N1G33C001A-A and C001B-B discharge to containment penetration #88, j Flued Head G505. Four snubbers are deleted on four supports. j i

2. RWCU pump discharge from Containment Penetration  !

1. 88, Flued Head G505, to RWCU Regenative Heat {

Exchanger N1G33B001A; from Drywell Penetration

1. 337, Flued Head G506, to RWCU Regenative Heat ,

Exchanger N1G33B001A; to and from anchors on the l piping to filter demineralizers; and to l Containment Penetration #83, Flued Head G502, for [

feedwater from Regenative Heat Exchanger "A", and  !

bypass. Twenty snubbers on twelve supports are deleted.  ;

I

. 3. RWCU pump suction from reactor pressure vessel j drains, RWCU non-regenative heat exchangers, and l Reactor Recirculation Pumps A and B suction to j Containment Penetration #87, Flued Head G504, . j Drywell Penetration #366, and Drywell Penetration i #337, Flued Head G506, including low point drains r to clean radwaste. Thirty-eight snubbers on 1 twenty-seven supports, of which thirty snubbers  ;

are on twenty-two reactor recirculation system  !

supports, are deleted, seven of the thirty-eight j snubbers are changed to rigid supports, one spring l hanger is revised with load setting changes and _i two restraints are modified as a result of this l optimization. Twenty-two snubbers on thirteen j supports remain in this stress problem. j

4. RWCU Pump N1G33C001A-A and C001B-B suction from j Containment Penetration #87, Flued Head G504. Six  !

snubbers on four supports are deleted. l S. Valve stem leak-off lines to drains. Six snubbers on six leak detection system supports are deleted ,

as a result of this optimization.

REASON FOR CHANGE- Removal of snubbers is desirable since there are a large number of them installed at GGNS and they require a significant maintenance and testing effort during refueling outages. Worker dose eduction and potentially improved outage schedules are among the benefits of snubber reduction.

[65]

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Attachment to GNRO-93/00001 NPE-92-054 Page 2 ,

i SAFETY EVALUATION: The modifications made by this change will not affect the system function, operation,  ;

or performance in any way.. The capability of the .

Reactor Water Cleanup System to perform its safety functions, including the prevention of excess reactor -

coolant loss and the prevention of radioactive material released from the reactor, is not affected. The  :

capability of the Reactor Recirculation System, as a  !

result of the reanalyses, to perform its safety l functions, including providing an adequate fuel barrier i thermal margin, a refloodable volume and pressure integrity, is not affected. The Leak Detection System l

was not designed to meet any safety design bases. The '

deletion and modification of supports will not j adversely affect the structural integrity of the ,

i associated piping. The piping, tubing, and their  !

] support designs meet ASME Section III/ ANSI B31.1 t requirements and are qualified as Seismic Category I. ,

In order not to exceed allowables on Inservice Inspection (ISI) Platform P-2, two signs depicting a  ;

live load limitation are to be attached. Therefore, i the piping, tubing and their supports will function in  !

l their intended manner. All equipment nozzle loads are  !

acceptable. No existing system interfaces are affected  :

and no new system interactions are created. l 4  !

The reanalysis of the piping systems has shown that all

^ ASME Section III/ ANSI B31.1 code allowables have been

met and therefore the probability of a piping failure j d

has not increased. The modification will not introduce

)

any new postulated piping failures and the existing hazards evaluations are not affected.

4 No new system interfaces with any equipment have been created and no existing interfaces have been adversely affected. No new failure modes for the system or any equipment have been created.

i By remaining within the same allowables specified by the applicable codes as stipulated for piping, tubing, ,

supports and supporting structures, and with approved i equipment nozzle loads, the margins of safety provided 1 by these allowables are not affected.

No revision is required to Technical Specification i 3/4.7.4, " Snubbers", since individual snubbers are not .

identified and the modifications being made per this }

change will not create an unreviewed safety question.

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[66]  ;

t Attachment to GNRO-93/00001 SRASN: NPE-92-055 DOC NO: CN-91/0208 ,

t DESCRIPTION OF CHANGE: This change installed an Annubar (flow sensing device) in the minimum flow recirculation line for the condensate pumps as  ;

permanent plant equipment. ,

REASON FOR CHANGE: The Annubar-is used for flow testing and calibration purposes and is. installed as -

permanent plant equipment to facilitate future system testing.

SAFETY EVALUATION: As described in the UFSAR, the condensate system serves no safety function and systems ,

analysis has shown that failure of this system will not compromise any safety-related system or prevent safe  :

shutdown. Incorporation of the Annubar as permanent t plant equipment to facilitate future system testing will not affect the function or operation of the condensate system. The piping was designed to ANSI  !

B31.1 code requirements and the pipe support has been determined to be structurally adequate and acceptable for the additional load of the Annubar. Since the condensate system is not required to effect or support  !

the safe shutdown of the reactor or perform in the operation of reactor safety features, incorporation of the Annubar as permanent plant equipment will not create an unreviewed safety question.

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Attachment to GNRO-93/00001 SRASN: NPE-92-056 DOC NO: MCP-91/1059 DESCRIPTION OF CHANGE: This change enhances the operational reliability of the emergency brake on the main hoist of the refueling platform and the fuel handling platform.

REASON FOR CHANGE: This change eliminates the  :

integrally mounted disengagement rod and follower spring, since these items were the cause of the brake failure during Refueling Outage 4. The new removable disengagement rod and tension springs will enhance the reliability of the emergency safety brake on both the ,

refueling platform and the fuel handling platform. ,

SAFETY EVALUATION: UFSAR Section 15.7.4 and 15.7.6 evaluate fuel handling accidents due to the postulated failure of the grapple head, fuel handle or improper grappling of a fuel bundle, for the refueling and fuel .

handling platforns. The modifications to the emergency ,

brake will not affect any of the parameters associated with this previously evaluated malfunction / accident.  ;

Section 3/4.9.7 of the Technical Specifications l addresses load limits for crane travel over the spent ,

fuel and upper containment fuel storage pools. The modifications to the emergency brake have no affect on the allowable load limits for crane travel or affect any technical specification limits for the platform r hoists. The modifications to the emergency brake have  ;

~

been evaluated for Seismic II/I concerns, and other design considerations such as, but not limited to, material composition, environmental factors, and frequency of use. The modifications will comply with the original design criteria and the brake will perform the same function as the existing equipment. Existing accident analyses are not impacted, nor are new accident scenarios postulated. No new failure mechanisms are created either directly or indirectly. ,

Margin of safety as defined in the Technical  !

~

Specifications is in no way reduced by the modifications to the emergency brake.

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I Attachment to GNRO-93/00001 i i

I SRASN: NPE-92-057 DOC NO: MCP-91/1130, Rev. 0 l DESCRIPTION OF CHANGE: The FSAR outlines the requirements for settlement survey for determining j building settlement and tilt. Additional settlement markers shall be installed in the plant to directly measure differential settlement and to monitor seismic gaps between buildings.

REASON FOR CHANGE: On April 16 and 17, 1990, the NRR I Structural and Geosciences Branch of the NRC met with GGNS and toured the facility to discuss, among other things, settlement monitoring. These changes are a ',

result from the discussions with the NRC on GGNS Settlement Monitoring Program.

SAFETY EVALUATION: The changes to GGNS's Settlement Monitoring Program are enhancements and provide direct ,

determination of differential settlement and direct

• monitoring of seismic gaps. Previously, these values  ;

were determined from the survey taken for total building settlement. ,

f The changes provide requirements, methods and data recording requirements for surveying the new markers. ,

The change requires installation of markers to be  !

performed under Plant Quality Inspection to prevent the ,

cutting of rebar and to ensure damage to the concrete walls and floors does not occur and proper placement of the markers is performed. The installation of.these markers will not impose any additional loads or forces  ;

on the buildings and will not interfere with operation  ;

of any equipment. Consequently, this change does not ,

increase the probability of occurrence or the consequences of an accident or malfunction of equipment ,

important to safety previously evaluated in the safety i analysis report.

This change installs markers in locations where they do not interfere with the operation or performance of any equipment. These markers will not infringe upon the seismic gap between structures. Also, no equipment ,

relies upon these markers to perform their function and no new accident initiators are being created. i Therefore, the possibility of an accident or j i

malfunction of a different type than any previously evaluated in the safety analysis report is not created. ,

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3 Attachment to GNRO-93/00001 i

NPE-92-057 i Page 2 These markers are being installed to gather data on .

building settlement. The requirements are addressed in the FSAR. Foundation settlement is not described in,

• nor governed by the GGNS Technical Specifications. i

~

Also, the installation of these markers will not affect any plant boundaries or structures. Therefore, the margin of safety as defined in the basis for any Technical Specification will not be reduced. ,

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Attachment to GNRO-93/00001 SRASN: NPE-92-058 DOC NO: DCP-88/284-1, Rev. O DESCRIPTION OF CHANGE: To determine the actual reactor recirculation pump torque during various operating modes, a pump torque measurement system is to be installed. The system chosen for this purpose is the Bently Nevada Torximitor" Torque Measurement System.

This system is " state-of-the-art" and capable of continuous monitoring of steady state and dynamic torque.

The presently installed C86 (Vibration Monitoring System)/B33 (Recirculation System) instrumentation junction boxes (located inside the pump driver mount) will be relocated so as to not interfere with reactor recirculation pump or seal work.

REASON FOR CHANGE: On three separate occasions, replacement of the reactor recirculation pump rotating elements was required due to cracks in the pump shafts.

On all three occasions through wall cracking occurred on the "B" pump. On the "A" pump cracking was .8" deep in 1989 and in 1990 the cracking was no more than .18" deep. The temperature extremes on the inside and outside diameters of the shaft where cracking occurs creates a severe steady state thermal stress which combined with various mechanical loadings have resulted in through wall cracking without the presence of stress i

risers. A critical piece of data concerning stresses on this area of the shaft is determination of actual (vs. calculated) pump torque. At present there is no installed instrumentation for measuring this parameter accurately.

Also, the presently installed vibration monitoring instrumentation junction boxes interfere with recirculation pump disassembly / assembly and seal replacements.

SAFETY EVALUATION: Installation of the Torximitor"

. torque measurement system on the reactor recirculation pumps will have no impact on plant safety. The Torximitors" are non-safety related parts. The increase in the pump mass and moment of inertia are only .2% and .00127%, respectively and therefore have an insignificant impact on the dynamic loading including seismic analyses related to the pump. The missile hazard presented by the addition of the

Torximitor* rotating component is negligible due to its design (single piece with no bolting or fasteners) and high material strength (approximately 180,000 lb/in2

[71]

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Attachment to GNRO-93/00001 ]

NPE-92-058  !

I Page 2 tensile strength). Design, materials, and construction i of the pump remains basically unchanged from the j original design and therefore the probability and  :

consequences of analyzed accidents and transients, l including shaft failure, are not increased. The i specified performance of the pump with regard to flow, ,

discharge head, critical speed, flow coastdown, etc.

remains unchanged from the original design.  :

i Relocation of the B33/C86 instrumentation junction boxes to outside the recirculation pump motor driver mount will have no impact on plant safety since the new I and/or modified supports are to be either "Q" Seismic ,

Category I or "Non-Q" Seismic Category II/I as appropriate. l i  !

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Attachment to GNRO-93/00001 l

SRASN: NPE-92-059 DOC NO: DCP-88/0284-4, Rev. O I

DESCRIPTION OF CHANGE: This change provides a l recirculation pump lower wear ring to replace that l portion of the pump case (i.e., the pump case ,

surrounding the bottom of the pump impeller) in the l event that inspection determines that the present condition of the pump case is unacceptable. Included  ;

in the scope of this effort is the review of the vendor analyses concerning the acceptability of the new design ,

- notably the revision to the ASME code required pump [

case stress analysis. l REASON FOR CHANGE: In May 1989, December 1990, and f again in December 1991, replacement of the reactor ,

recirculation pump rotating elements was required due  :

to cracks in the pump shafts. . Chi all three occasions '

through wall cracking occurred on the "B" pump. During inspection of the "B" recirculation pump following the i initial occurrence of cracking (May 1989) an imprint-similar in pattern to the threads on the bottom of the  ;

impeller were noted in the pump case at that location  ;

adjacent to the bottom of the impeller (indicating that '

contact had occurred between the bottom of the impeller i and the pump case). During inspection after the third cracking occurrence on the "B" pump (December 1991) it -

was noted that the thread pattern noted during the j first inspection was no longer present and that there t were wear indications along a substantial portion of  :

the pump case adjacent to the bottom of the impeller.

The vendor has stated that increased wear in this area  !

results in increased flow bypassing back to the pump i suction, decreased stiffness, and decreased damping. I Of particular concern is the eccentricity of the wear  !

area. The vendor has provided acceptance limits for  !

maximum allowable wear and eccentricity at this ,

location. The recirculation pumps will be inspected l

during RF05 to determine the present actual condition

! of the pump case adjacent to the bottom of the

, impeller. If the wear or eccentricity exceed the vendor provided acceptance limits then this area of the pump case will be machined and a new wear ring will be  !

installed.  :

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i

__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________________l

I Attachment to GNRO-93/00001 )

i NPE-92-059 Page 2 SAFETY EVALUATION: Installation of the new pump case wear rings in the reactor recirculation pumps will have I

no impact on plant safety. The wear rings are non-pressure boundary parts and the portion of the pump case that will remain after machining (to allow installation of the new wear rings) is sufficient to meet all ASME code requirements. The pump mass is .

virtually unchanged since the wear ring is to replace a  !

portion of the pump case that is constructed of nearly identical material and therefore has an insignificant impact on the dynamic loading including seismic analyses and missile analyses related to the pump.

Design, materials, and construction of the pump remains basically unchanged from the original design and therefore the probability and consequences of analyzed accidents and transients, including shaft failure, are not increased. The specified performance of the pump with regard to flow, discharge head, critical speed, flow coastdown, etc. remains unchanged from the original design.

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Attachment to GNRO-93/00001 SRASN: NPE-92-060 DOC NO: DCP-88/0284-S02-R00 DESCRIPTION OF CHANGE: The presently installed vendor supplied recirculation pump seals (Byron Jackson SU(M) i type) are to be replaced with Atomic Energy of Canada Limited (AECL) CAN8 model seals. The CAN8 seal design is an improved version of the CAN2 seal design which

• has been customized for GGNS service conditions. The CAN2 seal design was based upon the original Byron Jackson SU type seal (which was the predecessor to the ,

SU(M) type seal) that incorporates design improvements such as improved materials of construction, better i deflection characteristics, and increased lubrication to correct the pressure oscillation problems that were i

recurrent with the SU seal design.

REASON FOR CHANGE: The SU(M) seal design had corrected  :

! the pressure oscillation problem by the use of a slotted seal face design. This slotted seal face design has proven to be particularly susceptible to ,

crud induced failures when seal purge flow to the recirculation pumps is secured. Operation of the recirculation pumps with reduced seal purge flow has been recommended by General Electric to aid in mitigation of recirculation pump shaft cracking l problems. The CAN8 seal is designed to be operated either with or without seal purge flow and is fully  :

compatible with the remainder of the recirculation  !

pump.

SAFETY EVALUATION: Replacement of the precent recirculation pump seals (Byron Jackson type SU(M))

with AECL CAN8 seals will have no impact on plant safety. The seals themselves are non-safety related parts. The increase in pump mass is only .04% and therefore has an insignificant impact on the dynamic loading including seismic analyses and missile analyses related to the pump. The CAN8 seals maintain the original clearances between rotating and stationary parts, are of an improved material and configuration design, and are fully compatible with the remainder of the recirculation pump. The specified performance of

, the pump with regard to flow, discharge head, critical speed, flow coastdown, etc. remains unchanged from the original design.

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Attachment to GNRO-93/00001 SRASN: NPE-92-061 DOC NO: DCP-88/0064, Rev. O DESCRIPTION OF CHANGE: This change provides the design and installation of a relocated portion of the 12"-

KSF-2 fire protection underground water main (fire main). The fire main is a component of the Fire Protection System (P64) which is non-safety related.

The existing section of the 12" O fire main will be

. abandoned in-place with capped-off ends to prevent Category I backfill from entering the pipe. A divisional valve located on the abandoned section of the fire main will be replaced by two (2) post indicator valves. e REASON FOR CHANGE: This fire main is being relocated ,

approximately 37'-0" west of its present location cue to close proximity with the west wall of the Interim ,

Modification and Engineering Building (M&E) which was recently constructed as part of the GGNS site master plan projects. This change is being initiated to prevent any structural damage to the building, foundations and floor slab of the M&E facility, should a rupture of the 12" O fire main occur and for accessibility to the 12" 0 fire main for maintenance or repairs.

SAFETY EVALUATION: The ductile iron which this new fire main is composed of is resistant to soil corrosion. Also, this pipe is coated with asphaltic '

coating which provides an added protection against soil ,

corrosion.

• i The relocated 12" O fire main will pass under one and over six Division I, II, or III, Seismic Category I, i

electrical ductbanks. At those locations where the relocated fire main is constructed over or under a  :

safety related electrical ductbank, protection of the  !

ductbank from tornado-generated missiles is provided during the time frame in which the soil cover over the .

! ductbanks is reduced owing to excavation requirements. l

, Protection is provided for these electrical ductbanks by the placement of 1" thick carbon steel plates on top of the electrical ductbanks or at the bottom of the excavation above the ductbanks. The 1" thick steel plates will remain in place after the fire main is installed to provide adequate protection until backfilling is completed.

[76]

Attachment to GNRO-93/00001 i

NPE-92-061 e Page 2

+

Where the relocated fire main is constructed under the safety related ductbank, protection of the ductbank from a pipe rupture or water leak during plant  ;

operation is provided by a 24" O guard pipe placed under the ductbank. The relocated fire main is routed through the 24" 0 guard pipe which, when sealed at its ends, will provide sufficient protection against fire main rupture or leakage.

Portions of the fire main are relocated inside the tieback wall, necessitating excavation of Category I structural backfill. Materials used for backfilling these excavations will conform to the specifications for backfill of Category I structures.

Instructions to safely excavate in the vicinity of ,

safety related items and pressurized lines are provided. The excavation beneath the Division III;  ;

Seismic Category I, electrical ductbank will not '

adversely affect the structural integrity of the ductbank or its functional operability. In addition, the fire main has been evaluated for surface imposed '

load effects due to an H2O highway loading. The excavation of the trench for the fire main considers ,

the impact on adjacent piping and electrical ductbanks.

Final site grading in the affected areas will not ,

adversely affect the probable maximum precipitation (PMP) drainage. l The changes to be implemented per this modification do not inhibit the fire protection water main from providing adequate fire suppression water to confine and extinguish fires occurring in any portion of the facility where safety-related equipment is located, does not introduce new or different failure criteria and does not adversely affect or invalidate existing  !

analysis for postulated design basis fires. The newly installed section of fire main is installed in accordance with the requirements of the same specifications to which the abandoned section of the original main was installed. This change will not create any unreviewed safety question nor does it affect the GGNS Technical Specifications.

[77]

Attachment to GNRO-93/00001 l SRASN: NPE-92-062 DOC NO: MCP-92/1087, Rev. O DESCRIPTION OF CHANGE: This change provides details for the replacement of non-Class 1E Battery 1L3. The replacement cells being used are the serviceable cells remaining after the changeout of batteries 1A3 (Div. I) and 1B3 (Div. II) by another design change. The replacement batteries are type LC-29.

REASON FOR CHANGE: Battery 1L3 requires replacement i because it is approaching the end of its service life.

• This replacement will help to preclude failure of the 125V DC Battery "L" system and the possibility of reduced voltage level outputs that could affect operation of the non-Class 1E equipment downstream of the battery.

SAFETY EVALUATION: An electrical calculation has been performed to verify acceptance of the capacity and voltage ratings of the new battery. The battery being installed as 1L3 is a type LC-29. The ampere-hour rate for the replacement battery will be 2030 at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Battery 1L3 is primarily used to supply DC power to BOP inverters 1Y81 and 1Y97. This replacement is acceptable because it will meet existing requirements for design and operation of the 125V DC system. This installation meets the intent of applicable design standards and does not create any adverse affects to the existing 125V DC system. This evaluation is written to address this replacement and the revision of SAR Section 8.3.2.1.1, Rev. 5 for battery design i considerations, and Figure 8,3-10A, Rev. 3, to reflect the new ampere-hour rating of Battery "L". The SAR will be revised to reflect that the BOP battery meets the intent of the industry standards in sizing to i maintain capacity for end of life discharge capacity.

The battery does not meet the recommended 80% of nameplate rating capacity but is acceptable by 3 evaluation for expected service life. GGNS Technical Specifications do not address Battery "L" installation or operation and are riot af fected by this modification.

[78]

Attachment to GNRO-93/00001 SRASN: NPE-92-063 DOC NO: MCP-90/1070 DESCRIPTION OF CRANGE: This change will remove the twelve (12) stop-check valves from the return lines to the Upper Containment Pool (UCP). The original design utilized these stop-check valves as vacuum breakers to prevent siphoning of the UCP. The vacuum breakers are located on the high points of the return piping.

REASON FOR CHANGE: As a result of the return piping elevation, with the exception of one scenario, the vacuum breakers will not limit the minimum water level ~

established during a draindown or siphoning event to any higher elevation than the system would have if the vacuum breakers were not present. The single scenario that would allow the water level to decrease to a lower elevation has been determined not to be credible and would not uncover the fuel; however, it was addressed in this safety evaluation. Additionally, ten (10) of the twelve (12) vacuum breakers are below the minimum required elevation of the UCP water level. This could present a potential draindown path should the stop-check valve or packing leak. These valves are being physically removed from the system to eliminate the potential problems associated with valve seat or valve packing leakage.

SAFETY EVALUATION: An engineering report and the associated safety evaluation determined that the vacuum breakers do not perform any active safety-related function. As a result of the engineering report and the safety evaluation, the subject stop-check valves have been closed and removed from the testing program.

The passive safety-related function of the vacuum breakers will be retained as the pipe capping or plugging will meet the requirements of the applicable codes for Class 3, Seismic Category I piping. An unreviewed safety question does not exist. A change to the Technical Specifications is not required.

[79]

Attachment to GNRO-93/00001 _

SRASN: NPE-92-064 DOC NO: MCP-91/1044, Rev. O DESCRIPTION OF CHANGE: The objective of the change is to eliminate contamination around the solid radwaste fill ports which results from tank blowdowns. This is to be accomplished by cutting and plugging both 1" HCD-320 fill port vent lines at the connections to the 8" HCD-321 lines of the Radwaste Building Ventilation System. The 1" HCD-320 vent lines will be cut and removed from the T-sections of the 8" HCD-321 piping.

The exposed T-branches will then be plugged such as to minimize potential for crud traps. Thus, the 1" vent lines from the 8" HCD-321 to the fill port flexible ~

hose connections can be physically removed.

REASON FOR CHANGE: Solid radwaste fill ports, G18-D006A and B, are not currently used in radwaste processing. At this time, solid radwaste is processed by a contractor. Vent lines for fill ports G18-D006A and B are connected to the Radwaste Storage Tank Vent System. The basic function of the 1" vent lines is to provide ventilation from the fill ports to prevent overpressurization of the radwaste containers when the mixing tanks are drained. However, as stated above, this equipment is no longer used since radwaste is now processed by a contractor. Thus, the vent function is no longer required. Per procedure, during normal operation, the storage tank vent filter train exhaust fan is idle. The exhaust fan is run only during maintenance on a tank to reduce the amount of exposure to personnel performing maintenance. Since the exhaust fan is not normally in operation, the open fill port vent lines exhaust air and moisture into the room when radwaste tanks connected to the same exhaust system are blown down. This venting has resulted in unnecessary contamination around the fill ports.

SAFETY EVALUATION: The function of the vent lines to prevent overpressurization of radwaste containers is no longer required since use of the fill ports has been discontinued. The removal of the vent lines does not result in any operational or functional change to the affected systems (solid radwaste and Radwaste Building ventilation). The design provided by this change has been evaluated against applicable design criteria, installation, and operational requirements. It has been determined that all necessary requirements and commitments are met. Piping modifications have been designed in accordance with ANSI B31.1 code requirements as applicable. Therefore, this design change will not require a change to the Technical Specifications and does not involve an unreviewed safety question.

[80)

Attachment to GNRO-93/00001 l

, SRASN: NPE-92-065 DOC NO: DCP-89/0087-S00-R00 l r

i DESCRIPTION OF CHANGE: The revere load measuring i j system for the 125 ton hook of the containment polar  !

i crane will be replaced witn an Area Brown Boveri (ABB)  !

load measuring system. The outdated Revere system has {

an accuracy of .5%. The state-of-the-art ABB system l will have a design accuracy of .2%. The Revere i system has a small indicator / operators unit located on  !

the pushbutton pendant which is not convenient to use. i The ABB system will have a convenient indicator / I operators unit inside the cab and a large LED [

scoreboard type readout that should be visible from all l locations on the refueling floor. The Revere system l has two setpoints but the contact outputs are not used. .

The ABB system will utilize two alarm setpoints that will activate the crane siren and two setpoints that i will disable the main hoist. The Revere system utilizes one strain gage load cell in the 125 ton upper l block. The ABB system will utilize two pressductor i load cells in a new 125 ton bottom block. The existing i 125 ton hook / bottom block assently will be replaced with the spare (Unit 2) 125 ton hook and the new bottom  !

block. ,

i REASON FOR CHANGE: The existing load measuring system ,

j is simply not acceptable. During reactor assembly /  !

disassembly, heavy loads are lifted by the polar crane, j Loads such as the reactor steam dryer and steam  ;

separator are lifted from a submerged condition.

Determination of binding and/or lack of complete

disengagement of the hold-down devices is very  ;

i difficult.  !

i

, During RF03 the steam separator lifting lugs were bent  !

as well as one of the separator hold down bolts. Per

, plant staff, an accurate load measuring system could ,

have prevented this. l i

SAFETY EVALUATION: There are no Technical j

, Specification requirements for the containment polar i

crane or its load measuring system. The changes of  ;

i this modification will not require that they be added  !

, to the Technical Specification and no other Technical Specification requirements will be affected.  ;

Therefore, the Technical Specification margin of safety  ;

is not affected.  ;

i Although the containment polar crane is designated in j the UFSAR as Seismic Category I, it has no active safety function. The only evaluated accident / j i

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i Attachment to GNRO-93/00001 i

l J

i NPE-92-065

, Page 2 malfunction applicable to the containment polar crane  ;

is a load drop accident. The Unit 2 hook and the new

, bottom block will be certified to the containment polar  ;

crane design specification which corresponds to the UFSAR design criteria. The Unit 2 hook has passed a l 150% ANSI B30.2.0 load test. The Unit 2 hook /new '

bottom block assembly must pass a 125% ANSI B30.2.0 load test. All required field changes will be in 3

accordance with applicable design criteria and Seismic Category II/I design requirements. The seismic qualification of the crane will be maintained. ,

Therefore, neither the probability nor the consequences l of an evaluated accident / malfunction is increased. The  :

new load measuring system is an improvement in terms of I j both reliability and safety. ,

Although NUREG-0612 requires the hook and bottom block i be considered as a heavy load, the dropping of the I bottom block and hook is not addressed in the UFSAR.

The dropping of the hook and bottom block alone is not considered credible (reference USNRC Docket No. 50-416, j

" Evaluation of Heavy Load Handling Operations at Grand

Gulf Nuclear Station"). No new failure modes are ,

l created. This design change creates no interfaces with other safety related systems or other equipment l important to safety. The replacement of the hook and  !'

j bottom block will not increase the probability of dropping the hook and bottom b1cck. Therefore, the  ;

. possibility of an unevaluated accident / malfunction is (

2 not created. j l  !

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Attachment to GNRO-93/00001 I i

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5 SRASN: NPE-92-066 DOC NO: MCP-91/1028, Rev. O  ;

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DESCRIPTION OF CHANGE: The manner in which Check i Valves N1N23F050A and N1N23F050B were constructed and installed in the piping system requires that the i associated piping be cut to allow removal of the -

valves. An Engineering Evaluation Response (EER) ,

requested installation of flanges in the piping to facilitate removal of the valves for inspection.

~,

REASON FOR CHANGE: The objective of this change will be to install a set of flanges at each valve to '

facilitate removal of the check valves for the recommended inspections.  !

SAFETY EVALUATION: The adding of flanges will not  !

adversely affect the structural integrity of the i associated piping. The piping designs meet ANSI B31.1 r code requirements and supported for the appropriate l deadweight and thermal loads. The modification made by l this change will not affect the function, operation, or  ;

performance of this or any other system. The heater, f vents and drains systems serve no safety function and {

as stated in the FSAR, failure of this system will not compromise any safety related system or prevent reactor i

! shutdown. Therefore, this change will not create an i unreviewed safety question. In addition, this system i is not addressed in the Technical Specifications and [

! this design change will not affect the function or operation of this or any other system. Therefore, no

. changes to the Technical Specifications are required.

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Attachment to GNRO-93/00001-1 RASN: NPE-92-067 DOC NO: MCP-92/1072, Rev. O DESCRIPTION OF CHANGE: This change replaces the existing offgas condenser tube side (Turbine Building i 3 cooling water side) relief valve with a new reli

• valve which is not vendor supplied. Therefore, u. 3R Figure 9.2-026 must be revised to reflect that the new 7 tube side relief valve is not vendor supplied.

4 l REASON FOR CHANGE: The new relief valves are required to ensure that the connecting Turbine Building Cooling Water (TBCW) piping and maintenance block valves are protected from over-pressurization as required by ASME ,

, Section VIII B & P V code and ANSI B31.1.

SAFETY EVALUATION: The Offgas System (N64) is discussed in the Technical Specifications. However, the proposed change does not affect the technical specifications since the hydrogen concentra _su i specified (4% by volume) is not affected and this is not major modification of the Offgas System. The TBCW ,

System (P43) is not addressed by the technical specification since the system has no safety function and failure of the system will not compromise any safety related system or component. Since the proposed  ;

modification will not alter the non-safety related operating function of the N64 and P43 systems and since ,

the modification will not result in creation of an interface with or an effect upon safety related

components, structures, or systems the modification i does not constitute a change to the Technical .

j Specification, or an unreviewed safety question.

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Attachment to GNRO-93/00001 SRASN: NPE-92-068 DOC NO: MCP-91/1074, Rev. O DESCRIPTION OF CHANGE: Due to steam erosion of the extraction steam pipe this change replaces the pipe with an erosion resistant alloy steel (Chromium-Molybdenum), also change the size of the 1" and 2" GBD-5 & 6 to 2"-GAD-1 & 2 (from carbon steel to alloy steel).

REASON FOR CHANGE: GBD-5 & 6 piping developed steam leaks due to steam erosion. The 1" line was increased to 2" to minimize high velocity in the pipe.

SAFETY EVALUATION: Replacing carbon' steel piping and associated fitting with Chromium-Molybdenum piping will not affect the Extraction Steam System'(N36) function, operation, or performance in any way. The effect of the new piping on the system will have a significant improvement in the form of system reliability. The piping has been designed to ANSI B31.1 code requirements and supported for the appropriate deadweight and thermal loads only. Therefore, the piping and their supports will function in their intended manner. The system affected by this change has no safety-related functions. Failure of this system will not compromise any safety-related system or component and will not prevent safe reactor shutdown.

This system is not addressed in the Technical Specifications and no new technical specification requirements are being added. Therefore, this change will not require a change to the Technical Specifications and will not create an unreviewed safety question.

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SRASN: NPE-92-069 DOC NO: MCP-91/1094, Rev. O j l

DESCRIPTION OF CHANGE:

This change relocates System N33 (Main and Reactor Feed Pumps Turbine Seal Steam and Drain) pressure transmitter 1N33-PT-N042A(B) and .

pressure switch IN33-PSL-N044A(B) near the process line '

with the instrument tubing sloped to preclude trapping ,

water in the tubing. The presently installed pneumatic input pressure controller will be replaced with a controller having a 4-20 milliampere DC input. The pressure transmitter 1N33-PT-N042A(B) will supply the electrical input for the new controller [1N33-PIC-R010A(B)] in addition to its presently connected loads.

REASON FOR CHANGE: System N33 instrumentation high point vent valves are being left open approximately 1/4 l

, turn in order to dampen seal steam pressure oscillation. When these valves are in the normal operating position (closed), the seal steam pressure oscillates. The affected control loops contain  ;

instruments pneumatic controllers 1N33-PC-R010A(B),

pressure valves 1N33-PV-F510A(B) and pressure transmitters 1N33-PT-N042A(B). The oscillating pressure may also adversely affect pressure switch 1N33-PSL-N044A(B) resulting in spurious alarm by- .

annunciator 1N33-PAL-L610A(B). l l

SAFETY EVALUATION

The change in instrumentation described by this modification will provide improved .

i pressure regulation for the seal steam supplied to the

• main and reactor feed pump turbines. This instrumentation is part of the system which assures that only nonradioactive steam can flow out of the shaft glands into the Turbine Building.  ;

The change will not require any changes to the GGNS Technical Specifications.

The change will not have an adverse affect on accidents or equipment important to safety previously evaluated in the SAR and vill not create the possibility for an accident or malfunction of equipment of a different type. No margin of safety defined as the basis of any Technical Specification will be reduced.

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SRASN: NPE-92-070 DOC NO: DCP-84/0049, Rev. 1 DESCRIPTION OF CHANGE: Replacement of the existing power supply monitor cards in each Average Power Range l Monitor (APRM) page with new power supply monitor II I cards with transient qualification function l incorporated. The added circuitry operates such that {'

any +5V ABNORMAL +/-15V LOW TRIP initiation signal must persist for at least the period of the adjustable qualification delay before a trip is generated.

REASON FOR CHANGE: Reactor protection system trips initiated by "INOP" trip of the APRM system have been

, attributed to the effects of lightning during severe weather. This replacement is necessary to minimize Reactor Protection ~ System (RPS) trips initiated by the APRM system which have been attributed to brief transients during severe electrical storms.

SAFETY EVALUATION: This change is to replace the existing power supply monitor cards with new power supply monitor cards which incorporate an adjustable time delay before a 15V LOW /5V ABNORMAL signal is initiated. Presently, a voltage aberration causes a card trip output resulting in simultaneous de-energization of the " UPSCALE THERMAL", " UPSCALE NEUTRON", and " INOPERATIVE" trip sections for both channels, any one of which is sufficient to cause trip channel actuation (half-scram). Only the power supply voltage aberration initiation signal will incorporate this time delay. Operation of all other trip  :

initiation signals is unchanged. The change to the i

power supply monitor cards does not affect the operation or probability of malfunction of equipment

, described in the UFSAR. The new design utilizes the present power supply monitor interface so that the possibility of malfunction different than previously evaluated is not introduced. The change acts to prevent short term electrical disturbances from causing i undesirable trips.

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Attachment to GNRO-93/00001 SRASN: NPE-92-071 DOC NO: MCP-91/1084, Rev. 0 DESCRIPTION OF CHANGE: This change will eliminate the interface between the Load Shedding and Sequencing (LSS) System and Balance of Plant (BOP) Buses 11HD, 12HE, 13AD, and 14AE which causes these buses to be

disconnected (shed) during a Loss of Coolant Accident (LOCA) coincident with an 80% bus undervoltage (BUV) .

REASON FOR CHANGE: The BOP load shedding feature of the LSS System resulted in a non-valid shedding of the BOP loads on 09/16/90 and 07/28/91. The shedding of the BOP loads resulted in reactor trips (via turbine trips) and subsequent challenges to safety systems.

Removal of this fcature would prevent recurrence of this transient reculting from any event causing this feature's non-valid operation.

SAFETT EVALUATION: The basic function of the Load -

Shedding and Sequencing System (LSS) is to disconnect  ;

(shed) and connect, automatically in sequence, loads on the Class IE buses. The LSS accomplishes this without -

degrading the integrity, voltage, or frequency of the Class 1E supply. The 4.16 kV bus undervoltage BOP load i shedding circuits are also part of the LSS System.

The LSS System depends on two means of detecting l unacceptable voltage levels in the Class 1E electrical ,

system. These two methods are detecting the complete  ;

loss of voltage (primary method) or a sustained,  !

degraded voltage condition (secondary method) on a Class 1E 4.16 kV bus.

i One set of bistables monitor voltage at the 4.16 kV Class II bus of its safety division. These bistables ,

are the primary method for detecting unacceptable voltage levels. A measured nominal voltage of less than 70% for 0.5 seconds indicates unacceptable degradation (complete loss of voltage) of the preferred (offsite) source powering the bus. The LSS then disconnects the Class 1E electrical system from the offsite source and sequences the loads onto the onsite Class 1E source.

A second set of bistables act upon observing a nominal bus voltage between 70% and 90% for 9 seconds when the ,

bus is powered by the offsite source. These bistables provide the secondary method of detecting unacceptable voltage levels in the Class 1E electrical system. This condition is indicative of a severe degradation of the entire grid since automatic relaying on the network

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Attachment to GNRO-93/00001 '

NPE-92-071 Page 2 would normally restore nominal voltage in much less ,

than 9 seconds if the transmission systems were.

• manageable. The LSS then disconnects the Class 1E electrical system from the offsite source and sequences i the loads onto the onsite Class 1E' source.

i These two methods provide the basis for the bounding  !

accident analyses. .

A third set of bistables measure voltage on the Class 1E buses. These bistables are actuated by an observed bus voltage of 80%, or less, of nominal. Upon receipt 1 of both a Loss of Coolant Accident (LOCA) signal and.

the 80% bus voltage signal, the LSS then sheds the BOP .

6.9 kV and 4.16 kV buses from the offsite power source.

The BOP load shedding feature is inhibited if the second level of undervoltage protection provided by the. ,

LSS System has already actuated. The operation of the '

BOP load shedding feature does not affect the operation of the Class 1E onsite electrical power system. The successful operation of the BOP load shedding capability is not a prerequisite for any safety i function.

This BOP load shedding capability was designed to offer '

a possible improvement in the voltage profile at the 9

Class 1E electrical buses under extreme system [

operating conditions. These operating conditions  !

include minimum grid voltage during full-power ,

operation of both Units 1 and 2 with one service ,

transformer out of service followed by a LOCA on one l unit. This voltage improvement was desired to optimize Class 1E bus voltage when offsite power is available i during the Design Basis Accident (DBA). l i

The potential for the extreme system operating condition of minimum grid voltage during full-power ,

operation of both units no longer exists due to the cancellation of Unit 2. As a result, the potential for voltage improvement is outweighed by the potential for 4 transients resulting from the non-valid operation of the feature. The successful operation of this feature ,

is not credited in the system voltage calculation for i Unit 1. t L

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Attachment to GNRO-93/00001 SRASN: NPE-92-072 DOC NO: DCP-88/0060-1, Rev. 0 DESCRIPTION OF CHANGE: This change provides the design for the removal of snubbers from the following piping arrangements:

1. Main steam line"C" to main steam flow instrumentation: From the "C" main steam line flow measurement elbow tap in the drywell, 1"-DCA-32 to condensing chamber Q1B21D019C and 3/4"-DCB-48 from the condensing chamber to drywell penetration 426A.

2. Main steam line "D" to leak detection system:

From the "D" main steam line flow restrictor in the drywell, 1"-DCA-32 to condensing chamber Q1B21D009D and 3/4"-DCB-29 from the condensing chamber to drywell penetration 450F.

3. Reactor Pressure Vessel (RPV) to RPV -

instrumentation: From the reactor vessel, 1"-DCA- 1 26 to condensing chamber Q1B21D004B and 3/4"-DCB-7 from the condensing chamber to drywell penetration 431.  :

4. RPV head to various locations: From the RPV head vent, 2"-DBA-19 to valve Q1B21F002 and valve Q1B21F001, with a branch line 1"-DBA-20 to i 1"-DBA-26 to condensing chamber Q1B21D002 and .!

3/4"-DCB-10 from the condensing chamber to drywell penetration 442F with a branch line 2"-DBA-21 to valve Q1B21F005. Q1B21F002 and Q1B21F001 are I

motor operated isolation valves for the head vent and drain line that vents gases to the drywell equipment sump. Q1B21F005 is a motor operated ,

isolation valve from the head vent and drain line to a main steam line.

The portions of piping systems encompassed by this -i change were reanalyzed using Code Case N411, and i snubbers that could be deleted were identified. In l some cases, minor compensating modifications in  !

remaining pipe supports, or the replacement of a  !

snubber by a rigid strut, are found necessary to l maintain the system below allowable stress limits and these are also included in this change.

• The analyses that document this design demonstrate that all ASME code requirements and applicable regulatory criteria remain satisfied.

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i NPE-92-072 l Page 2 j REASON FOR CHANGE: Removal of snubbers is desirable since there are a large number of them installed at GGNS and they require a significant maintenance and testing effort during refueling outages. Worker dose reduction and potentially improved outage schedules are among the benefits of snubber reduction.

. r SAFETY EVALUATION: The modifications made by this change will not affect the system function, operation, or performance in any way. The capability of the t system to perform its safety functions, including RPV .

head vent valve operation, RPS function, MSIV actuation, containment isolation, and ECCS actuation, i is not affected. The deletion and modification of pipe ,

supports (i.e., snubbers) will not adversely affect the  ;

structural integrity of the associated piping. The i piping and pipe support designs meet ASME Section III '

2 requirements and are qualified as Seismic Category I.  !

. Therefore, the piping and pipe supports will function in their intended manner. No existing system interfaces are affected and no new system interactions  ;

are created.

I The reanalysis of the piping systems has shown that all  :

ASME Section III code allowables have been met and .

therefore the probability of a piping failure has not  !

increased. The modification will not introduce any new l postulated piping failures and the existing hazards i evaluations are not affected.

No new system interfaces with any equipment have been {

created and no existing interfaces have been adversely

, affected. No new failure modes for the system or any l l equipment have been created. }

By remaining within the same allowables specified by i the applicable codes as stipulated for piping, supports i and supporting structures, the margins of safety l provided by these allowables are not affected. i i  :

No revision is required to Technical Specification 3/4.7.4 since individual snubbers are not identified in this Technical Specification and these changes will not .

, create an unreviewed safety question. l

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Attachment to GNRO-93/00001 SRASN: NPE-92-073 DOC NO: MCP-92/1055, Rev. O  ;

1 DESCRIPTION OF CHANGE: The purpose of this change is j to provide repair instructions for pipe support structures based upon reanalysis of Residual Heat j Removal CRHR) Loop "B" piping system for higher design l temperatures. The piping reanalysis of RHR (E12) in _!

the shutdown cooling mode has been performed in i response to a Material Non-Conformance Report (MNCR).  ;

Pipe support modifications to be implemented include the replacement of three rigid struts with three e snubbers for RHR "B" pipe support. Additionally, two other pipe supports, one on RHR "A" and one on RHR "B", e shall have stiffeners added to embedded plates. ,

REASON FOR CHANGE: RHR Loop "B" piping has been  ;

reanalyzed for 360 F for shutdown cooling mode of  !

operation. Pipe support loads resulting from the  !

revised piping analysis have necessitated modifications to one RHR "A" pipe support and two RHR "B" pipe i i supports. i l SAFETY EVALUATION: The modifications to pipe supports l on the RHR "A" & "B" systems will not adversely affect  ;

the system function, operation or performance in any  !

manner. The capability of the system to perform its safety function is not adversely affected. The [

addition of the snubbers to the piping system will i ensure the process piping and pipe supports are in compliance with ASME Section III requirements as well as Seismic Category I criteria. The pipe support i modifications to be implemented by this change will i ensure the piping system will perform its intended  ;

function for all design conditions and modes of  !

operation. Pipe support loads transmitted to building  ;

structures have been evaluated and found acceptable. L No existing system interfaces are affected and no new  !

system interfaces are created. The reanalysis of the  !

?

piping system with the added snubbers has shown that  !

all ASME Section III code allowables have been met and i the probability of a piping failure has not increased. '

The modification of pipe supports to be implemented by this change will not introduce any new postulated i piping failures and existing hazards evaluations are not affected. Margins of safety as defined in the i Technical Specifications are in no way reduced by the j modifications to pipe supports on the RHR System.  !

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Attachment to GWRO-93/00001 SRASN: NPE-92-074 DOC NO: MCP-91-1001-S00-R00 l J

DESCRIPTION OF CHANGE: The following Bailey 731/732 .

recorders will be replaced with Westronics Series 2100 recorders: ,

1D17R600 - Offgas, Radwaste and Containment Vent ,

Radiation l 1D17R601 - Offgas Post Treatment Radiation  !

1D17R602 - Standby Service Water (SSW) Loop A & B .t Return Radiation 1D17R603 - Containment and Drywell Vent Radiation ,

1D17R604 - Offgas Post Treatment Radiation 1D17R605 -

Fuel Handling Area Vent Exhaust Radiation 1D17R606 - Fuel Handling Area Pool Sweep Exhaust Radiation 1D17R607 - Fuel Handling Area and Turbine Building Vent Radiation SD17R608 - Control Room Vent Radiation REASON FOR CHANGE: The Bailey 731/732 recorders are rbsolete. -

SAFETY EVALUATION: The changes specified by this modification do not affect any Seismic Category I structures or components. No Class 1E circuits or ASME Section III Class 1, 2 or 3 piping or components are added or modified by this change. All the recorders i are non-safety related and the panel in which they are located (H13-P600) is non-safety related. The function 7 of Recorders D17R600 and D17R607 is Regulatory Guide 1.97, Category 3 control room indication. The function of the other recorders is non-safety related t control room indication.

Recorders D17R603, 605, 606, and 608 receive their signals from safety related loops but are divisionally #

separated / isolated per Regulatory Guide 1.75. This separation / isolation is not compromised by the  :

implementation of this change.

No evaluated accident is predicated by the failure of i the recorders. This design change will be an l improvement in terms of reliability and monitoring  ;

capability. .

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Attachment to GNRO-93/00001 i

SRASN: NPE-92-075 DOC NO: MCP-90/1032-S00-R00 i I

DESCRIPTION OF CHANGE: This change deletes the pH I alarms for the Standby Service Water (SSW) System and removes or spares all equipment and tubing used for i remote monitoring of pH in the SSW basins. ,

REASON FOR CHANGE: The sample elements for the SSW Basin A, B pH analyzers are located after the check valve in the discharge piping of the SSW pumps. This i location allows the pH analyzers to sample stagnant water when the pumps are not running, resulting in nuisance Hi-pH alarms. The alarms are not being used  ;

by Operations, and have been placed on the inactive list. .

i SAFETY EVALUATION: No. accident evaluated in the FSAR .

is predicated on a failure of any SSW component or the  :

SSW system itself. The SSW system pH monitoring l instrumentation is not required to mitigate the ,

consequences of any accident previously evaluated in j the FSAR. The SSW system will continue to function as  ;

assumed to mitigate the consequences of accidents. The ,

pH monitoring instrumentation with associated alarms removed by this modification does not affect any other equipment. The pH monitoring instruments are not required to mitigate the consequences of a malfunction of equipment important to safety. ,

. The design function of the pH monitoring equipment is l to alert the control room of unacceptably high pH of the SSW basin water. The SSW system is not equipped with an automatic acid addition system and the j instruments do not perform any control function.

l Removal of this equipment will not affect the overall 2 quality of the SSW water, since samples are taken  ;

manually when the system is running and analyzed in the  !

laboratory. The pressure boundary of the SSW system '

will not be affected by closing and capping the  ;

instrument valves. The SSW system will continue to i

meet all system design criteria assumed in the margin of safety for the SSW Technical Specifications. '

Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced. l I

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Attachment to GNRO-93/00001 1

SRASN: NPE-92-076 DOC NO: MCP-90/1109-S00-R00 i

DESCRIPTION OF CHANGE: Previous changes provided for  !

the installation of an Alternate Rod Insertion (ARI)  :

System and the modification of the existing ,

Recirculation Pump Trip (RPT) System to meet the '

requirements of 10CFR50.62 as described in the UFSAR.

An evaluation identified an instance in which a relay race will occur in the initiating logic for Anticipated r Transients Without Scram (ATWS) ARI/RPT. In the event of loss of power for the ATWS ARI/RPT initiating logic, the process signals from the reactor level transmitters decay faster than the reference signals at the alarm

units. When the process signals fall below the trip setpoint following this loss of power, there is still sufficient residual voltage present to pick up the actuating relays and cause an ATWS ARI/RPT actuation.

This change is being made to replace the actuating relays with time delay on energization (TDE) relays.

REASON FOR CHANGE: To allow sufficient time for any residual voltage to fall below the minimum operating voltage of the relays.

SAFETY EVALUATION: The ARI system will provide an ,

alternate scram path to shut down the reactor which is diverse and independent from the RPS in the highly unlikely event of a failure to scram during or following a transient event. The replacement of the ARI/RPT actuating relays wich 1 second TDE relays will-prevent a nuisance ARI/RPT actuation should the above described scenario occur, but will not affect the system's ability to perform its intended function.

i The design of the ARI system and the modified ATWS RPT

, system is such that all-safety related criteria of '

other components and systems will continue to be met.

These systems are electrically independent and separated from the Reactor Protection System (RPS) and all other Class 1E circuits. The power source for these systems is BOP batteries.  :

Safety margins for ARI/RPT are not specifically delineated in GGNS Technical Specifications bases.

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Attachment to GNRO-93/00001 SRASN: NPE-92-077 DOC NO: MCP-90/1059-S00-R00 ,

DESCRIPTION OF CHANGE: This modification replaces three Rosemount Model 1151 transmitters with Rosemount Model 1153 Series B transmitters in the High Pressure  ;

Core Spray System and the Reactor Core Isolation Cooling System.

REASON FOR CHANGE: Model 1151 transmitters are no '

longer available as " safety related" and existing documentation for this model no longer meets future environmental qualification requirements.

SAFETY EVALUATION: The replacement instruments and the instruments to be added by this modification are environmentally / seismically qualified as required.

Each instrument and instrument loop will perform the l same instrument design functions as previously. The replacement Rosemount Model 1153 transmitters were  ;

qualified by testing. The square root extraction equipment to be added has been seismically qualified  ;

and applicable tests reports have been evaluated for '

this modification. The setpoint, control and indications provided by this modification remain unchanged except for the square root scale on 2 trip units. Therefore, the present margin of safety is maintained.

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Attachment to GNRO-93/00001 l l

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SRASN: NPE-92-078 DOC NO: DCP-90/0005-S03-R00 DESCRIPTION OF CHANGE: This modification changed the Automatic Depressurization System (ADS) air receivers from carbon steel to stainless steel.

REASON FOR CHANGE: To prevent corrosion.

SAFETY EVALUATION: The replacement of the air receivers with a stainless steel type will not affect the operation or the function of the ADS since the receiver size and location has not changed. The safety related receiver and piping designs meet ASME Section III requirements and are qualified as Seismic-Category I. Thus, the system will function in its intended manner. No other accident precursors evaluated in the UFSAR are affected by this change.  ;

This design change will not affect the operation of the Nuclear Boiler System (B21) _as analyzed in the FSAR.

The replacement of the air receivers with a stainless steel type will increase the reliability of the ADS by reducing the possible introduction of particles to the i ADS valve actuators.

This change will not increase the probability or the consequences of any accident evaluated in the UFSAR, does not create the possibility of a new accident or malfunction, and does not reduce any margin of safety defined in any Technical Specifications.  !

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Attachment to GNRO-93/00001 SRASN: NPE-92-079 DOC NO: DCP-90/0109-S00-R00 DESCRIPTION OF CHANGE: This modification adds 17 new permanent "O" annubars along with associated "Q" manual isolation valves, tubing, and dP gauges for the safety-related heat exchanges for Standby Service Water (SSW)

Loops A&C. Also required are 2 new temperature wells ,

(one installed temperature well will be moved to a new location) and 42 new temperature instruments (19 l

bimetallic thermometers and 23 capillary tube ,

thermometers). ,

REASON FOR CHANGE: To conduct thermal performance testing of safety-related SSW heat exchangers for  !

Loops A & C.

SAFETY EVALUATION: The safety function of the SSW system is to provide a reliable source of cooling for plant auxiliaries that are essential for 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 limiting failure coincident with a loss of offsite power.

The "Q" annubars and associated piping have been i designed to ASME Section III, Class 3 requirements and are qualified as Seismic Category I. None of the non-  :

"Q" temperature instrumentation to be installed will i penetrate the SSW process piping.

{

Installation of the instrumentation required by this modification will not adversely affect the UFSAR i analyses for piping failures in a fluid system. For the 14 annubars to be installed in the Containment and Auxiliary Building, the existing analyses have demonstrated that the hazardous effects from moderate  ;

energy pipe cracks (i.e., area flooding, low velocity r wetting of equipment in the area, and II/I concerns) will encompass the effects of any additional postulated leakage. For the 3 annubars in the Diesel Generator Building, the UFSAR states that no pipe cracks have been postulated due to flooding concerns in these rooms. Therefore, the annubar installations have been designed per the UFSAR to stress levels in which no  !

cracks are required to be postulated. The installation of capillary tube thermometers and associated tubing on the surface of the "A" Fuel Pool Cooling and Cleanup (FPCC) System heat exchanger piping, the "A" drywell purge compressor piping, and the "A" Residual Heat Removal (RHR) pump seal cooler outlet piping will not '

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Attachment to GNRO-93/00001 NPE-92-079 Page 2 t

adversely affect the piping analyses discussed in the UFSAR since these instruments do not penetrate the process boundaries and since the additional weight of the thermometers and associated 1/4" tubing is t

comparatively insignificant.

The installation of capillary tube thermometers in the ,

inlet and outlet air streams of the Engineered Safety

• Feature (ESP) Electrical Switchgear Room Coolers, the ,

Emergency Core Cooling Systems (ECCS) Room Coolers, and i the Reactor Core Isolation Cooling (RCIC) System Room i Cooler will not adversely affect the operation or function of any of these room coolers as described in i the UFSAR. These thermometers do not penetrate any process boundaries, are passive in nature, result in insignificant blockage of the air flows, and have i supports designed and fabricated to meet Seismic II/I concerns.  ;

The installation of bimetallic thermometers in the i Standby Service Water System temperature wells will not affect the operation or function of the SSW system as  !

described in the UFSAR since these thermometers are passive in nature, do not penetrate the process  ;

boundaries, and add insignificant weight to the SSW ,

piping.  ;

The installation of bimetallic thermometers in the Fuel Pool Cooling and Cleanup Room Coolers ventilation ducting will not affect the operation or function of these room coolers as described in the UFSAR since these thermometers are passive in nature, are supported to meet Seismic II/I concerns, result in insignificant 1 blockage of air flow, and due to the fact that the size l of the process (duct) penetrations (1" diameter) does i not present a significant bypass leakage path in the l event of these holes being uncovered due to failure of i these thermometer installations.

The operation or function of the affected systems as analyzed in the UFSAR is not affected by the addition of the instrumentation.

The instrument installations added by this modification 4

will not change the function or operation as defined by the bases of the Technical Specifications. Therefore, the margin of safety is not reduced.

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Attachment to GNRO-93/00001 I

l SRASN: NPE-92-080 DOC NO: DCP-87/0087-S00-R00 l l

DESCRIPTION OF CHANGE: This change installs two i handswitches in the Control Room to control incoming  !

feeder breakers 52-15405 and 52-16405 to Motor Control l Centers (MCCs) 15B42 and 16B42, respectively.

l REASON FOR CHANGE: To provide a remote control method i of closing these breakers from.the Control Room. ,

Presently, these breakers can only be manually closed ,

from the Auxiliary Building after a load shedding and  !

sequencing (LSS) actuation in the event of a loss of coolant accident (LOCA).  :

I SAFETY EVALUATION: The addition of handswitches in

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Control Room panel for the control of incoming breakers to MCCs 15B42 and 16B42 will provide a means of energizing MCCs 15B42 and 16B42 from the Control Room.

At present, these MCCs are energized by manually j closing breakers 52-15405 and 52-16405 in the Auxiliary  :

Building. Thus, this design modification is consistent .

with the existing design such that manual action is l required to reenergize these MCCs during LOCA. This change does not impact previously analyzed accident '

scenarios with respect to initiating events.

Sequence interlocks prevent inadvertent reconnection [

during LOCA sequencing. MCCs 15B42 and 16B42 loads  ;

will not exceed Emergency Diesel Generator (EDG) load  :

j capabilities; therefore, accident consequences as  !

previously evaluated will not be impacted. >

This design modification will meet all safety related criteria of the other components in the Class 1E power ,

system. The new handswitches in the Control Room will i perform a parallel function to the. existing switches in j the Auxiliary Building. The equipment to be used for l I this modification is Class 1E and meets applicable  !

Class 1E criteria with respect to separation,  ;

qualification, etc., and therefore, the probability of  ;

a malfunction of equipment important to safety  ;

previously evaluated in the FSAR is not increased.  !

This design modification is consistent with the existing design and the addition of the LSS interlock i prevents inadvertent operator error to energize MCCs i

15B42 and 16B42 prior to reconnection of all essential loads, on emergency diesel generators on LOCA. Thus, reliability of other systems and components will be unaffected. All equipment used for this modification is qualified to be used in safety related application

[100]

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Attachment to GNRO-93/00001 NPE-92-080 l Page 2 I 1

and will be installed seismically. Thus, this modification will not increase the consequences of a  ;

malfunction of equipment previously evaluated in the j FSAR. .

This design modification does not change the original l design intent. All modifications to the control

• circuit of these breakers are safety related and i Seismic Category I. All wiring and cabling per this  !

design modification meet Regulatory Guide-1.75  !

requirements. Therefore, this design will not reduce- '

the margin of safety as defined in the basis for any Technical Specification. ,

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Attachment to GNRO-93/00001 SRASN: NPE-92-081 DOC NO: DCP-87/0039-S00-R00 $

l DESCRIPTION OF CHANGE: The Control Room kitchen area was modified by replacing all cabinets, countertops, appliances, ceiling tile, flooring, dining booths, sink i' and water cooler. Additional lighting was added in the kitchen as well as new receptacles for increased -

electrical requirements. The kitchen was repainted and a sheet metal fascia and cabinet soffits added, as well as a communications console. Finishes in the adjacent areas, corridors, locker room, office and toilet shall  ;

be redone. The locker room was expanded. The fire extinguishers were relocated in highly visible locations. A dry chemical extinguishing system was ,

added for the new stove. l REASON FOR CHANGE: The Control Room kitchen was cited t for revamping as part of the proposed modifications resulting from the Detailed Control Room Design Review  ;

(DCRDR). The current kitchen facility is substandarc  !

and was designed for a staff of approximately 1/3 of $

the actual staff which presently utilizes the facility.  !

Architectural features such as room finishes, countertops, flooring and ceiling tile are permanently l soiled and require replacement. The seating capacity  ;

in the kitchen is insufficient as is the storage space in the adjacent locker room.

SAFETY EVALUATION: No accident previously evaluated in j the UFSAR has a probability of occurrence which would ['

be increased by changes implemented by this modification. Added conduits, boxes, transformer, and  ;

lighting fixtures which could pose a II/I hazard have l been designed seismically. All seismic designs are in l accordance with the applicable codes and used  !

applicable loads as defined by the UFSAR. The piping i installed by this modification meets all applicable l design codes. The piping system affected will function ]

in its intended manner. The removal of walls, to expand the locker room, were evaluated and found to i have no negative impact on adjacent walls which are designed seismically.

The materials specified in the modification minimized the use of combustibles as applicable to comply with the requirements of the UFSAR. To account for those combustibles that will be present in the kitchen area, the combustible heat load calculation was revised to document heat loads and durations for incorporation into the Fire Hazards Analyses (FHA). Subsequently, an FRA revision request was generated to document and

[102)

Attachment to GNRO-93/00001 NPE-92-081 Page 2 evaluate the acceptability of the changes to the fire

., zones made per this modification. For good fire .

protection practice, a dry chemical extinguishing i system shall be added for the new stove.

No accidents previously evaluated in the UFSAR had '

consequences related to the Control Room kitchen, locker room and adjacent toilet, or corridors. No systems, directly or indirectly affected by this modification, are used in mitigating the consequences of an accident as analyzed in the UFSAR.

t The changes to be implemented by this modification are i

such that they are not introducing any new hazards or reducing the margin of safety as defined in the basir for any Technical Specifications.

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[103]

Attachment to GNRO-93/00001 SRASN: NPE-92-082 DOC NO: DCP-87/0048-S01-R00 DESCRIPTION OF CHANGE: Stainless steel manifolds will replace existing carbon steel manifold on the diesel '

generator starting air manifold downstream of the ,

starting air solenoid valves.

REASON FOR CHANGE: To eliminate the corrosion problem i in the starting air manifold. i SAFETY EVALUATION: The operation and function of the affected system will not be altered. The piping supplied by this modification meets.all applicable design requirements and will function in its intended manner. This change in no way impacts the diesel 3 generator capability for mitigating the consequences of an accident. This modification does not change the limiting condition for operation or applicability of surveillance requirements as defined in the basis for Technical Specifications.

This change will not increase the probability or the consequences of any accident evaluated in the UFSAR, does not create the possibility of a new accident or malfunction, and does not reduce any margin of safety defined in any Technical Specifications. >

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Attachment to GNRO-93/00001 i SRASN: NPE-92-083 DOC NO: DCP-87/4023-S00-R00- -

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DESCRIPTION OF CHANGE: This design change provides i' continuous drainage of the High Pressure Core Spray ,

(HPCS) Diesel Generator engine exhaust lines when the .

engines are in standby mode, and allows drainage system' .

isolation via a manually operated valve when the engines are in operation. Additionally, this modification provides system operability from a r location accessible to operators.

REASON FOR CHANGE: To prevent the introduction of exhaust fumes into the oily waste hub and sump. To alleviate the requirement of using ladders and/or  ;

scaffolding with the current design.

SAFETY EVALUATION: The HPCS Diesel Engines (P81)

System operation and function will not change. The-piping and supports supplied by this modification meet  ;

the applicable code requirements and will function in  ;

their intended manner. Removal of the exhaust silencer '

drain subsystem will not affect the operability of the P81 system or the Diesel Generator Combustion Air  !

Intake and Exhaust System (DGCAIES).

Components of the P81 and the HPCS DGCAIES are ,

protected from rain, snow, ice, sleet, dust storms, missiles, and the effects of pipe whip or jet ,

i impingement from high and moderate-energy pipe breaks. '

This design change does not adversely affect these design features, therefore, these systems will function in their intended manner. .i An analysis has been performed demonstrating that the piping meets all applicable design requirements, and '

that the structural integrity of the piping system is assured such that a single failure of any component of the P81 system will not result in the loss or function '

of more than one generator or prohibit safe shutdown of )

the plant. These modifications maintain this design ,

feature.

These modifications are a maintenance enhancement intended to prevent the inleakage of water into the engine turbochargers and cylinders and subsequent damage to the diesel engines. ,

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Attachment to GNRO-93/00001 l NPE-92-083 i Page 2 l The addition / removal of the piping, valves, and supports by this modification does not change the  ;

limiting conditions for operational applicability, or l the surveillance requirements for the P81 system as defined in Section 3/4.8 of the GGHS Unit One Technical Specifications. Implementation of the actions described in this modification are an enhancement to the existing P81 exhaust line drain system which will only be used when the engines are in standby mode.

Therefore, there is no reduction in any margin of safety.

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Attachment to GNRO-93/00001 SRASN: NPE-92-084 DOC NO: DCP-86/4506-S00-R00 DESCRIPTION OF CHANGE: This modification provides requirements for replacement of existing drywell bulkhead vent covers.

REASON FOR CHANGE: To eliminate upper containment pool leakage into the drywell during refueling operations.

SAFETY EVALUATION: The modification is designed to perform the same function as the original vent covers and will have no affect on other equipment. M The vent covers are designed to remain leaktight when installed. This assists in maintaining the pool levels required to mitigate the consequences of irradiated fuel rupture. When the vent covers are not installed, they are stored in a manner that would not directly or indirectly ngravate other accident consequences.

The design ensures that the new vent covers do not directly or indirectly interact with equipment important to safety in a manner different than the original vent covers. Therefore, the probability of equipment malfunction is not increased.

The vent covers are designed to resist the applicable hydrostatic and seismic loads per the UFSAR. The new vent covers help maintain the Technical Specification basis water level that is needed to ensure that 99% of iodine gap activity released by rupture of an irradiated fuel assembly is removed by the water.

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Attachment to GNRO-93/00001 SRASN: NPE-92-085 DOC NO: DCP-86/0067-S00-R00 DESCRIPTION OF CHANGE: This modification provides the design changes necessary to install axially oriented anti-rotational pins on the turbine (System N31) shaft coupling bolts.

REASON FOR CHANGE: Enhances the previous design characteristics of the turbine shaft coupling bolt pins and enhances the integrity of the turbine shafts.

SAFETY EVALUATION: The design change is a system ~

enhancement and does not affect system performance, there will be no increase in the consequences of any accidents previously evaluated in the FSAR.

The GGNS FSAR does not classify the turbine in relation to safety or non-safety. The FSAR does however point out that the turbine is not required for safe shutdown of the reactor.

The design change modifications have been proven safe in operational testing. Since the design changes do not incorporate modifications to safety related equipment and since the changes have been designed and proven reliable in field applications, the changes do not create the possibility of an accident of a different type than already evaluated in the FSAR.

The GGNS Technical Specification does not address the turbine and/or generator shaft, with the exception of addressing the turbine overspeed protection system.

Installation of pins which enhance the capability of meeting the turbine shaft coupling bolt tightening requirements will not affect the margin of safety as defined in any Technical Specification.

[108]

Attachment to GNRO-93/00001 SRASN: NPE-92-086 DOC NO: DCP-84/0189-S00-R01 DESCRIPTION OF CHANGE: This modification installs a metal enclosure building covering the existing j equipment on the Radial Well (system P47) platforms for )

each well. Each building will have heating,  !

ventilation, lighting, and lightning protection. j REASON FOR CHANGE: The heating will maintain above I freezing conditions with no internal heat loads while the ventilation system will not allow excessive heat build-up during well operation. The lighting will be more uniform than at present and the grounded metal building with four lightning rods will provide greater j protection against lightning strikes.

SAFETY EVALUATION

The buildings are non-seismic, non-Category I designed to the same Uniform Building Code (UBC) requirements as the Ranney Wells. i Loss of the P47 system has already been postulated and  !

is not required for safe reactor shutdown. The Standby i Service Water (SSW) System has been provided to ensure <

safe reactor shutdown under all conditions.

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, The probability of a malfunction of equipment important  :

to safety is not increased because the buildings do not house any equipment important to safety and are ,

physically isolated from any equipment important to i safety.  ;

The loss of a well or loss of the total well system has ,

already been postulated. Since these are the worst  !

possible conditions that could occur, any accident ,

condition due to these buildings or their components is  ;

already bounded by a postulated well loss.  !

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The margin of safety will not be reduced by the addition of the Enclosure Building. Moreover, the l Radial Wells are not used as the basis for any ,

Technical Specification. [

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SRASN: NPE-92-087 DOC NO: DCP-84/0181-S00-R00  :

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DESCRIPTION OF CHANGE: This modification provides alternate floor and equipment drains in the Water Treatment Building. Also, this change provides a permanent connection to the mobile makeup water trailer  ;

and the polishing filter trailer. [

l REASON FOR CHANGE: To provide alternate floor and li equipment drains in the Water Treatment Building for the damaged drains utilized by the makeup water and l domestic water systems. Also, the permanent connection .

to the mobile makeup water trailer and the polishing i filter trailer provides makeup water. l SAFETY EVALUATION: The floor and equipment drain  !

system and the makeup water system are not included in  ;

the identification of causes or sequences of events and  !

i systems for accidents previously evaluated in the UFSAR. i I The changes implemented by this modification are not  !

associated with any system or component used in mitigating the consequences of an accident as analyzed in the UFSAR. ,

[

The UFSAR. states that failure of the makeup water .

system will not compromise any safety-related system or l component and will not prevent safe reactor shutdown.  !

The changes associated with the floor and equipment  :

drain system are confined to the Water Treatment

.i Building. 1 i

. The changes implemented will not alter the design (

j function of the floor and equipment drain system or the j makeup water system. The water quality of the makeup l

, water will continue to meet the requirements of the ,

i UFSAR.  !

i The floor and equipment drain system and the makeup i t

water system are not addressed in the Unit 1 Technical Specification. The margin of safety is not reduced.

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Attachment to GNRO-93/00001 l l

1 SRASN: NPE-92-088 DOC NO: DCP-83/0515-S00-R01 DESCRIPTION OF CHANGE: This modification provides for the installation of a re-flash unit in the annunciation circuits of the Process Radiation Monitoring System.

The subject annunciators are associated with the low sample flow in the Standby Service Water (SSW) A&B, the Component Cooling Water (CCW) and the Liquid Radwaste Effluent Systems.

REASON FOR CHANGE: The re-flash unit provides the capability to receive multiple indications of alarms as they occur.

SAFETY EVALUATION: This modification provides for additional annunciation capability of the Process Radiation Monitoring System. The FSAR identifies the systems affected by this modification as non-safety related but required for plant operation. The installation of the re-flash unit does not affect any equipment that provides for initiation of, or contributes to previously analyzed accidents.

This modification provides enhancement to the annunciation of the associated systems. Neither the annunciation of the monitoring system nor the systems being monitorec' are part of the basis of evaluation of any accident as previously evaluated in the FSAR.

There are no electrical or mechanical connections to systems important to safety that are part of an accident evaluation contained in the FSAR.

The installation of this DCP provides for the enhancement of the monitoring of the systems affected.

The operation, setpoints, or operating characteristics of the affected systems is not altered. The annunciation of the Process Radiation Monitoring System is not a part of any basis used to determine the margin of safety as described in the Technical Specification.

Therefore, this DCP will not reduce the margin of safety as defined in the basis for any Technical Specification.

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Attachment to GMRO-93/00001

SRASN: NPE-92-089 DOC NO: DCP-83/0139-S00-R00 l

DESCRIPTION OF CHANGE: This modification installs protective covers over the Low Pressure (LP) turbine 4

inlet piping expansion joints located inside the  :

condensers. Similar expansion joints designed by i Utility Power Corporation (UPC) have failed due to

erosion from falling condensate inside the condenser.

UPE has designed and supplied the expansion joints at i GGNS. This design change is based on design l information provided by UPC. ,

REASON FOR CHANGE: To enhance the reliability of the turbine inlet steam piping. l SAFETY EVALUATION: Installation of erosion protection i covers on the LP turbine inlet steam piping expansion joints will enhance the reliability of the main and reheat steam system CN11). Implementation of the ,

design change will not affect the functional .

characteristics of the system.  !

j The steam piping is a component of the power conversion l system which is classified as safety class "other" in  :

the FSAR. No other equipment important to safety as evaluated in the FSAR is adversely affected.

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turbine overspeed protection system, Main Steam Isolation Valve CMSIV) isolation on low condenser  :

vacuum, or the turbine stop and control valve closure i scrams as addressed in GGNS Technical Specifications.

1 This modification does not reduce the margins of safety as defined in the bases of any GGNS Technical Specifications. ,

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Attachment to GNRO-93/00001 l

SRASN: NPE-92-090 DOC NO: DCP-83/0417-S00-R00 DESCRIPTION OF CHANGE: Install platforms to gain  ;

access to the top of the Reactor Feed Pump Turbine '

(RFPT) Lube Oil Reservoirs.

REASON FOR CHANGE: To provide access to the reservoirs and to enhance personnel-safety. j SAFETY EVALUATION: The platforms are located in the Turbine Building which is classified as a non-Category I structure. The platforms are designed in accordance with the Uniform Building Code as there is -

no potential for safety related equipment to be affected. The platforms do not interface with any structure system or component required to protect against an accident as described in Chapter 15 of the FSAR.

The platforms are designed to withstand its dead load,  :

100 psf live load and seismic loads without failure and ,

will not affect safe operation of the plant.

There are no Technical Specifications applicable to  :

platforms in the Turbine Building. Furthermore, the addition of the platform will not prohibit or alter the surveillance requirements as specified by GGNS ,

Technical Specifications.

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Attachment to GNRO-93/00001 SRASN: NPE-92-091 DOC NO: MCP-90/1019-S00-R00 DESCRIPTION OF CHANGE: This modification revises affected documents to reflect the new room names from Unit 2: OC404, OC405, OC406, OC411, OC502, OC705, i OC708, and OC711 to " Unit 1 Support Area." This modification installs the flammable liquids storage cabinet and portable fire extinguisher in what was .

Room OC411.  !

REASON FOR CHANGE: The room names were changed because the original Unit 2 occupancies have been replaced with activities for support of Unit 1. Room OC411 includes ,

storage of combustible and flammable liquids which requires provision of a flammable liquids storage i cabinet and one additional portable fire extinguisher.

SAFETY EVALUATION: The fire extinguisher and flammable liquids storage cabinet have been evaluated for Seismic II/I hazards and the wall to which the extinguisher is mounted has been examined for structural integrity. The portable fire extinguisher ,

, does not present a missile hazard to safety related systems or components as per the UFSAR. j The inherent function of the equipment provided by this change is to reduce the consequences of postu?.ated fires. This change does not adversely affect the analysis of safe shutdown in the event of a fire as stated in the UFSAR. Installation of the flammable liquids storage cabinet, portable fire extinguisher, and reassignment of room names will not introduce new  ;

or different failure modes.

The flammable liquids storage cabinet and the fire extinguisher are listed and arranged in accordance with National Fire Protection Association (NFPA) 30.

Reassignment of room names, installation of portable fire extinguishers and flammable liquids storage cabinets will not affect the bases for any GGNS Unit 1 l Technical Specification. Consequently, the margin of  ;

safety as defined in the basis for any Technical Specification is not reduced.  !

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Attachment to GMRO-93/00001 i

SRASN: NPE-92-092 DOC NO: DCP-82/5074-S00-R01 DESCRIPTION OF CHANGE: Changes to the Control Rod Drive (CRD) Maintenance Facility as described below i were implemented. System G50 (CRD Flush Tank Filter i and Leak Test System) has been desig.,ed and developed '

to provide the maintenance processing required for CRDs. Other systems affected by implementation of this design change with a description of the required modifications are as follows:

a. System G41 - Fuel Pool Cooling and Cleanup System s Due to the need for additional floor spaca in the CRD Maintenance Facility, a 1" drain line cff the i Fuel Pool Cooling and Cleanup System piping requires rerouting. t

b. System P21 - Makeup Water Treatment System The supply of demineralized water required for the CRD Maintenance Facility is obtained from the Makeup Water Treatment System. j

c. System P52 - Service Air System  !

Service air required for equipment and tools vital  ;

for the operation of the CRD Maintenance Facility is obtained from System PS2.

d. System P48 - Floor and Equipment Drain System i Liquid radwaste generated by operation of the CRD

• Maintenance Facility will be disposed through System P48.  ;

e. System P64 - Fire Protection System

• A section of fire water piping will require rerouting to avoid interference with the monorail  !

being installed by this modification. A I combustible heat load calculation was prepared to incorporate changes in combustible loading.

f. System T42 - Fuel Handling Area Ventilation System

The existing exhaust duct in the CRD Maintenance ,

, Facility will be used as the exhaust path from the l room.  ;

g. System P53 - Instrument Air System  :

Air required for CRD Maintenance Facility instrumentation will be obtained from System P53 and a reroute of an instrument air line is required to avoid interference with the monorail 3 being installed by this modification.  !

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Attachment to GNRO-93/00001.-

NPE-92-092 Page 2

h. System T41 - Auxiliary Building Ventilation System A new air handling unit is being added to support the CRD Maintenance Facility. This unit will provide cooling and filtration of the return air from the facility,

i. System P71 - Plant Chilled Water System The Plant Cnilled Water System provides the cooling medium for the air handling unit being provided for the CRD Maintenance Facility.

Also, heating, ventilation and air-conditioning, lighting and smoke detectors were added to complete the CRD Maintenance Facility.

REASON FOR CHANGE: To incorporate approved changes recommended by Plant Staff into the design of the CRD Maintenance Facility.

SAFETY EVALUATION: System G50 serves no safety related function and its failure will not compromise any safety related systems or prevent the safe operation of the plant.

The piping modifications associated with System P71 are non-safety related and have been evaluated for and impose no Seismic II/I concerns. The piping modifications associated with Systems PS2, P53, and P21 have been designed as Seismic Category II/I, non-safety related. The affected portions of the Plant Chilled Water (P71), Instrument Air (P53), Service Air (PS2),

and Makeup Water Treatment (P21) Systems are non-safety related as defined in the UFSAR and failure of these systems will not compromise any safety-related system or component and will not prevent safe reactor shutdown.

l Portions of the Fuel Pool Cooling and Cleanup System

! (G41) are designed as Seismic Category I as discussed in the UFSAR; however, the piping reroute required by this modification is outside the ASME boundary and is therefore non-safety related. The piping reroute has however been designed as Seismic II/I, thus the safe operation and design of the system as described in the UFSAR is not impaired.

i The piping modifications associated with the Floor and Equipment Drain System (P48) have been designed as Seismic Category II/I, non-safety related and comply with the system design bases as described in the UFSAR.

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Attachment to GNRO-93/00001 i

i NPE-92-092 t

Page 3 The routing of a section of Fire Protection System ,

(P64) piping has been designed as Seismic l Category II/I, consistent with the UFEAR so that inadvertent operation or failure of the system will not  ;

impair safety related systems. Changes in combustible loading have been reviewed and detenmined not to  !

present unacceptable exposure to safety related systems  ;

or components. Furthermore, changes in combustible loading will not adversely affect the ability of existing fire protection systems and components to l

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perform their specified functions.

The air handling unit and its associated ductwork have been designed to incorporate features for ALARA

' considerations as discussed in the UFSAR. The ductwork .

supports associated with Systems T41 and T42 has been i designed Seismic Category II/I, non-safety related, thereby maintaining safe operation of these systems and ,

complying with the system design as described in the  !

UFSAR.  ;

Electrical circuits required for implementation of this [

modification are non-safety related and their failure -

would in no way affect the safe operation of the plant.

Conduit supports (with the exception for the lighting design) as required per this modification have been designed either Seismic Category I, safety related, or Seismic Category II/I, non-safety related to preclude any hazardous conditions. The new supports for light fixtures and associated conduit in the CRD Maintenance Facility and the clean CRD storage area are non-safety ,

related and impose no Seismic II/I concerns. '

Instrumentation, tubing, and tubing supports required for implementation of this modification are non-safety j related, non-seismic. They have been evaluated for and j impose no Seismic II/I concerns.

! Proper closures for penetrations through fire barriers

. separating safety related fire areas have been  :

specified in the modification, therefore compliance with technical specifications is maintained. The  ;

design temperature of 75'F for the CRD Maintenance i Facility is well within the limit of 104*F for general  !

areas of the Auxiliary Building as specified per  ;

technical specifications. Additionally, the  ;

modification has specified the operational  !

considerations to ensure plant compliance with the ALARA Program and with Technical Specifications; thus ,

the margin of safety as defined in the basis is  ;

2 maintained.  !

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SRASN: NPE-92-093 DOC NO: DCP-83/0170-S00-R00 DESCRIPTION OF CHANGE: This modification removes the large trumpet speakers from the public. address (P.A.) i system in Containment at Elevation 208 and replaces them with eight smaller speakers.

REASON FOR CHANGE: To reduce reverberation and make better communications.

SAFETY EVALUATION: Conduit and speakers additions have been reviewed for II/I considerations and have been _ i supported on "Q" Seismic Category I or "non-Q" seismic supports. The new speakers will meet the P.A. design criteria. Separation criteria has been followed.

The dedicated P.A. speaker system in Containment is not used to mitigate the consequences of an accident. The speakers being changed affect only the paging function.  :

The channel talk function will not be affected.

Pressure boundaries are not breached in this change. i Additional conduit, cable, and speakers to the dedicated P.A. system cannot increase the consequences t of a malfunction of equipment since it has no i

connection or associated function with equipment that is important to safety.

The dedicated P.A. system is a BOP system and as such does not function by itself as an accident prevention .

system. No new accident scenario will be created. I The dedicated P.A. system is not addressed in the bases of any technical specification. Changing speaker types on the P.A. system will be consistent with vendor design, therefore, the margin of safety will remain the same. l 4

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Attachment to GNRO-93/00001 ,

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-1 SRASN: NPE-92-094 DOC NO: DCP-84/0107-S00-R00 I i

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DESCRIPTION OF CHANGE: This modification seals 25 open j floor penetrations on Elevations 113' and 133' of the  ;

Turbine Building.

l REASON FOR CHANGE: Penetrations TC-12B and TC-40B will i be sealed to prevent injury to plant occupants. The _

two penetrations are large enough for someone to step  :

into or trip over. The remaining twenty-three i penetrations will be sealed to prevent liquids from -

flowing through the open penetrations.  ;

. SAFETT EVALUATION: The 25 Turbine Building floor penetrations do not perform safety related functions.

Closure of the subject penetrations in the Turbine  ;

Building will not affect the design of the Turbine  !

Building as described in the FSAR. Sealing the i penetrations prevents the spreading of liquid spills between the two floors and diverts the flow to the floor drainage system. The penetration closures specified will adequately perform their design function  !

without affecting any equipment required for safe t operation of the plant. '

i This change will not increase the probability or the i consequences of any accident evaluated in the UFSAR, ,

does not create the possibility of a new accident or i j malfunction, and does not reduce any margin of safety defined in any Technical Specifications.

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Attachment to GNRO-93/00001 SRASN: NPE-92-095 DOC NO: DCP-88/0050-S00-R00 DESCRIPTION OF CHANGE: This modification includes direction for the inclusion of the Unit II power block into the Unit I protected area (PA) boundary. Included I in the scope of this package is the addition of a l double fence north and east of the Unit II Turbine l Building that will tie into the existing perimeter at j the northeast corner of the Standby Service Water (SSW)  ;

basin and behind the Unit 1 Turbine Building east of  ;

the railway door.

The work included in this package is installation of l new ductbank, 3 manholes, 4 handholes, foundations for i 9 new camera towers, eight new security zones, and 7 modification to the existing E-fields and microwave for the SSW basin. Also, included is cable and conduit input to run raceway up to the Unit II Turbine Building penetration into the Control Building. This package also contains instructions for site grading in the existing Unit 1 PA required to ensure Probable Maximum Precipitation (PMP) effects by the modification will  !

not impact safe operation of Unit 1. l REASON FOR CHANGE: To incorporate Unit II facilities j into Unit I because Unit II has been cancelled. ,

SAFETY EVALUATION: The changes are limited to '

installation of security equipment (fences, alarms,

~

cameras), modification of the E-field and microwave for  !

the SSW basin (on the western perimeter of the PA) and 4 PMP grade changes. This package does not include information required for system operability (i.e.,

cable termination) and is therefore limited in l potential impact on plant safety. Grade changes are I being performed to ensure that the new perimeter does not impact plant operability from a PMP event and the security system is non-seismic and non-safety related.

These changes, when fully implemented, will provide a new security perimeter that provides for penetration ,

detection and alarm assessment consistent with that required for Unit I and will strengthen the GGNS

, Security Program with respect to potential for  !

~

radiological sabotage. The Security System (C83) is non-safety related. The Security System is not used as  ;

a basis for margin of safety.  !

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[120] l L

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Attachment to GNRO-93/00001 l NPE-92-095 Page 2 Because these changes do not adversely impact systems i required for plant safety and PMP requirements are met,  ;

no new accident scenario is created. PMP concerns have been addressed such that there will be no decrease in the margin of safety by including this area within the Unit 1 PA. This design change does not introduce any new tornado missiles that are not enveloped by the  !

existing design basis missiles evaluated in the UFSAR.

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[121]

• Attachment to GNRO-93/00001  !

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SRASN: NPE-92-096 DOC NO: DCP-84/0249-S00-R00 j

DESCRIPTION OF CHANGE: This modification installs spray nozzles in the ultrasonic level sensor housings of the waste holding tanks. Also, the floor and i equipment drain filter pressure and delta pressure  ;

instruments are replaced with a filled capillary system which is isolated from the process fluid by a diaphragm seal. ,

REASON FOR CHANGE: To allow flushing of the sensor housings to prevent clogging of these sensors. Also, the filled capillary system separated by a diaphragm seal will prevent the slurry from clogging these sample 3

lines. l t

SAFETY EVALUATION: These modifications will not contribute to the chances of process line leakage or  ;

radwaste tank rupture, as discussed in the UFSAR. The ,

modifications meet or exceed the design, material and construction standards used in the existing G17/G18 (Liquid Radwaste/ Solid Radwaste) systems. The G17/G18

, systems serve no safety function and its failure will {

not compromise any safety-related system.

This design change does not alter any existing equipment important to safety nor will it create any system interfaces with equipment important to safety. l In the unlikely event that the modifications to the  !

instrumentation, piping, waste holding tanks caused damage to the subject components or any associated ,

equipment, this occurrence would be bounded by the  ;

analysis in the UFSAR for the Reactor Water Cleanup I (RWCU) Phase Separator / Decay Tanks. The radiological  !

3 results of this accident are shown in the UFSAR to be 9 acceptable.

No flowpaths which bypass the previously analyzed l process train for the G17/G18 systems are established *

. by these modifications. Therefore, no new effluent release scenarios are postulated.

i These modifications will enhance the performance and reliability of the liquid and solid radwaste systems. 1 The modifications do not change the limiting conditions i for operation, applicability, actions, or surveillance l requirements as defined in the bases for Technical Specifications. Therefore, the margin of safety for

~

offsite release limits is unaffected. No change to the overall radwaste process is introduced. l I

[122]

. . _ - . _ - _ _ =. _. _ ._.__ _ . ___

Attachment to GNRO-93/00001 t

SRASN: NPE-92-097 DOC NO: MCP-90/1011-S00-R00 {

i i

i DESCRIPTION OF CHANGE: This modification provides for the installation of three personnel barriers in the  ;

Radwaste Building at the following locations: .

Elevation 151', Filter Area; Elevation 136';  !

Elevation 118', Penetration RP-60C. Two barriers ,

consist of a structural steel frame and/or door with expanded metal. The other barrier consists of miscellaneous 1/4" thick steel plates.

REASON FOR CHANGE: To prevent unauthorized personnel entry into very high radiation areas.

SAFETY EVALUATION: The barriers are installed to  ;

prevent unauthorized entry into very high radiation areas and serve no safety related function. Structural ,

integrity of the barriers has been justified by -i calculation. The barriers preclude creation of II/I hazards on the basis of their installed locations. The  !

barriers will not create any major ventilation flow ,

obstructions. Available fire protection measures will 1 not be affected by installation of the barriers, since they are constructed of mesh and allow smoke and water to pass through. The barriers will not affect the .

. seismic design of the Radwaste Building since small -

flexible members utilized for their construction will ,

not allow a transfer of loading between floors other than previously analyzed. l The barriers are constructed of non-combustible i materials to prevent an increase in the potential for ,

fire related damage to any nearby safety related  ;

equipment or components. [

The barriers are used for personnel safety only, and  !

are not considered in a basis of any Technical Specification; therefore, the original margin of safety  !

will not be reduced. j I

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Attachment to GNRO-93/00001 6 SRASN: NPE-92-098 DOC NO: DCP-91/0026-S01-R00  !

DESCRIPTION OF CHANGE: This modification changes each  !

of the drywell area equipment drain sump pump control circuits such that the pumps will not start in the recirculation mode until the associated recirculation valve is open ("NOT FULLY CLOSED").

REASON FOR CHANGE: This will provide the time required for the discharge valve to come off the open seat ("NOT ,

FULLY OPEN") and thus prevent a false high leakage alarm.

SAFETY EVALUATION: This change modifies the pumps '

control circuits such that the pumps will not start until their associative recirculation valve is open.

The pumps and valves will still function as per the i original design intent. The delay in the pump start signal until the valve is open ensures proper valve alignment. Failure of any component, system or ,

structure added or modified by this change will not initiate any evaluated transient or accident. '

The modification of the pump recirculation mode circuitry is not required to support the safe shutdown of the reactor or to perform in the operation of ,

^

reactor safety features. The changes made by this modification do not prevent any equipment relied upon to mitigate the consequences of any evaluated transient i or accident from performing its safety function.

The modification of the control circuitry does not affect a safety related system or design function. All existing design functions for equipment important to safety are maintained by this design change. The

failure of any component, system or structure added or modified by this change will not cause any malfunction ,

of equipment important to safety.  !

Modification of the pump control circuits such that the pumps will only start in the recirculation mode when their respective recirculation valve is open does not change the original design intent of any equipment. '

The Technical Specification requirements for the drywell equipment drain sump level and flow monitoring as described in the Technical Specification is not affected by this change. Therefore, the margins of safety as defined in the bases for the Technical Specifications are not changed by this modification.

[124]

Attachment to GNRO-93/00001 SRASN: NPE-92-099 DOC NO: EER 92/6184 DESCRIPTION OF CHANGE: This engineering evaluation requested that temporary lead shielding be attached to certain portions of the Reactor Water Cleanup (RWCU)

System in order to reduce radiation exposure to ,

personnel performing work in this area. The lead ,

shielding will be installed during Operating Modes 4 ,

and 5 only, and must be removed prior to restart.

REASON FOR CHANGE: The temporary lead shielding will be attached to certain portions of the RWCU system in r order to reduce radiation exposure to personnel performing work in this area.  ;

2 SAFETY EVALUATION: An evaluation has been performed on <

the associated stress problem. This engineering i evaluation shows that with the added weight of lead shielding, the structural integrity of the applicable i RWCU piping is maintained even in the unlikely event of an Operating Basis Earthquake f0BE), Safety Shutdown Earthquake (SSE), Safety Relief Valve discharge induced ,

loads from one valve's subsequent actuation (SRVONE) and Seismic Anchor Movements (SAM). All applicable ASME code stress allowables are met. Therefore, the .

system operability in Operating Modes 4 and 5 is not affected by the addition of the temporary lead i shielding. All the lead shielding must be installed during Operating Modes 5 and 5 only, and must be  :

removed prior to the plant restart after RF05. The lead blankets may be wrapped around the piping at the ,

locations and in the amounts shown in the EER response.

Also, no other lead shielding or any other additional '

weight can be attached to the piping while this shielding is attached.

Temporary addition of lead shielding does not result in I any permanent changes to location, routing, or type of l l supports, nor does it alter any component performance characteristics, design parameters, or operational ,

parameters of the affected system after the temporary )

i lead shielding is removed. <

This change will not increase the probability or the consequences of any accident evaluated in the UFSAR, does not create the possibility of a new accident or nelfunction, and does not reduce any margin of safety defined in any Technical Specifications.

[125]

J Attachment to GNRO-93/00001 {

l SRASN: NLS-92-001 DOC NO: Switchyard Fence Modification DESCRIPTION OF CHANGE: The modification will add 860 feet of fence and two thirty-foot gates inside the .

current switchyard perimeter fence around the existing MP&L warehouse. In addition, a third thirty-foot gate will be added to the fence that will access the MP&L warehouse only. A gravel road and associated culvert will be added to provide access to the warehouse ,

through one of the gates without entering the rest of-the switchyard.

i REASON FOR CHANGE: The perimeter fence to the switchyard will be modified to improve security of and i decrease traffic flow through vital areas of the switchyard. The modifications are in response to l NUREG-1410, the Vogtle event, and as a result of OA-91-10, Evaluation of GGNS Responses Related to the .

Vogtle Loss of Vital AC Power Event.

SAFETY EVALUATION: The addition of the fence called for by this modification will have no negative impact '

on the safe operation of the plant. The new fence and road will provide access to the MP&L warehouse while bypassing the switchyard, thus routine traffic to and l from the shop will avoid vital plant equipment. This '

change will effectively reduce the risk of any malfunctions of switchyard equipment.

i The switchyard perimeter fence is not used as the i technical justification.in the bases for any Technical Specifications. The margin of safety as defined in the basis for any Technical Specification is not reduced.  !

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Attachment to GNRO-93/00001  ;

SRASN: NLS-92-002 DOC NO: CR-PLS-91-002 DESCRIPTION OF CHANGE: UFSAR Table 9.5-11 provides a  ;

point-by-point comparison between the fire protection j program at GGNS and the positions of the NRC's  ;

Appendix A to Branch Technical Position APCSB 9.5-1 for i design, procurement, installation, and testing and the ,

administrative controls for the fire protection systems  ;

. of safety-related areas at nuclear power plants. UFSAR

] Table 9.5-11, Item C.2.2 (b) presents the GGNS position  ;

with respect to Quality Assurance Requirements for i control of procurement of items and services for  !

Nuclear Power Plants, as committed in Regulatory Guide 1.123 Rev. 1. The present GGNS commitment  :

requires documented visual verification to ensure that j non-identical replacement material / equipment is adequate (equal to or better than original). This .

table is being revised to replace the " visual verification" method with a more appropriate design review method. J REASON FOR CHANGE: This change is being made to update the UFSAR discussion of the Fire Protection Program to  :

reflect current and more appropriate procedures and  ;

practices which ensure the adequacy of non-identical i replacement material / equipment. The current commitment j for visual verification is being modified to reflect t initiation of a design change which would ensure that l material / equipment is reviewed for adequacy, as required by Plant Administrative Procedure 01-S-09-1 i (Rev. 29).

t SAFETY EVALUATION: UFSAR Table 9.5-11 states that visual verification is to be used to determine suitability of non-identical replacement material / l' equipment. The more appropriate methodology to ensure i that adequate replacement material / equipment is used in i the Fire Protection System is accomplished by the >

i design review process. The design review process is a l more conservative approach than performing a " visual verification" to determine adequacy of non-identical >

j replacements and therefore the change does not constitute an unreviewed safety question. i License Condition 2C(41) , states in part that "The Licensee may make changes to the approved Fire ,

Protection Program without prior approval of the l Commission only if those changes would not adversely l affect the ability to achieve and maintain safe  !

shutdown in the event of a fire." This change would not adversely affect the ability to achieve and ,

maintain safe shutdown in the event of a fire. i

[127)

Attachment _to GNRO-93/00001 ;

r SRASN: NSP-92-001 DOC NO: Cycle 6 Fuel Design

& Criticality Analysis DESCRIPTION OF CHANGE: Siemens Nuclear Power, Inc.

(SNP) has designed and is fabricating reload fuel for GGNS-1 Cycle 6 (reload fuel batch ANF-1.5 [1] ) . The t fuel is scheduled to begin arriving on site in late February, 1992. RF05 is scheduled to begin on  ;

April 17, 1992. ,

The ANF-1.5 fuel will be delivered to the plant site by truck transport, and will be stored in the new fuel vault and/or the spent fuel pool until the start of the refueling outage. This evaluation addresses the shipping, handling and storage of fresh fuel, which includes the following proposed actions:

a. Environmental effects of fuel transportation,

b. Movement of the new fuel to either the new fuel vault (NFV) or the spent fuel pool (SFP),

c. Storage of fresh reload fuel in the NFV, and

d. Storage of fresh and irradiated reload fuel in the SFP. j REASON FOR CHANGE: To address changes in Cycle 6 fuel design and criticality analysis.  ?

SAFETY EVALUATION: The safety evaluation addresses the  ;

similarity between the Cycle 6 and Cycle 5 fuel design i and the impact of the Cycle 6 design on the criticality analysis. Confirmatory analyses were performed to show that the fuel designs are compatible with and similar  ;

to previous reloads. The Cycle 6 fuel is bounded by  :

the Cycle 5 criticality analysis. Therefore, the activities associated with the movement and storage of }

the Cycle 6 reload fuel do not involve a Technical Specification change or an unreviewed safety question.

[128]

Attachment to GMRO-93/00001  !

SRASN: NSP-92-002 DOC NO: Fuel Movement / j Shuffle During RF05 T i

1 DESCRIPTION OF CHANGE: During RF05, with the reactor in Modes 4, 5, 4 in conjunction with *, or 5 in '

conjunction with *, fuel bundles will be moved in the Auxiliary Building and in the Reactor Building and fuel  ;

assemblies in the reactor core will be shuffled.  ;

REASON FOR CHANGE: During the fifth refueling outage -l (RF05) at Grand Gulf Nuclear Station Unit 1 (GGNS-1),  ;

Entergy Operations will replace depleted Siemens l 1

Nuclear Power (SNP) fuel assemblies with 272 new, unirradiated SNP-1.5 9x9-5 fuel assemblies. The new fuel assemblies are neutronically, thermal-  !

hydraulically, and mechanically similar to the Advanced Nuclear Fuel (ANF) fuel assemblies introduced into the GGNS-1 core during previous reloads. The major  ;

difference from previous reloads is the distribution of l" the uranium in the fresh fuel assemblies. The fresh fuel batch is composed of two sub-batches with a  ;

different average enrichment for each sub-batch. Both i sub-batches use an axially distributed uranium design  :

and have slightly lower bundle average enrichments than used in Cycle 5. All SNP fuel assemblies were designed }

and built specifically for the GGNS-1 reload core.

i Approval for changes to the GGNS-1 Technical  ;

Specifications required to support operations in '

Modes 1, 2 and 3 with the new fuel has been requested }

from the NRC. Approval of these changes is expected +

prior to the scheduled completion of RF05. Once RF05 j

begins, and until these changes to the Technical i Specifications are approved, plant operations will be  !

restricted to Modes 4, 5, 4 in conjunction with *, or 5 ,

in conjunction with *, which do not require any change t to the Technical Specifications. Mode 4 is defined as COLD SHUTDOWN, Mode 5 is defined as REFUELING and Mode l

• is defined as "where irradiated fuel is being handled  !

in the primary or secondary containment, and during CORE ALTERATIONS and operations with a potential for  !

draining the reactor vessel". j SAFETY EVALUATION: An evaluation of the continued ,

applicability of fuel-related analyses described in the GGNS-1 UFSAR to the Cycle 6 core showed that these ,

analyses, together with SNP's Cycle 6 reload analyses,  !

continue to remain applicable to the Cycle 6 core. An i evaluation of the radiological consequences of dropping  !

an SNP 8x8 or 9x9-5 fuel assembly showed it was bounded by the radiological consequences of dropping a GE 8x8  ;

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[129] l

Attachment to GNRO-93/00001 l l

l NSP-92-002  :

Page 2 i

fuel assembly; the offsite doses for the GE and SNP t fuel assemblies are within 25% of 10CFR100 limits, >

which is the acceptance criterion stated in the GGNS-1 ,

Safety Evaluation Report. Consequently, the  ;

radiological consequences of dropping an irradiated ,

fuel assembly onto irradiated fuel assemblies for RF05 .

, are within the acceptance criterion. Analyses have l been performed to show that adequate shutdown margin during interim fuel shuf fling can be naintained subject to the identified restrictions in the fuel shuffle procedure. The proposed action involves no changes to '

, the Technical Specifications, no unreviewed safety '

questions, and no unreviewed environmental questions.  !

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Attachment to GNRO-93/00001 I

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SPASN: NSP-92-003 DOC NO: MNCR-0166-92 j DESCRIPTION OF CHANGE: Continued use of fuel batch XNC  ;

as planned for Cycle 6 given the recently noted near-  !

closure of the fuel rod to upper tie plate gap.  :

This evaluation is based upon and contingent upon the issuance of the NRC SER for PCOL-91/23 for Cycle 6.

REASON FOR CHANGE: The third Siemens Nuclear Power (SNP) 8x8 reload fuel (batch XNC) is experiencing the near-closure of the fuel rod to upper tie plate gap.  ;

The gap closure is due to unexpected, uneven .

differential growth between the fuel rods and the tie  ;

rods. Gap closure could result in increased rod bow, ,

which could impact mechanical, thermal-hydraulic, and t neutronic parameters. This fuel has previously been >

irradiated two cycles. All XNC fuel planned to be used  !

in Cycle 6 will be discharged at the end of Cycle 6 (EOC-6).

SAFETY EVALUATION: Analyses were performed to assess {~

the impact of the closure on this fuel batch for Cycle 6 operation. The analyses addressed the mechanical, thermal-hydraulic / transient, and LOCA >

considerations. The analyses demonstrated the acceptability of operation with this fuel, as designed, for Cycle 6. In addition, an analysis was performed for the 9x9-5 fuel design. The Cycle 6 core consists of 560 9x9-5 assemblies and 240 8x8 assemblies. All of  !

the 8x8 fuel will be discharged at the end of Cycle 6.

t Because of certain design features, the 9x9-5 fuel [

throughout its designed life will be readily able'to ,

accommodate the increased growth experienced in the 8x8

• design. All design criteria will continue to be met. i The upper tie plate will continue to be removable and  ;

no increase in rod bow due to rod to upper tie plate l gap closure will occur. Hence, this safety evaluation -

will address to the operability of the third batch of SNP 8x8 fuel (which will be referred to as the '

"affected fuel').

MECHANICAL All the mechanical design criteria were reviewed.

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[131] ,

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Attachment to GNRO-93/00001 NSP-92-003 l Page 2

)

The increased growth has resulted in the inability to remove the upper tie plate on at least some of the XNC bundles. This has no effect on the safety or operational performance of the fuel in Cycle 6.

Increased loading on the tie plates due to the projected closure and interference is negligible.

The increased growth may potentially result in additional rod bow. The impact of this additional .

bow was evaluated for its effect on Linear Heat Generation Rate (LHGR) limit. The affected fuel will remain within the mechanical design limits.

Minimum Critical Power Ratio (MCPR):

The amount of additional bow was determined and a ,

Critical Power Ratio (CPR) reduction factor was developed for the affected fuel. The affected ,

fuel will operate at low power and even with the CPR reduction factor will not be limiting. The MCPR operating and safety limits will not need to be changed.  ;

PEAK CLAD TEMPERATURE:

The additional rod bow has no impact on the LOCA i analysis. The overall system and hot channel hydraulic performance are not affected. The effect of bow on the-local peaking is within the range considered in the fuel design and does not result in a reduction in margin. The affected fuel will operate at low power, and therefore low initial MAPLHGR (Maximum Average Planar Linear Heat Generation Rate) value, and will not be limiting.

Follow-up inspection was performed on a limited number of bundles from other batches. One XNB bundle had been irradiated for only two cycles (the same as the XNC bundles) and the other for three cycles. Both bundles had the near closure of the gap. Also, near-gap closure has been experienced at Monticello without any adverse effect on safety or operational performance of the fuel.

[132]

Attachment to GNRO-93/00001 NSP-92-003 Page 3 l DESCRIPTION OF TESTING AND EVALUATION PROGRAM OF EEFERENCE 1 Inspection of several XNC fuel bundles was performed to determine the margin to solid compression on the springs between the upper tie plate and the fuel rods. '

Based on these measurements, plus knowing the minimum possible as-fabricated margin to solid compression and the assembly burnup, the maximum rate of gap closure was determined. From this, the maximum amount of ,

interference at the highest anticipated assembly exposure at EOC-6 was calculated.

To determine the expected behavior of the XNC bundles at this interference, a test was performed on a full scale, unirradiated 8x8 fuel assembly. In the test, an axial load was applied to an individual rod to simulate the predicted interference. Rod-to-rod gap measurements were made then at different axial spans (between grid spacers) including at the critical span where boiling transition is predicted to occur. This measured rod bow was then used to evaluate the MCPR, LHGR, and LOCA limits.

The testing and evaluation methodology incorporated several conservative assumptions:

A conservative rate of gap closure was used. This was based on assuming that full gap closure has already developed at GGNS (even though measurements have shown that some margin still exists) and that the maximum as-fabricated gap existed at BOL (Beginning of Life).

The minimum as-fabricated BOL rod-to-rod spacing was used based on the limiting tolerances of both the spacer grids and the rod bow between the spacer. _

The rate of gap closure / interference was conservatively assumed to continue throughout l Cycle 6 (i.e., the closure would be linear with ,

continued operation). The effects of fuel rod compression and increased tie rod tension was not i credited.

Two high growth fuel rods were assumed to be adjacent and bowed toward each other.

[133] 1 1

Attachment to GNRO-93/00001 6

SRASN: NSP-92-004 DOC NO: Fuel'Related Cycle 6 Operational Issues DESCRIPTION OF CHANGE: Cycle 6 will operate with the second reload batch ot Siemens Nuclear Power (SNP) 9x9-5 fuel. This safety evaluation is written to demonstrate the overall appropriateness of the documentation for issues not addressed in the Cycle 6 reload PCOL (Proposed Change of License) or other 50.59 safety evaluations.

This evaluation is based upon and contingent upon the ,

issuance of the NRC SER for PCOL-91/23 for Cycle 6.

REASON FOR CHANGE: Replacing depleted SNP 8x8 fuel with new, unirradiated SNP 9x9-5 fuel assemblies and planned operation with the Cycle 6 core configuration have necessitated revising the plant licensing documentation addressing fuel-dependent issues.

Reviewed and Approved Issues:

1. Safety and operating limits for power operation, which are discussed in the Technical Specifications up to, and including, the recently .

approved changes that were requested per the Cycle 6 Reload PCOL. Approval for revisions to the Technical Specifications affected by the Cycle 6 reload was requested from the NRC,

2. Refueling operations associated with the receipt of new fuel and the movement and storage of new and irradiated fuel in the Auxiliary Building and containment, and continued use of fuel batch XNC as planned for Cycle 6 given the recently noted ,

near-closure of the fuel rod to the upper tie '

plate, which have been addressed in previously approved 50.59 safety evaluations.  :

SAFETY EVALUATION: Changes to the Technical Specifications for Cycle 6 operation are pending approval by the NRC. These changes address operation at power (i.e., Modes 1, 2, and 3). Activities involving new fuel receipt and operations relating to fuel during Modes 4, 5, and

• were addressed in approved safety evaluations. Continued use of fuel batch XNC as planned for Cycle 6 is addressed in an approved safety evaluation.

[134]

r Attachment to GNRO-93/00001 j i

NSP-92-004 Page 2

The primary scope of the other fuel-dependent issues l tied to Cycle 6 operations comprises operation at .

power, including startup, normal operation, and hot  !

shutdown modes. The evaluations / analyses that address  ;

these issues demonstrate the acceptability of operation  !

with Cycle G fuel.

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Attachment to GNRO-93/00001 {

SRASN: NSP-92-005 DOC NO: Transfer of Environmental Surv.

Program ,

DESCRIPTION OF CHANGE: Transfer of Environmental .

Surveillance Program (ESP) from Vice President, Operations Support, to Vice President, Operations GGNS. i More specifically, the ESP transfers from Corporate i

'(Manager, Environmental Services) to Plant Operations (Chemistry Superintendent). .

REASON FOR CHANGE: A Quality Action Team (QAT) studied ,

the organization of Environmental Surveillance Programs at Entergy Operations' three nuclear sites and ,

recommended this transfer of responsibility.

SAFETY EVALUATION: This change will transfer the Environmental Surveillance Pragram and staff from l Corporate to GGNS. This change will not modify  !

requirements associated with GGNS's Environmental Surveillance Program. Since environmental surveillance ,

will continue as before the transfer, there will be no ,

increase in consequence or probability of an accident i

or malfunction. Furthermore, environmental l surveillance activities occur outside and independent of plant operation. Therefore, no adverse affect is '

., possible. i I

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[136]

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t Attachment to GNRO-93/00001 l i

i SRASN: PLS-92-001 DOC NO.: TSTI-1C34-92-002-O-N i I

t i l DESCRIPTION OF CHANGE: Alarm Units 1C34K626A and B i monitor the output of summer 1C34K651. This signal is i the sum of the level channel selected for control by l the Feedwater Control System and the steam program  ;

function generator output, 1C34K652. This alarm unit l causes the reactor recirculation pumps to transfer to  !

low speed when vessel level drops below level 3  !

(-11.4-inches) plus steam program level. The change  ;

involves inhibiting this trip for a short period of  ;

time (approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) with the reactor  !

recirculation pumps on high speed. '

REASON FOR CHANGE: This change eliminates the risk of l a reactor recirculation pump transfer to low speed  !

1 while performing a surveillance test with a potentially .

faulty level select switch and with reactor ,

recirculation pumps on high speed.  !

SAFETY EVALUATION: The reactor recirculation pump  ;

transfer to Low Frequency Motor Generator (LFMG) at j level 3 appears on FSAR Figure 7.7-6, but is.not-  ;

otherwise deecribed in the FSAR. It is not used in any 1 accident ana~.ysis in Chapters 6 and 15. Where a pump l trip is required to insert negative reactivity, j l

transfer at Reactor Pressure Vessel (RPV) level 2 via l

! End of Cycle-Reactor Pump Trip (EOC-RPT) logic is  ;

assumed. The purpose of this trip is for pump  ;

protection, i.e., to maintain sufficient head of water for net positive suction head (NPSH) in a scram to 2 i prevent extended cavitation. The procedure has a

, provision for manual transfer to LFMG in case of a .

I scram. No unreviewed safety question is involved. i a

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[137]

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Attachment to GNRO-93/00001 SRASN: PLS-92-002 DOC NO.: UFSAR 5.2.2.10 I I

DESCRIPTION OF CHANGE: The requirement to remove main .

steam safety / relief valves (:MSRVs) from the plant at i every refueling outage is being changed to a requirement to remove them at appropriate shutdowns  ;

during each refueling cycle.  :

REASON FOR CHANGE: This will allow the valves to be removed for testing during outages other than refueling ,

outages.  ;

SAFETY EVALUATION: This change does not affect the design of the MSRVs or their function in the plant.

The only change is the frequency at which the MSRVs are ,

set pressure tested and the duration of their operation  !

in the plant between set pressure tests.

Since this UFSAR change requires us to continue testing  ;

at least 50 percent of the MSRVs during each refueling cycle, we will continue to meet or exceed the testing requirements of ASME Code Section XI. Therefore, this  !

change does not increase the probability of occurrence of an accident previously evaluated in the UFSAR. ,

Analysis of the results of set pressure testing previously conducted on the MSRVs shows that the MSRVs have consistently operated when they were subjected to  ;

overpressure conditions. None of the valves have  ;

failed to open during set pressure testing.  ;

i There is no evidence to suspect that the MSRVs would fail to protect the plant during an accident or that ,

they would allow larger overpressure transients than i those analyzed in the UFSAR; therefore, the consequences of a malfunction of equipment important to safety will not be increased.

The only credible accidents in which these valves could realistically participate are overpressure accidents and loss of coolant accidents (both small leaks and large leaks), including missile hazards associated with i l

these accidents, all of which have been previously evaluated in the FSAR. Therefore, this change would not create the possibility of an accident of a different type than any evaluated in the FSAR.

Also, there is no reduction in the margin of safety as defined in the basis for technical specification.

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Attachment to GNRO-93/00001  ;

l SRASN: PLS-92-003 DOC NO.: CR-PLS-92-001 i

DESCRIPTION OF CHANGE: This change combines the i Electrical Superintendent and Instrumentation & Control i Superintendent positions into one position. i REASON FOR CHANGE: This change has the net effect of f' flattening the organizational structure of the department and; therefore, vill result in better communications, more consistent implementation of requirements and enhanced performance of the affected  :

sections.

SAFETY EVALUATION: This change only affects the -

organizational structure of maintenance. The change will not relax any requirements associated with the performance of maintenance at Grand Gulf. This change ,

will not result in any reduction in the-duties, i responsibilities or authority of the combined positions  ;

and; therefore, will not have an affect on current  :

practices. This change will not result in any l reduction of the qualification requirements (ANSI 18.1 i and Regulatory Guide 1.8) for the positions and will therefore, not result in any reduction of the i I

capabilities of the department. This-combined position will assume the PSRC duties currently held by the I&C  !

Superintendent which will ensure continued high t performance of this body. The combined position will l not be overloaded'due to the current similarities i between the two positions.  :

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l Attachment to GNRO-93/00001  !

l SRASN: PLS-92-004 DOC NO.: TEMP ALT 92-0011 DESCRIPTION OF CHANGE: This safety evaluation ,

addresses implementation of Temporary Alteration i 92-0011.

This temporary alteration will isolate and replace the l P21 (Makeup Water Treatment System) supply to ['

circulating water pump lube water with a construction water supply. The construction water supply will be i routed from the water treatment building through the  ;

turbine building corridor Elevation 133 to the west l wall of-the circulating water pump house via a 2-u 1/2-inch fire hose where it will be tied into the existing P21 supply piping.

The construction water supply will provide the-same

., automatic parallel lube water supply function as  ;

originally provided by the P21 supply. The existing j domestic water supply to lube water will not be i' changed. A check valve in the construction water supply line will prevent loss of domestic flow to [

construction water. [

A fire hose connection with isolation valve will be f'

installed in the construction water supply line to allow a fire water hose station to be connected and i valved in as a contingency (short term) lube water  ;

supply in the event both the domestic water and  ;

construction water supplies will be lost. She capacity of the domestic water storage tank will provide ,

adequate time for this operator contingency action in ,

the event construction water is out of service.

REASON FOR CHANGE: The P21 supply to the circulating ,

water pump lube water is not reliable due to degraded l pressure and flow. The degraded pressure and flow is ,

believed to be caused by buildup of corrosion in the  !

2-inch piping supply line from the water treatment  !

building to the circulating water pump house.

SAFETY EVALUATION: No unreviewed safety or  !

environmental questions were identified as a result of f this safety evaluation. j Implementation of this temporary alterations will restore a reliable backup lube water supply to the ,

suction of the circulating water pump lube water pumps. t l This will only reduce the chance of a circulating water j pump trip on low lube water flow and subsequent plant  ;

scram on loss of condenser vacuum. The Circulating  ;

Water System serves no safety function.  ;

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Attachment to GNRO-93/00001 i

PLS-92-004 Page 2 Loss of the circulating water pumps during plant operation has been evaluated in UFSAR Section 5.2.5. i The Construction Water System serves no function ,

related to safety. Routing of temporary fire hoses is limited to the Water Treatment Building and the Turbine .

Building corridor elevation 133. Leakage from these  !

hoses would not affect any safety related equipment or  !

prevent safe shutdown of this plant. i The disconnected P21 supply will be isolated such that no new leakage paths or failure mechanisms will be  !

created. ,

Use of the fire water as a contingency (short term) -

lube water supply will not adversely impact operation of the Fire Protection System or prevent it from performing its design function. The total required fire water flow to be supplied as lube water is 50 gpm.

This 50 gpm (short term) loss to the 4500 gpm capacity Fire Protection System is insignificant and therefore j will not adversely affect the system operation. ,

However the 50 gpm flow requirement is beyond the .

30 gpm capacity of the fire water jockey pump which normally maintains fire water system pressure and will -

thus cause the electric motor driven fire pump to run i intermittently to maintain fire water system pressure.

Use of the motor driven fire pumps to maintain system pressure on a continuous basis is not a normal r configuration but is well within the bounds of the  !

system design and thus will not adversely affect the i system operation. No other affects to the Fire i Protection System are postulated.

2 The radiological consequences remain the same as the  ;

consequences of a loss of condenser vacuum as discussed j in UFSAR Section 15.2-.5.5.

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Attachment to GNRO-93/00001 I SRASN: PLS-92-005 DOC NO.: W.O. #65005 DESCRIPTION OF CHANGE: A 1/2-ton jib crane and two (2) !

one-man work platforms (outriggers) will be installed  !

on the auxiliary platform during RF05.

REASON FOR CHANGE: The crane and outriggers will i greatly enhance the constructibility of Shroud Head Stud Assembly Modification (SHSAM) installations and ,

reduce impact on other critical path refueling i activities by reducing dependence on the polar crane for this work.

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. SAFETY EVALUATION: Use of the jib crane and outriggers within specified limitations represents temporary usage  :

of tools to improve outage progress of RF05 SHSAM work  !

and does not increase the probability or consequences of any previously analyzed UFSAR accident or equipment malfunction. Restrictions imposed on the use of these tools and structural / failure evaluations performed for j the tools and interfacing structures / components ,

minimize the probability and consequences of an SSE-caused overturning event postulated during the interim -

when auxiliary platform seismic qualification is not  ;

applicable. Based on these restrictions and  !

evaluations, adequate assurance is established that the possibility of a different type accident or equipment malfunction is not created. In addition, since  :

I 4

structural integrity of the tools and postulated failure effects of the structures / components were  ;

justified by evaluation, RF05 usage of these temporary tools does not affect any technical specification and  ;

does not reduce the margins of safety for any technical specification.

i The use of the subject jib crane and outriggers has never been proposed at GGNS. Therefore, no UFSAR accidents have evaluated probabilities based directly  ;

on the use of these tools. Considering indirect affects on previously evaluated accidents in the UFSAR, these l tools cannot increase the probability of the UFSAR fuel ,

drop accident because interference with the refueling ,

platform is precluded and the tools are restricted to the separator pool area. Also, load drops from the l temporary jib crane are no more probable than those  :

evaluated in the UFSAR polar crane load drop study.

A:.1 rigging must comply with the same GGNS rigging and  ;

load handling procedures / standards. Probability  ;

assessment of an accident caused by outrigger failure l is not necessary because structural integrity of the '

outriggers was justified, by evaluation, for the  !

specific application.

[142]

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Attachment to GNRO-93/00001 6

PLS-92-005 Page 2 ,

Technical specifications for the auxiliary platform and  !

the refueling platform are not affected. In addition, the margins of safety for these technical a

specifications are not affected since the subject tools t used on the auxiliary platform will not enable fuel handling except as intended by the refueling platform main hoist, will not affect load capacities of any auxiliary hoist, and will not compromise protection features that prevent the inadvertent overstress of ,

core internals or pressure vessel. The limitations imposed on the use of the jib crane will restrict its  ;

use to the separator pool only and auxiliary platform handling of control rods will be rendered temporarily impossible. Interference with fuel handling using the refueling platform will continue to be prevented by the ,

collision interlock, and heavy load handling over fuel racks in the upper containment pool is physically prevented by the auxiliary platform's inability to i travel beyond the west wall of the reactor cavity. No '

other technical specifications apply. l t

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l Attachment to GNRO-93/00001 l i

SRASN: PLS-92-006 DOC NO.: 02-S-01-4, R24;  ;

02-S-01-5, R23 l

DESCRIPTION OF CHANGE: Current requiremt.nts of 02-S-01-4, Shift Relief and Turnover, and 02-S-01-5, Shift Logs and Records, require that the oncoming operator sign an area turnover checksheet in order to document acceptance of the watch and to provide a place to itemize remarks. Signature by the encoming oparator also signifies status of equipment in his/her area of responsibility and responsibility for the watch, r This change will eliminate these redundant requirements  ;

to allow use of the operations computerized rounds taking system. Hand held computers used to_ enter data do not have a turnover sheet attached. A comment '

section for the entry of area equipment status may be used to document important turnover information.

Section 6.6 of 02-S-01-5 provides requirements for building operator logbook entries. The guidance for building logbook entries and documentation of area i responsibilities is redundant to the area turnover  ;

checksheet attached to building rounds. Use of this  ;

building logbook will also provide a centrally located book for each area and will require the level of detailed entry for complete turnover status. j i

REASON FOR CHANGE: Procedures 02-S-01-4, Shift Relief l 4

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and Turnover, and 02-S-01-5, Shift Logs and Records, do  !

not allow for implementation of computerized rounds  !

without an attached area turnover checksheet. Hand '

held computers are used to take rounds and the building  ;

operator logbook will provide an alternate for entry of  !

important turnover information per Section 6.6 of Shift Logs and Records.

SAFETY EVALUATION: No manipulations are performed to  !

gather data and rounds are passive data gathering  ;

optrations intended to help determine and prevent  !

degradation of important plant equipment. Readings taken on equipment important to safety employ ,

techniques which does not cycle equipment. Passive data gathering does not include methods of abnormal aligrment or operations which are not prescribed by -

permanent plant procedures.

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[144)

Attachment to GNRO-93/00001 l

PLS-92-006 Page 2 l

t Rounds format or method of readings are not described {

in the station technical specifications, only what data  ;

should be taken for required readings. By the  !-

additional information and abilities provided in i additica to those required by surveillance procedures, trends may be established which will prevent any ,

reduction in the margin of safety as defined by technical specifications for required safety related equipment.

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Attachment to GNRO-93/00001 SRASN: PLS-92-007 DOC NO.: TEMP ALT 92-0012 l

DESCRIPTION OF CHANGE: This temporary alteration '

defeats the circulating water (CW) pump lube water flow low auto trip signal and pump start permissive  ;

interlock. The automatic trip signal will be replaced by procedurally controlled operator action to manually trip the CW pump (s) within 4 minutes (5 minutes total -

including valve closure time) in the event loss of CW pump lube water is verified and cannot be restored. ,

Per vendor response, loss of lube water to the pump and loss of cooling flow to the motor for a period of 5 minutes will not damage or adversely affect pump or motor operation. Defeating the auto trip function will prevent an erroneous CW pump trip and subsequent r reactor scram caused by a spurious pressure spike in the lube water supplies. The CW pump lube water low .

flow start permissive interlock will be replaced with procedural controls to ensure lube water flow is acceptable prior to CW pump start.

REASON FOR CHANGE: Pressure spikes in the CW pump lube water supplies cause momentary spurious low flow '

readings to the float type flow switches. The flow oscillations are caused by pressure spikes which are due to the lube water pressure regulator response to 1 external supply pressure changes. These small flow oscillations will on occasion cause a low flow condition to be sensed by the float type flow switches. ['

These momentary flow oscillations are of no consequence to the CW pumps operation and do not adversely degrade <

the lube water supply but will subsequently cause an t erroneous trip of the CW pumps and subsequent reactor ,

scram on loss of condenser vacuum.

i SAFETY EVALUATION: Implementation of this temporary l alteration will only prevent an unnecessary trip of the  :

i CW pumps due to spurious lube water flow oscillations  !

and will thus reduce the chance of an inadvertent trip ,

of the CW pumps and subsequent plant scram on loss of i condenser vacuum.

Loss of the circulating water pumps during plant t operation has been evaluated in the SAR.

Per the SAR, the Circulating Water System serves no i safety function. System analysis has shown that a

failure of the Circulating Water System will not compromise any safety related systems or prevent safe i shutdown.

[146]

Attachment to GNRO-93/00001 I

PLS-92-007 '

Page 2 The radiological consequences remain the same as the  !

consequences of a loss of condenser vacuum as discussed l in the SAR.  !

The technical specifications do not contain any margins j of safety for the operation or design of the CW System.  ;

The changes do not affect or prevent safe shutdown of l the reactor vessel. Therefore, implementation of the ,

changes does not reduce the margin of safety as defined .

, in the basis for any technical specification.  ;

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Attachment to GNRO-93/00001  ;

SRASN: PLS-92-008 DOC NO.: W.O. #67130 DESCRIPTION OF CHANGE: The water in Standby Service Water (SSW) "A" or "B" basin is required to be pumped -

out to perform maintenance. The requirement to minimize time for pump down is very important since work cannot begin until the basin is empty. To pump down a basin two or three temporary electric sump pumps will be lowered to the bottom of the basin in a i quadrant without an SSW pump and connected to either i rigid or flexible piping. The pumps and associated  ?

piping are put in place prior to declaring the SSW basin inoperable. The piping from the pumps will be routed to either the Unit 2 circulating water basin or the Unit 2 discharge basin. To prevent a siphon from forming, the piping going to the discharge basin will have at least one joint not connected up in each line.

The line going to the Unit 2 circulating water pit will not extend down into the pit so a siphon can form. To prevent actual pumping of water, the breakers to the temporary pumps will be tagged open under the GGNS ,

tagging program.

The piping will have the final connections made up 1 after the SSW basin is no longer required to be i' operable. The tags will be cleared when ready to start 1

pumping.

REASON FOR CHANGE: To perform maintenance on SSW "A" or "B" basin.

. SAFETY EVALUATION: The SSW system is designed to supply cooling water and a heat sink for a safe shutdown of the reactor following a Design Basis '

Accident (DBA) Loss of Coolant Accident (LOCA). The I worse case accident evaluated involving SSW is a DBA LOCA with failure of one standby diesel generator which removes one of the SSW loops from service. The i combination of the temporary pumps being tethered, the  ;

low flow currents in the basin, the location of the temporary pumps, the construction of the basin with the slab over-hand and the screen over the intake of the SSW pumps will prevent any equipment or parts of equipment affecting operation of the SSW pump.

3 The installing of the pumps will not affect the operation of the SSW or any other system.

The closed loop design of the SSW system adds a barrier between the environment and radioactive material that may have leaked from the Residual Heat Removal (RHR) l i

[148]

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Attachment to GNRO-93/00001 4

1 PLS-92-008 Page 2 System. The possible pathway created by installing pumps and piping into the SSW basin is controlled so that the basin contents cannot be removed from the basin until plant conditions allow. The tagging open of the temporary pump breakers and the leaving open the final connections of piping ensures that the pathways from the basin are controlled so as to ensure basin contents remain in the basin. The leaking of contaminated SSW water from the basins is prevented; thus, preventing any increase in the consequences of an j accident previously evaluated in the SAR. ,

should the tethering device fail, the equipment being  !

installed would descend in a near vertical trajectory even if the SSW pump was running. The low fluid velocities present in the basin would not affect the equipment falling. The velocity of equipment free falling through water is much slower than through air.

The bottom of the SSW basins are made of 4-foot thick reinforced concrete. The force of the resultant impact -

of the equipment free falling through the basin fluid and striking the basin floor is not considered to be a significant hazard to the integrity of the basin. The probability of some piece of the temporary equipment  ;

getting pulled to the pumps is extremely low  :

9 considering the low fluid velocities and basin layout.

If some small piece of equipment did get to the pump sump, the suction screen would not allow anything into ,

the pump that would damage the system. The temporary equipment would not affect the system therefore is bounded by the existing analysis and not create the '

possibility of an accident of a different type than any evaluated in the SAR.

The installation of temporary pumps in the SSW basin will not affect system operation and therefore will not l affect its cooling ability. No margin to temperature j

limits will be decreased due to any degraded cooling 3

ability of the SSW system. The integrity of the SSW

, system will not be breached or challenged in such a manner to cause a release of system contents. Controls  !

will be in place to prevent the loss of water from the  :

SSW basin while it is required to be operable. The ,

margin to radiological limits for release to the  ;

environment will therefore not be reduced. The >

installation of the temporary pumps will not affect any functions of the SSW system and therefore will not

] reduce the margin of safety as defined in the basis for

]

any Technical Specification.

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Attachment to GNRO-93/00001 SRASN: PLS-92-009 DOC NO.: W.O. #28560 l

DESCRIPTION OF CHANGE: This temporary alteration will I install vibration probes on the Standby Service Water (SSW) fan shaft bearing pedestals inside the cooling ,

towers of SP41C003A, B, C, D. The vibration probes will be mounted on the outside surface of the bearing support using magnetic bases. The vibration probes are heavy enough and have a small enough surface area that the force from the air flow is not enough to draw the probes up into the fans if they should become detached from the surface they were installed on. The cables leading from the probes will be run along the grease line and out through a hole in the fan tower wall. The cables will be run under the drive shaft along a grease line support and secured to the greaseline below the fan with wire tires. The cables will be secured in the ,

motor rooms and connected to portable test equipment i under 17-S-03-25, " Portable Vibration Monitoring  ;

Program" when recording data.

REASON FOR CHANGE: To check the vibration on SSW fan shafts. j SAFETY EVALUATION: The SSW system is designed to  ;

supply cooling water and a heat sink for a safe i shutdown of the reactor following a Design Basis Accident (DBA) Loss of Coolant Accident (LOCA). The  !

worse case accident evaluated involving SSW is a DBA LOCA with failure of one standby diesel generator which ,

removes one of the SSW loops from service. The i installation of vibration probes will not remove a SSW  ;

loop from service because the probes cannot affect the fans or coolant flow through the tower. If the .

4 magnetic base probes were to become detached from the 4

pedestal support, they woald hang suspended by the l instrument cord below the fans or fall to the fill l

) material below.

The closed loop design of the SSW system adds a barrier 3 between the environment and radioactive material that i may have leaked from the Residual Heat Removal (RHR) -1 System. This temporary alteration will not breach the l system or create a path for system contents to enter '

the surrounding environment. The addition of vibration monitoring probes to the SSW fan shaft bearing pedestals will not impact the barrier protection that the SSW system gives between the environment and any radioactive material released from the RHR system.

[150] l l

i Attachment to GNRO-93/00001 !

{

PLS-92-009 i Page 2 l r

The SSW system is important to safety because it is  !

used to provide a heat sink to dissipate reactor heat in accident conditions as well as normal cool down. l The SSW system is the only system the probes are close i to. The malfunction of SSW being caused by the probes  ;

is not likely to happen. The probes would not be ,

sucked up into the fans because they are too heavy for the area they displace. If the probes fell, they would ,

not reach the pump suction because they would land on  !

the fill material. The failure of the probes to stay r installed would not affect the function of the SSW  :

, system and therefore would not increase the probability i of occurrence of a malfunction of equipment important  !

to safety previously evaluated in the FSAR. l The installation of vibration probes on the SSW fan .

. bearing pedestals will not affect system operation and i therefore will not affect its cooling ability. No l margin to temperature limits will be decreased due to l any degraded cooling ability of the SSW system. The integrity of the SSW system will not be breached or challenged by the installation of the probes so no release of system contents will be caused by them. The margin to radiological limits for release to the ,

environment will not be reduced.

The installation of the probes will not affect any  :

functions of the SSW system and therefore will not reduce the margin of safety as defined in the basis for any Technical Specification. l h

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Attachment to GNRO-93/00001 SRASN: PLS-92-010 DOC NO.: 06-OP-1000-D-0001 l R40, TCN 106 1

DESCRIPTION OF CHANGE: Procedure 06-OP-1000-D-0001 is ,

being changed to delete the requirement for use of a  ?

graph in determining sampling requirements for ,

Technical Specification 4.11.2.7.2.b. A fifty percent- l change in activity in four hours with an administrative limit of 100 mR will be used to meet the Technical l Specification requirement. '

REASON FOR CHANGE: Small fuel leaks have developed during fuel cycle five. A small change, from thirty to forty mR/hr increasing to fifty to seventy mR/hr, has been noted in the offgas pretreatment radiation monitor indication. During normal steady state power operation the increase has been small enough to pass the present surveillance test using the graph on Pages 5, 23 and 49 of 71, Attachment I of 06-OP-1000-D-0001. This graph ,

is characterized by a linear response to power caused .

by a fission gas release recoil distribution pattern.

Operation with a fuel failure negates the linear response of the fission gas. release and when reactor  ;

power is decreased a corresponding decrease in the ,

offgas pretreatment radiation monitor does not occur.  ;

This causes a situation requiring repeated chemistry  !

offgas analysis when there has not been an increase in -

offgas pretreatment activity. The graph relied upon for determining sampling requirements is useful only at ,

a steady state power and is overly conservative t requiring sampling when there has been no increase in activity. The graph is not representative of activity versus reactor power after a fuel leak has occurred and triggers sampling even though an increase in activity ,

has not occurred.

SAFETY EVALUATION: Using a fifty percent increase in activity in four hours to trigger action per Technical '

Specification 4.11.2.7.2.b is satisfactory because the purpose of the offgas pretreatment radiation monitor is to detect a change in condition of the fuel cladding.

A leak in a fuel assembly even a small one will be transported via the reactor coolant to the offgas pretreatment radiation monitor; therefore, fifty  ;

percent changes that occur at time intervals greater l than four hours need not be considered as part of the  !

surveillance requirement. To ensure that long term  ;

increases in activity are considered when determining sampling requirements, the administrative limit will

[152)

i Attachment to GNRO-93/00001 '

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l PLS-92-010 l Page 2 l require sampling even though there has not been a fifty percent increase in activity. This limit will be based i upon the condition of fuel being used and will be i

evaluated at startup from refueling outages and upon a change in condition of current cycle fuel.

The new method for determining offgas pretreatment  ;

sampling requirements will not increase the operational time before failed fuel is detected nor will it change the response to a fuel failure.

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Attachment to GMRO-93/00001 SRASN: PLS-92-013 DOC NO.: UFSAR 11.5.2.

3.2 DESCRIPTION

OF CHANGE: The change deletes the 15%

calibration accuracy value for continuous radiation monitors in the UFSAR and deletes the use of control charts for installed airborne radioactive monitoring systems.

B REASON FOR CHANGE: To correct inaccuracies in the UFSAR as identified by NRC Inspection Report  ;

50-416/91-19 and Quality Deficiency Report 0222-91.

SAFETY EVALUATION: The plant monitors / instrumentation affected by these UFSAR Sections, and the changes being evaluated, will continue to function as per the manufacturer's specifications. The accident bases evaluated in the UFSAR. remain unchanged and the possibility of an accident occurring as evaluated in the UFSAR is unaffected.

The accuracy requirements for individual calibrations are adequately controlled by individual surveillance i procedures. The instruments continue to perform their intended function within the limits of their design.  ;

The deletion of the requirement for control _ charts is consistent with industry practice and poses no change to the operational capability of these instruments, and they will continue to perform their design function.

As no change in instrument / monitor operation or capabilities occurs from this UFSAR change, the l probability for a malfunction of equipment important to  :

safety of a different type than any previously evaluated in the UFSAR is not created.

The margin for safety defined the basis for Technical  !

Specifications will not be reduced. Inconsistencies between these sections of the UFSAR and Technical Specifications are resolved by this change.

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Attachment to GWRO-93/00001 SRASN: PLS-92-014 DOC NO.: MWO 68359 l l

DESCRIPTION OF CHANGE: This work order covers the '

temporary installation of the fuel inspection station, i the reconstitution operation and the removal /decon of the reconstitution equipment prior to completion of ,

refueling activities and plant startup.

REASON FOR CHANGE: This work order is needed to perform bundle reconstitution.

SAFETY EVALUATION: All irradiated fuel assemblies will be handled in a normal manner with the fuel handling ,

platform and the fuel prep machine. The events leading up to a fuel handling accident are unchanged.  !

Operation of the fuel handling platform is controlled by plant procedure. Operation of the fuel prep machine is controlled by plant procedure also. Since all fuel  ;

assembly handling will be in accordance with plant  !

procedure there is no increase in the probability of a fuel handling accident (in the Auxiliary Building) l' UFSAR 15.7.4.

The fuel handling accident (in the Auxiliary Building), f UFSAR 15.7.4, assumes a channeled fuel assembly drops  !

from a height of six feet onto unchanneled bundles ,

stored in the fuel pool racks. The consequence of the fuel handling accident (in the Auxiliary Building) is the release of radioactivity. The release from a l bundle drop associated with inspection / reconstitution ,

are bounded by the release determined for the fuel i handling accident. All fuel bundle movements will be performed in the normal method per plant procedure. .

Individual fuel rod movements will be performed with  !

, the use of a fuel rod grapple. Any failure of the fuel rod grapple would result in an accident bounded by the

• fuel handling accident. An increase in consequences ,

could only occur if more than one' bundle dropped or it dropped from a distance greater than six feet. ,

, Administrative controls of crane operations per plant i procedure will remain in effect. In addition, no loads i greater than 1140 pounds will be moved by the new fuel bridge crane hoist. There is no increase in the  ;

consequences of the fuel handling accident (in the -

Auxiliary Building) during inspection / reconstitution .

operations since these activities are bounded by the  !

Puel Handling Accident Analysis (in the Auxiliary l Building). {

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Attachment to GNRO-93/00001

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i PLS-92-014 l

Page 2  ;

Since the fuel prep machine's seismic characteristics i are maintained and the spent fuel pool criticality l analysis is bounding, the probability of occurrence of i a malfunction of equipment important to safety ,

previously evaluated in the SAR has not increased.  !

Safety considerations are a significant part of the  :

Siemens Nuclear Power Corporation's inspection /

reconstitution equipment design. All tools used to ,

move irradiated components will be marked at 6' to l prevent an inadvertent reduction in the safe water ,

shield level. The worst malfunction of the inspection /  :

reconstitution equipment would result in a dropped fuel  ;

rod. Therefore, the fuel rod inspection / reconstitution >

operation is bounded by the previously evaluated fuel i handling accident (in the Auxiliary Building).  ;

Therefore, no new failure modes are created and no possibility for creating a malfunction of equipment important to safety different than previously evaluated -

in the FSAR. j

. The fuel rod inspection / reconstitution operation will l 4

not require any changes to normal fuel assembly j handling nor will there be any changes to crane  !

operations in the Auxiliary Building. As a result, the margin of safety as defined in the basis for Technical 4 Specification 3/4.9.6, " Refueling' Equipment" and j 3/4.9.7, " Crane Travel" will not be reduced.

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Attachment to GNRO-93/00001 y SRASN: PLS-92-015 DOC NO.: MWO 66729 -i DESCRIPTION OF CHANGE: This work order covers the temporary installation of the fuel sipping equipment  !

(sipping cans, operating consoles), the sipping  ;

operation, and the removal /decon of the sipping-equipment prior to completion of refueling activities  ;

and plant startup.

1 REASON FOR CHANGE: This work order is needed to perform sipping operations.  ;

i SAFETY EVALUATION: All irradiated fuel will be handled i in a normal manner with the fuel handling platform. ,

The events leading up to a fuel handling accident are l unchanged. Operation of the fuel handling platform is  !

controlled by plant procedure.  :

The fuel handling accident (in the Auxiliary Building), ,

UFSAR 15.7.4, assumes a channeled fuel assembly drops {

from a height of six feet onto unchanneled bundles l stored in the fuel pool racks. The consequence of the  ;

fuel handling accident (in the Auxiliary Building) is ,

the release of radioactivity. The release from a

• bundle drop associated with sipping are bounded by the i release determined for the fuel handling accident. All  !

fuel bundle movements will be performed in the normal  !

method per plant procedure. An increase in {

conscquences could only occur if more than one bundle ,

, dropped or it dropped from a distance greater than six  !

feet. Administrative controls of crane operations per i plant procedure will remais in effect. In addition, no ';

loads greater than 1140 pounds will be moved by the new  :

fuel bridge crane hoist. Since-all fuel handling will  !

be in accordance with plant procedure, and no loads l greater than 1140 pounds are involved, there is no  ;

increase in the consequences of the fuel handling i accident (in the Auxiliary Building) during sipping  !

operations.  !

. The additional weight placed in rack H1 (sipping cans) I beyond the weight currently installed (4 garbage cans,  ;

1 defective fuel can, and 2 weighted half guides) will i not alter the rack's ability to withstand all credible  ;

static and dynamic loadings as evaluated in UFSAR j 9.1.2.1.1.1. The rack's integrity prevents damage to  !

the contained fuel. Protection of the contained fuel ,

from excessive damage may mitigate the release of

• radioactive materials during normal or abnormal l conditions.  !

e

[157] l l

Attachment to GNRO-93/00001 PLS-92-015 Page 2 i

Safety considerations are a significant part of the [

General Electric vacuum sipping equipment design. The key safety considerations are that a fuel assembly .

should not be damaged by either inadvertent-lid l openings / closings or be allowed to overheat while f

, inside the vacuum sipper can. Such safety systems as j the 1) can temperature monitor, 2) water carry-over

, trip, 3) service air overpressure trip, and 4) lid isolation in existing position upon system loss of '

i power helps to assure continual safe sipping operations.

The sipping operation will not require any changes to normal fuel handling nor will there be any changes to +

crane operations in the Auxiliary Building. As a ,

result, the margin of safety as defined in the basis for Technical Specifications 3/4.9.6, " Refueling  ;

Equipment" and 3/4.9.7, " Crane Travel" will not be  !

reduced. I i

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[158] i

Attachment to GMRO-93/00001  !

t SRASN: PLS-92-016 DOC NO.: 05-1-02-VI-2, Rev. 17 DESCRIPTION OF CHANGE: Division 1 and 2 diesel ,

generators are required to be started, loaded, and the  ;

associated buses separated from offsite power during tornado warnings and severe weather. This change eliminates this requirement, leaving the diesel generators in a normal condition.

REASON FOR CHANGE: Separating the emergency buses from offsite power during severe weather provides a i potential for plant transients in the event of operator error or equipment malfunction. This change reduces this potential.

SAFETY EVALUATION: Accident analyses assume emergency diesel generators are in standby conditions at the >

onset of the accidents and, although diesel generators are also required to respond properly if already in operation, no safety benefit is apparent. A loss of offsite power will result in a plant transient whether or not the emergency buses are tied to offsite power, ,

while the practice of separating the buses from offsite sources increases the potential for a transient. NRC  ;

Information Notice 84-69 also supports this conclusion.  ;

The only event of concern related to this change is a  ;

loss of power event, which can initiate accidents such >

as increase in reactor pressure, decrease in reactor  :

coolant flow rate, and decrease in reactor coolant

inventory. The probability of a loss of power event as analyzed in the UFSAR is not increased by this change, ,

because this probability is governed by factors such as-number, reliability, and separation of offsite sources  !

and is unrelated to diesel generator availability or  ;

status. A loss of power event is not prevented by operating the diesel generators.

l The UFSAR requires electrical power availability l (offsite or onsite) in order to mitigate the '

consequences of a number of accidents, such as loss of coolant accidents, feedwater line break outside containment, and steam system piping break outside containment. None of these accidents require the ,

, diesel generator be in operation prior to occurrence of  :

the accident. Radiologic &2 consequences of accidents  !

analyzed in the UFSAR are therefore not affected by ,

this change, since the assumptions of these analyses i are maintained (i.e., diesel generators in standby when  :

accident occurs).

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[159] j l

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Attachment to GMRO-93/00001 f

PLS-92-016 Page 2

?

The diesel generators, as described in the UFSAR, are '

designed to automatically start and load when necessary. This ability is demonstrated every 18 months as required by Technical Specification 4.8.1.1.2.d and 4.8.1.2. While the probability of a loss of offsite power is greater during severe weather,  ;

this potential is assumed in the UFSAR, Section 8.2,  ;

Offsite Power System, and is not affected by this  !

change.

f The malfunction of equipment related to this change is ,

a loss of voltage to emergency buses. Consequences of loss of voltage to a bus are not affected by this change, since the change only affects the onsite and offsite AC sources supplying the emergency buses. The assumptions of the UFSAR related to AC source alignment are preserved by this change.

Electrical power is required to be provided to the '

emergency buses to preserve margins of safety, for example, for fuel temperatures and containment pressure after a loss of coolant accident. This change preserves the assumptions for offsite and onsite power i

, sources; therefore, since the change affects nothing

• t else, none of these margins of safety are reduced. All accident and transient barriers are maintained as assumed. ,

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Attachment to GNRO-93/00001 i

1 SRASN: PLS-92-017 DOC NO.: OpCon 4 or 5 in Action c or d DESCRIPTION OF CHANGE: This safety evaluation documents the analysis of entry into Operational i Condition 4 or 5 when one or more suppression pool l level instrumentation divisions are inoperable. This  !

evaluation specifically addresses the use of Technical Specification (TS) 3.0.4 to enter Operational Condition 4 or 5 while complying with Action Statement c or d of TS 3.5.3. The evaluation makes the following assumptions: -

a. Either or both suppression pool level instrumentation divisions may be inoperable.  ;

b. An alternate indicator of suppression water level is utilized in accordance with the applicable Action Statement (c or d).

c. Operational Condition 4 is entered from Operational Condition 3 or 5. Operational Condition 5 is entered from Operational Condition 4 or 5 (high water level).

NOTE: This evaluation does not cover entry into Operational Condition 4 from Operatienal Condition i 3 with 2 suppression pool instrumentation divisions inoperable.

d. Lowering the reactor cavity water level is not an [

operation with the potential to drain the reactor  !

vessel, because the cavity drains are external to i the reactor vessel at or about the reactor vessel flange elevation. I

e. With no suppression pool level instrumentation  !

OPERABLE, there are no ongoing evolutions with the j

possibility of depleting suppression pool inventory or draining the reactor vessel in l progress. I REASON FOR CHANGE

During refueling outage activities,

situations arise where suppression pool level  :

instrumentation is inoperable in order to perform maintenance activities, surveillance testing, or design i modifications. In these situations, utilizing the flexibility allowed by TS 3.0.4 will reduce additional  ;

and unnecessary activities and complications which  :

would be necessary to comply with without the use of ,

TS 3.0.4.

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, [161) c l

I l Attachment to GNRO-93/00001 PLS-92-017 Page 2 SAFETY EVALUATION: Under the actions of TS 3.5.3 in Operational Condition 4 and 5, with one or more division of suppression pool level instrumentation inoperable, the inoperable division (s) is to be .

restored to OPERABLE or verify suppression pool water I level is greater than or equal to 12 feet 8 inches at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by an alternate indicator. The suppression pool is required to be OPERABLE as part of the ECCS to ensure a sufficient water supply is available to the HPCS, LPCS, and LPCI systems in the event of a LOCA. The suppression pool level instrumentation is not required OPERABLE in Operational Condition 4 or b for the Suppression Pool Makeup System or for accident monitoring. The use of TS 3.0.4 does not alter the function or operation of either ECCS or the suppression pool or introduce new failure mechanisms. Complying with Action c or d of TS 3.5.3 ensures the suppression pool level is maintained such that the minimum water level required based on net positive suction head (NPSH), recirculation volume and vortex prevention is maintained. This application of TS 3.0.4 does not change design features or plant operating procedures or introduce any new mechanism for draining the reactor vessel or the suppression pool.

The NRC reviewed and approved a oi.e timo exception for use during RF03 to the "old" TS 3.0. requirements via amendment 58 to the GGNS TS. This ext.cotion was ~

applicable for entry into Operational Condition 5 from 4 or 4* (high water level). In addition, safety evaluations PLS 030/90 and NL-SE-90-025 evaluated the use of this flexibility for LCO 3.5.3. This safety evaluation encompasses these previously evaluated conditions.

Because of the above considerations, reliance on Action c or d of LCO 3.5.3 after entry into the specified operational condition as opposed to relying on these actions after having already been in that specified condition, is not adverse to safety.

Therefore, the application of TS 3.0.4 to TS 3.5.3 Action c or d is acceptable.

[162]

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J SRASN: PLS-92-018 DOC NO.: Tech Spec 3/4.9.7 l l

l f DESCRIPTION OF CHANGE: This change modifies the bases l statement for Technical Specification 3.9.7 to agree i with the UFSAR and NRC correspondence. The bases  :

currently states that movement of non-fuel loads over  :

the fuel assemblies in the storage pools is restricted i to non-fuel loads weighing less than the nominal weight of a fuel assembly and that the activity released in l the event that this load was dropped would be limited  ;

to that contained in a single fuel assembly. As (

discussed in UFSAR Section 15.7.4, UFSAR Appendix 9D, j SER Supplement 5, AECM-81/427, AECM-82/17, and  ;

GNRO-91/00129 the current technical specification value  ;

is actually based on the weight of a channeled fuel l assembly and its associated handling tool (when the i NF400 mast is in use). As discussed in UFSAR Sections 15.7.4 and 9.1.4.3 and UFSAR Appendix 9D, the  :

activity released in the event of a load drop is  !

limited to well within the 10CFR100 limits.

Inadvertently, this bases statement was not mooified when the Technical Specification value was changed as  :

the result of analysis performed in accordance with j NUREG-0612 (AECM-81/427). ,

t REASON FOR CHANGE: Update the description of the bases I for Technical Specification 3.9.7 to reflect the bases i for the Technical Specification value as identified in j the UFSAR, NUREG-0612, 9nd NRC correspondence.  !

SAFETY EVALUATION: Although this change modifies the i i definition of a heavy load as presented in the bases l

) for Technical Specification 3/4.9.7, it does not change j l the probability of the occurrence that a non-fuel load will drop onto fuel bundles in the storage racks. As discussed in UFSAR Section 15.7.4, UFSAR Appendix 9D, AECM-81/427, AECM-82/17, and SER Supplement 5 the drop 3 of a non-fuel load onto the storage racks has been 4

evaluated for loads up to 1140 pounds. AECM-81/427 transmitted the proposal to change the TS value to heavy load limit of 1140 pounds. This change to the TS was evaluated and found to be acceptable in SER Supplement 5, Appendix L. These documents state: "A j heavy load is defined as a weight exceeding the weight of a channeled fuel assembly and its associated handling tool (approximately 1140 pounds) " . This change does not modify the maximum allowable non-fuel load which can be transported over spent fuel in the

, storage racks, change the load handling devices which ,

can transport these loads, or any other administrative  !

j control over the handling of these loads. l

[163] ,

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Attachment to GNRO-93/00001 PLS-92-018 Page 2 I

UFSAR Sections 15.7.4 and 9.1.4.3 and UFSAR Appendix 9D {

evaluate the consequences of a heavy load drop onto the i

spent fuel in the storage racks. These sections  ;

conclude that the heavy load limit ensures the  !

radiological effects are below the does limits '

specified in NUREG-0800, Standard Review Plan 15.7.4 (i.e., well within the guidelines of 10CFR100). .This change removes the statement that the heavy load limit  !

ensures "the activity release will the limited to that  :

contained in a single fuel assembly". The basis of  :

this limit is to ensure that the release is limited to  ;

that assumed in the safety analysis, i.e., to well i within the guidelines of 10CFR100.

[164]

Attachment to GNRO-93/00001 i

SRASN: PLS-92-019 DOC NO.: TEMP ALT 92-0017 DESCRIPTION OF CHANGE: The changes made by this temporary alteration will supply demineralized water from the makeup water treatment (MWT) system to the refueling floor during RF05. A 1-1/2" non-collapsible  !

hose will be connected from the MWT system at wye strainer 1P21-D010 to a booster pump located on i Elevation 208 of containment. The booster pump will i supply water to the spray header of the dryer / separator i strongback to provide a covering spray during transfers l and to prevent the possibility of airborne radioactive ,

particles. The pump will be powered from a 480 V .

welding receptacle, and the power cord will be enclosed in conduit. The temporary alteration shall be removed .

, before the plant has entered Mode 2, and the booster [

pump along with hard piping will be made permanent l plant equipment.  ;

REASON FOR CHANGE: When the dryer or separator is transferred between the reactor and the upper containment pool, they can dry which results in airborne radioactivity and subsequent containment ,

contamination. To prevent this from occurring, a spray i

l header is used to provide a covering spray during component transfers. The present water supplies on  :

Elevation 208 of containment are Fuel Pool Cooling and  !

Cleanup (G41) and Condensate & Refueling Water Transfer i (Pil) systems which are both potentially contaminated  ;

and not suitable for this purpose. i i

, SAFETY EVALUATION: The temporary alteration will not

present a change to technical specifications nor will  ;

it impose any new Technical Specification requirements.

Possible failures of the non-collapsible hose were j considered, and it was determined by researching l postulated piping failures in the FSAR that a failure would not alter the operation of any safety-related system or component. During use of the booster pump,  !

] the demineralized water jockey pump will be operated to pressurize the MWT piping to prevent the possible ,

introduction of radioactive water into the system.  !

Also, a check valve will be installed onto the end of the hose that is connected to the pump suction piping to prevent back flow of potentially radioactive water into the system if MWT piping pressurization is lost.

The pump will be powered from a non-divisional welding .

receptacle, and ampacity of this feed will be in accordance with Article 310-15 of the National Electric  !

Code. The power cord will be enclosed in its own i

i [165)

Attachment to GNRO-93/00001 l

PLS-92-019 Page 2 conduit and routed in accordance with separation requirements of Regulatory Guide 1.75. It has been determined that the temporary alteration will not ,

affect the function of the MWT system nor will it alter l t the operation of any safety-related equipment or >

component. It will not increase accident or malfunction probabilities or consequences. It will not create any risk of a different type of accident or  !

malfunction and does not reduce a margin of safety as  ;

described in the Technical Specification bases. >

Therefore, the temporary alteration will not involve an  ;

unreviewed safety question. .j i

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[166) r

Attachment to GNRO-93/00001 i SRASN: PLS-92-020 DOC NO.: TSTI-1B33-92-011-0-S I

i DESCRIPTION OF CHANGE: This document evaluates the chemical decontamination of the Reactor Water Cleanup (RWCU) System, Reactor Recirculation Loop A, and Reactor Recirculation Loop B using the Pacific Nuclear Services LOMI-AP-LOMI process. This document reviews the implementation procedures, the heavy load evaluation, disposal of liquid and solid wastes, and j chemical and radioactive liquid spills.

REASON FOR CHANGE: Chemical cleaning and i decontamination of the RWCU and recirculation systems is necessary to reduce the radiation exposure of  :

personnel working on or in the vicinity of these l' systems.

. SAFETY EVALUATION: Engineering Evaluation Response l (EER) 92/6026 documents the heavy load evaluation for j installation of the Pacific Nuclear (PN) chemical decon  ;

equipment. The evaluation provides: (1) evaluation of containment building structural beam (located at Elevation 161'-10" and at-135'-4" area 11) for Seismic l II/I concerns, (2) evaluation of steel grating at 114'-  ;

0" containment and drywell for structural adequacy, (3) evaluation of a temporary platform at 114'-0" over the  !

suppression pool to support the PN booster pump, (4) restraint requirements for PN flexible hoses, and (5) evaluation for the temporary 4" PN piping. The ,

e I

evaluation concurred with the decontamination issues  :

listed above and identified no adverse impact. ,

Temporary electric power for the PN equipment is

. provided from non-safety related BOP breakers. This ,

temporary power is not routed with, nor does it affect  ;

any divisional power trains.

1

The decontamination of the RWCU and reactor l recirculation systems will change area radiation dose i

. rate when the Vanadous Formate (VF) is injected. Test j steps are identified with a warning prior to  !

implementation of the step. Also, prior to injection, l Health Physics (HP) will be notified per the test .

J procedure. Radiation work permits under the direction of the HP department will ensure appropriate controls '

are in place to ensure that personnel exposure to radiation and radioactive materials is within the requirements of 10CFR20 and that such exposure is kept as low as reasonably achievable (NLARA) . Water shields will be installed to minimize radiation from the ion

< exchange columns. With the administrative controls in 1

[167)

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Attachment to GNRO-93/00001 PLS-92-020 Page 2 place and special precautions including the water y shields and lead blankets, no adverse radiological ,

affects are anticipated as a result of the chemical l decon activities.

The chemical cleaning process has no impact on the ability of the reactor recirculation system or the RWCU system to perform any safety function. The system /

loop to be deconned will be isolated and out of service during the decon process.

The resulting solid radwaste will be placed in a high integrity container and disposed of in accordance with existing plant radwaste procedures for shipping ,

offsite.

In conclusion, implementation of the TSTIs to ,

chemically decon portions of the RWCU and reactor recirculation systems will not increase the probability or the consequences of any accident evaluated in the i SAR, does not create the possibility of a new accident ,

or malfunction, and does.not reduce any margin of safety defined in any Technical Specifications. No  !

unreviewed safety questions exist.

o

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[168)

Attachment to GNRO-93/00001 i

SRASN: PLS-92-021 DOC NO.: WO #64619 l DESCRIPTION OF CHANGE: ESF Bus 16AB, System R21, j provides power to safety and non-safety related components and instrumentation. Required maintenance and cleaning of the 16AB ESF Bus requires that it be  :

de-energized for approximately 24-48 hours. This work  ;

will be conducted when the reactor is in Mode 4.

This safety evaluation addresses the operability concerns associated with supplying temporary power from -

BOP Buses 11HD and 12HE to loads normally supplied by l Bus 16AB.

REASON FOR CHANGE: During maintenance and cleaning of 16AB ESF Bus, provide temporary power to those loads ,

normally supplied by the 16AB Bus with power from Buses 11HD and 12HE. .

SAFETY EVALUATION: Temporary power will be supplied in

.f a similar manner as the normal power supply. Cable sizing and breaker selection will be such that adequate .

circuit protection is maintained. The loads being  !

supplied temporary power will not be relied upon to perform a safety function.

~

With Technical Specifications met by Divisions I and/or I III, shedding and sequencing of the affected loads is  ;

not required during postulated accidents. These loads  ;

will be powered from Buses 11HD and 12HE for [

convenience only and will be treated similar to-other  !

BOP loads during accident conditions. j Using Buses 11HD and 12HE does not diminish the quality of power to the temporary loads, nor does it decrease  ;

the reliability of the power available to the 2 nads i normally supplied by those buses. In addition, the j circuits that will receive temporary power do not i perform any safety functions or aid in the mitigation of accidents.  !

l The only possible failure of the circuits supplying i i tem,rorary power is their loss of power. Regardless of i how that loss occurs, the end result is failure of i component to function. Loss of power to all components has already been considered and is postulated as .

acceptable. i This change w'.ll not increase the probability or the j consequences of any accident evaluated in the UFSAR, does not creace the possibility of a new accident or .

malfunction, t.nd does not reduce any margin of safety defined in any Technical Specifications.

[169] l

1 Attachment to GNRO-93/00001 '

SRASN: PLS-92-022 DOC NO.: WO #67130 l

l DESCRIPTION OF CHANGE: The water in Standby Service Water (SSW) "A" or "B" basin is periodically required to be pumped out to perform maintenance. The  ;

requirement to minimize time for refill is very j important since testing requiring the basin full will be held up. To fill the basin as quickly as possible, i the normal fill line is supplemented with a line from '

the operable basins fill line. The water that was l pumped into the unit two circulating water pump pit i will be pumped back to the empty basin as well, this is  !

less than 1/7 of the total water volume required. The  ;

temporary line connected to the cperable basin's fill i line requires disconnecting the automatic fill piping i to the operable basin and gagging open the fill control ,

4 valve to allow flow to the drained basin. The  !

operability of the basin only requires that the basin contain a 30-day supply of water. The automatic fill  ;

system makes maintaining the level more convenient by }

removing the need for operators to take action. The  !

temporary line going to the drained basin will have a  :

branch line with a valve on it to manually add water to {

the operable basin. i i REASON FOR CHANGE: To refill SSW "A" or "B" basin as quickly as possible.

l j SAFETY EVALUATION: The SSW system is designed to '

9 supply cooling water and a heat sink for a safe shutdown of the reactor following a DBA LOCA. The water level in a SSW basin must be maintained above the 130'3" to perform its designed function in all modes of  !

operation. The water level in the SSW basin will be i controlled manually inside its normal operating band.

n The maintaining of basin level in its normal operating  ;

range will allow the SSW system to be operable and meet all requirements of previously evaluated accident ,

scenarios.  ;

) '

The closed loop design of the SSW system adds a barrier between the environment and radioactive material that may have leaked from the RHR system. The connection of temporary piping to the fill line of the operable basin will not create any new pathway for leakage from the d

SSW system.

The SSW basin fill system is from the non-safety related Plant Service Water (PSW) system and is not considered to be available in an accident. Redirecting the automatic fill line to the drained basin will

[170]

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Attachment to GNRO-93/00001 ,

PLS-92-022 Page 2 affect only the equipment associated to the automatic i fill system. Manual control of basin level will ensure basin level is maintained. Only non-safety related equipment will be modified.

The SSW basins contain a 30-day supply of water without makeup capabilities. The required level is assured by the level monitoring system which reads out in the

] Control Room and has a high/ low alarm. The level i

monitoring system will be in service for the operable  ;

basin. The low alarm is 6" above the Technical Specification required level, this will allow enough i time to direct makeup water to the operable basin. i The installation of a temporary manual fill system for j a SSW basin will not affect SSW pump operation and  !

therefore will not affect the system's cooling ability. i This change will not increase the probability or the  !

4 consequences of any accident evaluated in the UFSAR, does not create the possibility of a new accident or malfunction, and does not reduce any margin of safety ,

defined in any Technical Specifications. t 4

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Attachment to GNRO-93/00001 SRASN: PLS-92-023 DOC NO.: WO #67130 l

DESCRIPTION OF CHANGE: The water in the Standby l Service Water (SSW) "B" basin is required to be pumped 1 out to perform maintenance. To pump down one basin  ;

requires blanking off the siphon line between the two basins. Technical Specifications define an operable basin as having a 30 day supply of water either self contained or by means of an operable siphon. The siphon will be blanked off with a flange on the end in the SSW "B" basin. i

. REASON FOR CHANGE: To perform maintenance on SSW "B" '

basin.

SAFETY EVALUATION: The SSW system is designed to supply cooling water and a heat sink for a safe shut '

down of the reactor following a DBA LOCA. The probability of occurrence of most accidents evaluated

in the SAR reduces greatly as the plant goes to j Operational Conditions 4 and 5. The radioactive i'

release from a subsystem and component is the least  :

affected accident with the change to operating modes 4  !

4 and 5. The draining of SSW "B" basin will not affect any other systems other than those in the division  !

associated with the SSW "B" basin. The conditions of .

GGNS Technical Specification will be met.

The closed loop design of the SSW system adds a barrier between the enviromment and. radioactive material that '

may have leaked from the RHR system. The SSW "B" -

system will be inoperable and the SSW pump will not be running to cause flow to or from the basin. The RHR  :

"B" system will also be inoperable. The introduction i 4 of radioactive material into the basin as it is being 1 drained is highly unlikely due to no flow in the SSW l 2

"B" system and RHR "B" system. Since radioactive material will not be in the draining basin radioactive

discharge or exposure limits will not be exceeded. The '

siphon line plug is a blank flange bolted to the SSW ,

"B" basin side of the SSW siphon line. The blind l l' flange prevents any transfer of contents from the operable basin to the inoperable basin thus isolating ,

the two basins. The isolating of the SSW "A" basin  ;

from the SSW "B" one maintains the barrier between the environment and any radioactive material introduced into the SSW "A" system.

This change will not increase the probability or the ccnsequences of any accident evaluated in the UFSAR, does not create the possibility of a new accident or

malfunction, and.does not reduce any margin of safety j defined in any Technical Specifications.

[172) I

Attachment to GNRO-93/00001 I i

SRASN: PLS-92-024 DOC NO : EER-92/6127 DESCRIPTION OF CHANGE: This Engineering Evaluation  ;

Response (EER) evaluates temporary installation of  ;

blind flanges in place of the strainer screens on  ;

suppression pool suction strainer 1E12D008 to support modification of valve 1E12F004B, Residual Heat Removal l (RHR) Suction Valve Train "B".  !

REASON FOR CHANGE: This change is being made to provide isolation from the suppression pool water inventory while piping between suction strainer D008 ,

and valve F004B is being modified. These flanges will prevent the potential for a non-isolable leak from the '

suppression pool to the RHR "B" room during .

modification activities.

SAFEIT EVALUATION: The safety evaluation determined ,

that temporary installation of the blind flanges does .

not require a change to the GGNS Technical l Specifications or represent an unreviewed safety  !

question because installation requirements specified in i this EER will prevent a loss of suppression pool l inventory, and eliminate any potential effects on other Emergency Core Cooling Systems (ECCS) Systems. The RHR "B" system will be inop: dale during modification j

, activities and for the p -iod that the blind flanges  ;

are installed. System et 4.guration will be restored l to normal prior to declaring the RHR "B" system {

cperable. i i

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Attachment to GNRO-93/00001 SRASN: PLS-92-025 DOC NO.: EERR-92/6109 i

l

) DESCRIPTION OF CHANGE: The change allows for the i storing of fuel bundles in spent fuel pool (SFP) rack

, H1 inside a defective fuel storage container and the storage of fuel rods in a fuel rod storage basket in a defective fuel storage container in SFP rack Hl.

REASON FOR CHANGE: This 50.59 is needed for storage of '

defective fuel.

SAFETY EVALUATION: The storage of defective fuel in rack H1 is acceptable subject to decay and loading restrictions. All irradiated fuel assemblies will be handled in a normal manner with the fuel handling i platform. The events leading up to a fuel handling accident are unchanged. ,

The consequence of the fuel handling accident (in the Auxiliary Building) is the release of radioactivity.

The release from a defective fuel storage container  ;

drop with the fuel rod storage basket is bounded by the  :

release determined for the fuel handling accident. The l additional weight placed in rack H1 will not alter the )

rack's ability to withstand all credible static and dynamic loadings. The geometric configuration of r irradiated fuel stored in non-adjacent locations in H1 with at least one empty cell between them and the other. -

fuel racks will maintain the design basis in the spent fuel pool. The storage of 32 fuel rods in the fuel rod storage basket inside a defective fuel storage container will also maintain the design basis in the spent fuel pool. The storage of fuel assemblies or the ,

fuel rod storage basket inside a defective fuel storage [

container will in no way impair the structural or i criticality characteristics of rack Hl.

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i Attachment to GNRO-93/00001 i i

SRASN: PLS-92-026 DOC NO.: 07-S-14-184,  !

Rev. 11, TCN #13 ,

i DESCRIPTION OF CHANGE: This change allows for the I

lifting of the vessel head without stud protectors installed. i REASON FOR CHANGE: A change to the stud protector used j during past outages, prior to RF05, from aluminum to i stainless steel and the machining of the stainless stud ,

protectors has necessitated the need to lift the vessel  !

head without the protectors installed. ,

SAFETY EVALUATION: Evaluation of not using stud

• protectors during head removal only creates a risk of i damage to the stud itself. It does not establish any j accident of a different type or equipment malfunction -

and would only create a maintenance rework / replacement  ;

of any damaged studs should that occur.

}

Accidents evaluated for polar crane heavy load drop of the vessel head will not be affected by this action.  !

Current load drop analysis of the vessel head shows that such postulated occurrences would not affect the  ;

availability of the plant to remain shutdown or result in release of significant amounts of radioactive  !

materials. t i

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Attachment to GNRO-93/00001 I SRASN: PLS-92-027 DOC NO.: Plywood Containment Hatch DESCRIPTION OF CHANGE: The temporary plywood

. containment hatch cover is installed during irradiated l fuel handling and during evolutions with the potential to drain the vessel. The change will require that the hatch cover be staged but not installed unless required under direction from the Control Room. ,

REASON FOR CHANGE: It has been necessary in past refueling outages to remove and replace the hatch cover frequently in order to allow movement of large pieces of equipment into and out of containment. '

SAFETY EVALUATION: Use of the temporary hatch cover is  !

not credited as a mitigative measure in any safety analyses or other design or license basis documents.

For potential events of concern (e.g., fuel handling .

accident, vessel draindown and loss of shutdown l cooling), sufficient time exists to install the hatch cover prior to release of radioactive material from the plant. Therefore, no unreviewed safety question exists for this change.

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Attachment to GNRO-93/00001 i SRASN: PLS-92-028 DOC NO.: WO #71304 i

DESCRIPTION OF CRANGE: Connect construction water (CW) to 2P53C001, Unit 2 Instrument Air Compressor and SP52C001B, Service Air B Compressor as a temporary source of cooling water.

REASON FOR CHANGE: Cooling Water Isolation Valve ,

1P43F275 has a broken yoke and must be replaced. l Valve 1P43F275 must be isolated for replacement to be i

installed. When Valve 1P43F275 is isolated then a temporary source of construction water must be used for i Unit 2 Instrument Air Compressor and Service Air B ,

Compressor. .

SAFETY EVALUATION: The loss of instrument air is <

evaluated in the FSAR as an incident of moderate  !

frequency with no radiological consequences. The total loss of instrument air evaluated in the FSAR states that the equipment using instrument air is designed to fail to a safe position. Construction water provides a good source of cooling water for the instrument and service air compressors' oil coolers and air coolers. i The loss of service air would have no bearing on plant operation or safe shutdown. l i

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Attachment to GNRO-93/00001 i

i SRASN: PLS-92-029 DOC NO.: TSPS 128, Rev. 0 1 DESCRIPTION AND REASON FOR CHANGE: The safety I evaluation documents the evaluation of the effect rf ,

the Division II ESF switchgear room coolers being out of service during a refueling outage (Operational  ;

Condition 4 or 5) with temporary ventilation provided  !

to maintain temperature below the Technical i Specification limit of 104*F.

SAFETY EVALUATION: Calculations were performed to  !

determine how much, if any, cooling is required to l maintain these rooms within technical specification limits while the associated room coolers are out of i service. The calculations show that minimum flow rates i for the following rooms must be maintained:  ;

1A221 - 1382 CFM l' 1A207 - 208 CFM 1A308 - 1313 CEM 1A320 - 253 CFM i 1A407 - 45 CFM Previous tests and calculations have shown that j equipment in those rooms will remain functionally l capable of performing the safety functions at ,

temperatures well in excess of 104 F. The alternate  ;

method of room cooling establishes sufficient air flow t to maintain room temperatures below this limit during this period of time when Division II heat loads are  ;

minimal.  ;

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i Attachment to GNRO-93/00001 I l

SRASN: PLS-92-030 DOC NO.: EER-92/6127 i

1 DESCRIPTION OF CHANGE: This Engineering Evaluation  ;

Response (EER) evaluated the temporary installation of f blind flanges in place of the strainer screens on suppression pool suction strainer 1E12D008 to support j modification of valve 1E12F004A. The same modification will be made during the RHR A subsystem outage of RF05 to the RHR A suction strainer 1E12D007 to support )

modification implementation on valve 1E12F004A. .

3 REASON FOR CHANGE: As stated above, this change is being made to provide isolation from the suppression pool water inventory while piping between suction ,

strainer D007 and valve F004A is being modified. These ,

flanges will prevent the potential for a non-isolable leak from the suppression pool to the RHR "A" room i during modification activities.

SAFETY EVALUATION: The safety evaluation determined [

that temporary installation of the blind flanges does  :

not require a change to the GGNS Technical  !

Specifications or represent an unreviewed safety

, question because installation requirements specified in l this EER for 1E12D008 will be applied to 1E12F007 and >

these requirements will prevent a loss of suppression pool inventory, and eliminate any potential effects on

other Emergency Core Cooling System (ECCS) Systems.

The RHR "A" system will be inoperable during l modification activities and for the period that the blind flanges are installed.

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Attachment to GNRO-93/00001-l SRASN: PLS-92-031 DOC NO.: MNCR 171-92 DESCRIPTION OF CHANGE: Evaluated the need for additional inspections of main turbine stop and control valves in accordance with the requirements of Technical Specification 3/4.3.9 due to the material' condition of the 1N11-F026A valve seat.  ;

REASON FOR CHANGE: During the inspection of the  !

1N11-F026A stop valve several defects were identified on the seat. Two of these defects are of sufficient depth (approximately .5 mm to .7 mm) so that l operability of the valve must be evaluated to determine  ;

the necessity for further valve inspections. i SAFETY EVALUATION: The material condition of the seat for valve IN11-F026A (two chips in seat approximately i 7/32" wide and .5 mm to .7 mm deep) does not present an i operability concern for this valve. The valve continues to provide overspeed protection for the turbine as designed and therefore the chips in the valve seat are not an unacceptable flaw. There is no evidence of a generic material or configuration flaw that would require an inspection of any of the other main turbine-stop/ control valves. Therefore, no  !

further stop/ control valve inspections are necessary.

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Attachment to GNRO-93/00001 i

SRASN: PLS-92-032 DOC NO.: 07-S-14-186, R12 DESCRIPTION OF CHANGE: This safety evaluation assesses the transport of the reactor steam separator from its  ;

storage location to the reactor vessel without being .;

submerged underwater. It has been past practice to transport the steam separator underwater in the upper containment pool, except for the brief period of time ,

necessary to clear the weir wall. The UFSAR will be .

i changed to clarify that the timing of the flooding with respect to separator installation is not critical.

t REASON FOR CHANGE: This evolution allows scheduling ,

flexibility during refueling outages.  !

e SAFETY EVALUATION: During previous refueling outages .

at GGNS, the steam separator has been transported from l its storage location back to the reactor vessel while  :

remaining submerged for the most part under water.

This transport method is being changed to permit  !

J returning the steam separator back to the reactor t vessel, while not submerged. This change in transport method for the steam separator is dependent on meeting the applicable criteria of NUREG-0612, Section 5.1.1  ;

.(Phase I requirements), for handling heavy loads over  !

irradiated fuel. A review of the referenced documents ,

has determined that this change continues to satisfy the applicable requirements of NUREG-0612. j 1

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Attachment to GNRO-93/00001 -

SRASN: PLS-92-033 DOC NO.: TCN 35 to '

04-1-01-G41-1 i

l DESCRIPTION OF CHANGE: This Temporary Change Notice (TCN) provides the necessary procedural controls to  ;

permit alignment of one fuel pool cooling and cleanup (FPCC) train consisting of one FPCC pump and both FPCC heat exchangers to one train of standby service water i (SSW) following a single failure of one FPCC train.

REASON FOR CHANGE: This "one pump /two heat exchanger" l configuration of the FPCC system is required to  !

eliminate reliance on the spent fuel pool cooling assist mode of the Residual Heat Removal (RHR) system for loss of a FPCC pump and a SSW pump.

t

SAFETY EVALUATION

Operation in the "one pump /two heat ,

exchanger" configuration of the FPCC system does not  ;

prevent the SSW or FPCC systems from performing their ,

design heat removal functions consistent with the  !

assumptions of the UFSAR accident analyses since this  !

system configuration is dependent upon the initiating single failure of the operating FPCC train. Since an .[ '

additional single failure or single operator error is not required to be postulated, the FPCC and SSW systems ,

are no more likely to fail when required to function  ;

than before. Since the heat removal capabilities of the FPCC and SSW systems will not be affected by the "one pump /two heat exchanger" FPCC system )

, configuration, the change does not constitute a change i to the technical specifications or an unresolved safety issue.

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j Attachment to GNRO-93/00001 l

i SRASN: PLS-92-034 DOC NO.: TEMP ALT 92-0031 i DESCRIPTION OF CHANGE: Liquid seals are employed at l several locations on the offgas system piping. The l ~

purpose of the liquid seal is to drain the excess water from offgas process flow without allowing the offgas to escape through the drain lines. This is accomplished by use of loop seals. These seals are equipped with {

solenoid valves that close if release from the offgas  :

system exceeds established limits. This activity l evaluates installation of a temporary loop seal arrangement using tygon tubing for the offgas holdup line loop seal which has been determined to be clogged from the accumulation of debris.

REASON FOR CHANGE: This temporary loop seal will allow manual draining of the offgas holdup line volume until 1

the loop seal line can be opened and inspected (during _  ;

plant shutdown).  !

SAFETY EVALUATION: The installation of the temporary loop seal will not reduce the structural design requirements of the offgas system piping as the .

temporary loop seal tygon tubing will be isolated from ,

the offgas process flow by a level instrumentation isolation valve, when not manually draining the holdup line process volume. The isolation capability of the solenoid valve is not affected, should release from the offgas system exceed the established limits.

Disconnecting the level switch instrumentation will not  !

detrimentally affect loop seal operation of the loop l seal as it is presently clogged and will not allow l offgas process flow to escape. The offgas system is  !

, non-safety related and needed only for power l production. For the reasons stated above, this  !

activity does not affect the design objective of the  !

gaseous waste management system and will not l

,. detrimentally affect the maintenance of dose to offsite persons from routine station releases to significantly less than the specified limits and will not  !

detrimentally affect plant operation within the  ;

emission rate limits established in the station j operating license. j i

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Attachment to GNRO-93/00001.

SRASN: PLS-92-035 DOC NO.: TEMP ALT 92/0033  ;

1 I

DESCRIPTION OF CHANGE

This modification removes the i electronics from the drywell equipment drain sump high- l

, high level switch (1P45-N221). Also, a jumper is.added from the high-high level switch to the high level j switch of the drywell-equipment drain sump.

REASON FOR CHANGE: The drywell equipment drain sump l high-high alarm switch failed in alarm condition and i cannot be repaired until the plant shuts down. ,

Removing the electronics will clear the locked in alarm condition. The installation of the jumper restores the l second pump start function and the annunciator which is removed by the removal of the electronics from the switch. j i

SAFETY EVALUATION: The Drywell Equipment Drain Sump l System is non-safety related and performs no function ,

i important to safety. Neither the removal of the switch  !

which is used to start the second drain sump pump and l energize an annunciator on HI-HI drywell equipment l drain sump level or the installation of the jumper  !

could initiate any action or event that would increase {

the probability of an accident as previously evaluated j in the SAR since either a primary or secondary plant 4

isolation will isolate the discharge path for this  ;

system and render this system inoperable.

The function of this switch is to start the second pump f and energize an annunciator on HI-HI drywell equipment  :

drain sump level. None of the plant conditions that  !

were used for evaluating an accident will be affected [

by the removal of this switch or the installation of l this jumper. The radiological consequences of this l temporary modification to the non-safety related Drywell Equipment Drain Sump System, which is not used l in mitigating the radiological consequences of an  ;

accident, cannot therefore increase the consequences of '

an accident previously evaluated in the SAR.

The removal of the switch and installation of the i jumper will affect only the non-safety related pump l 4

start circuits and annunciator and will not affect the i limits of any systems as described in the SAR; j therefore, it will not reduce the margin of safety as l defined in the basis for any Technical Specification.  ;

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Attachment to GNRO-93/00001 {

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SRASN: PLS-92-036 DOC NO.: LDCR Change i No. PLS-92-007 l

l DESCRIPTION OF CHANGE: Storing defective fuel bundles  !

in the spent fuel pool " fuel racks" is acceptable. t Defective fuel containers are only needed for storage  !

of severely damaged fuel bundles.

i REASON FOR CHANGE: This change is needed for storage j

, of defective fuel bundles without defective fuel -

, storage containers. l I

i SAFETY EVALUATION: The defective fuel is placed into the spent fuel pool (SFP) storage racks in the same le s

manner as used for non-defective fuel. There are no changes to fuel handling activities or sequences of actions. Therefore, the initiating events that result i in a fuel handling accident in the Auxiliary Building or a fuel handling accident inside containment are i 3 unchanged.

The reactivity of a fuel assembly is not affected by

, small cladding perforations that do not threaten the structural integrity of the fuel bundle. The reactivity is unchanged as long as fuel fragments are not being released from the assembly. Therefore, there ,

is no impact on the criticality analyses performed for  !

storage of fuel in the SFP. i i

1 The dynamic response of a fuel assembly during a  !

seismic event is primarily affected by the presence or  :

absence of a fuel channel. Any small cladding damage l would not affect the channel and hence the seismic  :

response of the assembly. Defective fuel will be  ;

stored in the spent fuel pool with a channel. Any  ;

assemblies with gross cladding damage will still be  !

. stored in the defective fuel storage container.  :

Therefore the affects of a seismic event on the spent  !

fuel pool are unchanged. j Since there are no affects to the fuel handling f accidents, criticality of the spent fuel pool, and seismic response, the storage of defective fuel in the l

. SFP racks without a defective fuel storage container '

i will not increase the probability of occurrence of an accident previously evaluated in the UFSAR. I l

The consequence of the fuel handling accident is the  !

release of radioactivity. The release from a bundle j l drop associated with defective fuel is the same as a ,

bundle drop associated with non-defective fuel since  !

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i Attachment to GNRO-93/00001 {

l t

PLS-92-036 '

Page 2 i l

. j all rods in the dropped assembly are assumed to fail. l The release from a bundle drop onto defective fuel is i unchanged or reduced since most of the fission gas in i the defective rod has been previously released. I l

The criticality analysis for the spent fuel pool is not [

affected by the storage of defective fuel as long as i the assembly geometry is maintained. Fuel assemblies i that are damaged to the extent that their geometry cannot be guaranteed will be stored in a defective fuel i i

storage container.

Integrity of the boraflex sheets in the spent fuel pool j is not compromised by the storage of defective fuel. F Gamma flux level is the only factor affecting boraflex

! gap growth. Gamma flux levels will not be affected by minor cladding damage. Introduction of defective fuel  ;

. into the spent fuel racks will not create a new type of  ;

malfunction.

The storage of defective fuel in the spent fuel racks will not require any changes to normal fuel assembly  !

handling nor will there be any changes to crane  :

operations in the Auxiliary Building. Since gamma flux [

levels are not increased in the spent fuel racks there l is no impact on the fuel rack criticality analyais. .As i a result, the margin of safety as defined in the basis i for Technical Specifications, " Refueling Operations" {

and " Fuel Storage" will not be reduced. j i

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Attachment to GNRO-93/00001 i t

SRASN: PLS-92-037 DOC NO.: EER-89/6229  !

t DESCRIPTION OF CHANGE: Increase Area Radiation Monitor (ARM) K605 setpoint from 15 mr/hr to 85Hmr/hr. This is

above the UFSAR radiation zone value of 515 mr/hr and i

, . is above the initial UFSAR setpoint of 15 mr/hr.

REASON FOR CHANGE: ARM K605, 119' containment 1

Transversing In-Core Probe (TIP) area, is in continuous alarm. The general area readings are 20-75 mr/hr. Rad levels in the area have increased due to plant [

operation and are influenced by use and resulting activity buildup in the TIP mechanisms. The current '

alarm setpoint is 15 mr/hr. The proposed change '

increases the setpoint to 85 mr/hr. The proposed change restores the ARM function, which is to warn.

personnel of increasing or abnormally high rad levels.

l SAFETY EVALUATION: This ARM has no active emergency '

shutdown features. The system is not essential for safe shutdown of the plant and it serves no active emergency function during operation. Consequences of ,

~

an accident previously evaluated in the SAR are not 4 increased because this ARM neither mitigates nor monitors radioactive releases to the environment. ,

i The proposed change does not affect system function o.. e operability, the system will continue to function as ,

intended. The system is not essential for safe shutdown, nor does it serve an active emergencv ~

shutdown function.

The ARM system serves no radiological effluent monitoring function, therefore the margin of safety associated with radiological consequences of accidents is not reduced. This ARM is not the basis for any GGNS ,

Technical Specification. i i

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Attachment to GNRO-93/00001- 'i SRASN: PLS-92-038 DOC NO.: MWO 82439 '

DESCRIPTION OF CHANGE: Siemens Power Corporation (SPC) will be performing fuel bundle inspections on two SPC 8x8 assemblies (XNC-827 and XNC-582) as a followup to the 8x8 fuel bundle compression spring issue that ,

surfaced in RF05.

t After the inspection the bundles will be placed in storage in the spent fuel pool. They will not be ,

reloaded into the reactor core.

REASON FOR CHANGE: The SPC 8x8 reload fuel was found during RF05 to have unexpected differential rod growth. l To determine the cause of this behavior, SPC will '

conduct a detailed fuel inspection of two selected discharged 8x8 bundles.

SAFETY EVALUATION: The process for removing the .

I adjusting nuts will ensure that at all times the bundle 1

can be re-assembled to a structurally sound condition which is able to withstand expected handling and seismic loads. The required drilling on the tie rods l will be performed adequately above the rod plenum that the interior integrity of the rods will not be breached. 4 The process for re-installing the upper tie plate, either with the use of special flanged nuts and/or inert tie rods will result in a bundle which is still capable of withstanding the expected handling and 4 seismic loads. No special equipment or procedures will be needed for handling the bundles after re-assembly. I The ability of the tie rods to withstand the i compressive loading during a fuel handling accident is '

unchanged. The reactivity, decay heat, and fission product inventory of the bundle will not increase. j Should it be decided to remove any identified failed i rods from a bundle prior to re-installing the upper tie  ;

plat, it is acceptable to leave those locations vacant.

, Any vacancies will not increase the reactivity of the  !

bundle. The procedure will ensure that any vacant tie rod locations within the bundle will not result in a ,

bundle which cannot withstand the expected handling or seismic loads.

[188]

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Attachment to GMRO-93/00001 l l

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PLS-92-038 j Page 2 1 1

l Since the two bundles will not be reloaded into the core, the thermal-hydraulic /neutronic transient considerations of operating with either inert fuel rods or vacant rod locations are not applicable. Similarly, the hydraulic forces acting on the upper tie plate  !

during operations is not a concern. l The fuel bundle inspection will not increane accident or malfunction probabilities or consequences. It will ,

not create any risk of a different type accident or ,

malfunction and does not reduce the margin of safety in -

the Technical Specifications bases. There are no unreviewed safety questions.

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l Attachment to GNRO-93/00001 SRASN: PLS-92-039 DOC NO.: 01-S-06-5 Rev. 22 DESCRIPTION OF CHANGE: This change removes Incident Reports (IR) from the corrective action process, allows the use of new definition of significant, and requires  !

that other programs be used to resolve irs. ,

REASON FOR CHANGE: To streamline current process for irs and to focus attention to those that are significant.

SAFETY EVALUATION: These are primarily process changes to the administrative functions associated with handling incident reports. There will be no reduction in overall effectiveness of the program and should enhance the process by freeing management time and focusing more attention to significant items.

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[190]

Attachment to GNRO-93/00001 I l

l SRASN: PLS-92-040 DOC NO.: Site Directive G4.110 I Rev. 3

^

DESCRIPTION OF CHANGE: A " pre-screening" has been  !

added to the applicability review form. If any of the i pre-screening questions are answered in the affirmative, then neither a safety evaluation applicability review nor a 10CFR50.59 safety evaluation is necessary. Also, documentation of the basis for screening conclusions has been made optional. A check-  :

off list for the revir,wer of an applicability review t has been added to record how the reviewer performed his review that reached the same conclusions as the preparer.

REASON FOR CHANGE: This change allows exclusion of ,

unnecessary applicability reviews and safety i evaluations when one of the pre-screening criteria is '

fulfilled. This change adds flexibility to documenting ,

, the basis of applicability review conclusions.

SAFETY EVALUATION: This change does not revise the method of performing an applicability review. A change i to the Technical Specifications or to information  !

described in the FSAR will continue to be identified j and safety evaluations will be performed as required. l This change provides a method for determining that an ,

1 applicability review or safety evaluation is not  ;

necessary, i e., a pre-screening. An applicability j review or safety evaluation is not necessary if the  ;

change is editorial, 10CFR50.54 applies to the change i instead of 10CFR50.59, an approved safety evaluation on j this subject already exists, the change already has j

been approved by the NRC or the change is an FSAR l change that meets the exclusion criteria outlined in l Site Directive G4.803. l l L

] This change provides options for documenting the basis  :

for applicability review conclusions. In lieu of l

reviewing the basis documentation, the independent  !

reviewer will complete an independent applicability

• review or perform a verbal review with the preparer.

Because these alternatives provide for at least an ,

i equivalent level of independent review quality as a  !

documentation review, the changes to the applicability  !

review process do not reduce the effectiveness of the  ;

safety evaluation process. The applicability review process will continue to ensure that a safety l evaluation is performed when required. i 1

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1 Attachment to GNRO-93/00001 l l

i SRASN: PLS-92-041 DOC NO.: FSAR CR PLS-92-009 )

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I DESCRIPTION OF CHANGE: Change Fire Detection System l Requirement 3.3.7.9 Action (a) to require establishing a fire watch patrol only in areas affected by fire ,

detector inoperability. i REASON FOR CHANGE: Allow reliance on operable fire ,

detection instrumentation for early warning when >

portions of the Fire Detection System are impaired or otherwise inoperable, f SAFETY EVALUATION: This change is to provide more flexibility in applying administrative controls of SAR Section 9B.6.n and in meeting Technical Requirement t Manual (TM4) action requirements at times when  ;

individual fire detectors are impaired or are otherwise l inoperable such that unaffected sensors in the same zone remain capable of performing their design function i

as described in the UFSAR.

Fire zones envelope multiple defined areas or rooms. l The current wording for Appendix 16A, Requirement l 3.3.7.9, Action (a) requires an hourly firewatch be established for all areas in a fire detection zone when ,

less than the minimum number of fire detection instruments in that zone are operable, even if some j areas in that zone contain no inoperable detectors, and i the mode of the detector inoperability would not mask l or impede the operation-of the detectors in the  ;

unaffected areas. This change will allow credit to be taken for fire detection instrumentation in areas / rooms  !

where all required detectors are operable, while other l portions of the defined zone containing inoperable  !

detectors are assigned a firewatch patrol per l Requirement 3.3.7.9 Action (a), as applicable, without ,

compromising the intent of fire protection program i requirements. {

t This change will ensure firewatch patrols are still j mandated for areas in which required fire detection  :

instrumentation will not function as designed while allowing reliance on unaffected instrumentation for ,

fire detection and early warning.

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1 Attachment to GNRO-93/00001 l

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SRASN: PLS-92-042 DOC NO.: 04-1-03-N19-1 l l

- I i DESCRIPTION OF CHANGE: Individually remove from l service minimum flow valves for condensate, condensate  !

booster and reactor feed pumps to verify those valves l will operate properly during reactor plant shutdown.

REASON FOR CHANGE: To allow removal from service  :

minimum flow valves required by FSAR to be in service while plant is in operation. This will allow the valves to be stroked open and closed to ensure the i valves work properly when required by changing flow requirements during down power manuevers.

SAFETY EVALUATION: The condensate and feedwater t systems are not required to affect or support the safe  ;

shutdown of the reactor or perform in the operation of  !

the reactor safety features. The isolation of the minimum flow valves to test their function for a '

limited duration will have no effect on the operation of the condensate or feedwater systems since these valves are normally closed in steady-state power

operation. The probability of a loss of feedwater f event occurring during the testing of these valves for i a limited duration is not discernably greater than the  !

probability of a valve failure during a transient.which  !

leads to loss of feedwater event. This conclusion is l based on the high failure history of these valves  ;

during startups and shutdowns at GGNS. Therefore,

there is no increase in the probability of an accident.

In addition, this procedure helps ensure proper minimum flow valve operation prior to condition when valve  !

3 operation is needed.

i Isolation of the minimum flow valves in the condensate 2

and feedwater systems do not increase the consequences of an accident in the SAR. Loss of feedwater flow to  !

the vessel is the accident associated with failure of j the condensate and feedwater systems minimum flow {

1 protection to function properly. This accident is i analyzed in the SAR as having no radiological  !

consequences. Therefore, there is no increase in the  ;

consequences of an accident in the SAR. t The condensate and feedwater systems will continue to operate as normal providing no transient occurs which  ;

effects feedwater flow. If a transient occurred during i the isolation of the minimum flow valves, the condensate and feedpumps will trip on low suction l 1

pressure. Therefore, the equipment will perform as  :

required for protection of the pumps. l t

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Attachment to GNRO-93/00001 PLS-92-042 Page 2 Any transient that occurred associated while testing the minimum flow valves would have the same result as a failure of the minimum flow valves which is bounded by a loss of feedwater flow analyzed in Chapter 15 and the reload analyses. No other accident is expected to occur other than evaluated in the SAR.

Isolation of the minimum flow valves in condensate /

feedwater does not affect any margin of safety as defined in bases of any Technical Specification. Also in the event a loss of feedwater flow occurs, the reactor protection system initiates an automatic scram placing the plant in a safe condition due to low water level.

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3 Attachment to GMRO-93/00001 i

SRASN: PLS-92-043 DOC NO.: WO #84785  !

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DESCRIPTION OF CHANGE: The wiring modification to be performed will remove all primary water low flow trips to the generator. j REASON FOR CHANGE: A failure of a cord in the primary l' 1

water system section of the Electronic Generator Protection (EGP) System has made it necessary to bypass all generator trips associated with the primary system ,

low flow.

SAFETY EVALUATION: The main accident of concern when dealing with the turbine / generator is the generator load reject with failure of bypass flow. This event is  ;

categorized as an incident of moderate frequency. The i primary water system generator trips are used to protect che non-safety related generator from equipment i damage. With the primary water system low flow '

generator trips bypassed, operators will be relied on to monitor primary water system flow parameters  !

continuously. With the primary water generator trips bypassed, the system could not initiate any action or ,

event that would increase the probability of a  !

generator trip. With the primary water system generator trips bypassed, only the non-safety related .

, generator which is not used in mitigating the  !

radiological consequences of an accident is affected i and therefore would not increase the consequences of an  ;

accident.

The result of the bypassed primary water generator ,

trips would only effect the generator and no Technical  ;

Specification limits would be affected. The bypassing  !

I of primary water system generator trips will only '

affect the performance of the generator trip system and not affect any safety systems or limits as described in -

the SAR. Since the bypassed primary water system generator trips will not affect the limits of any i systems as described in the SAR then it will not reduce j the margin of safety as defined in the basis for any Technical Specification.

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Attachment to GNRO-93/00001 SRASN: PLS-92-044 DOC NO.: Transfer of Environmental Surveillance Program DESCRIPTION OF CHANGE: Transfers the Environmental Surveillance Program (ESP) from Vice President, Operations Support, to Vice President, Operations GGNS.

More specifically, the ESP transfers from Corporate (Manager, Environmental Services) to Plant Operations (Chemistry Superintendent).

This change also deletes the second sentence from UFSAR Section 13.1.1.1.1.6.4 because the Final Environmental Report is not required to be updated or revised and revises UFSAR Section 13.5 to reflect incorporation of environmental procedures into Volume 8 of the GGNS Operations Manual.

REASON FOR CHANGE: A Quality Action Team studied the organization of Environmental Surveillance Programs at Entergy Operations' three nuclear sites and recommended this transfer of responsibility.

SAFETY EVALUATION: This change..will transfer the Environmental Surveillance Program responsibilities and staff from Corporate to GGNS. As a result of this change, Nuclear Support Administrative Procedures for the Environmental Surveillance Program will become Plaat Administrative Procedures. Changes to UFSAR Section 13.5 show incorporation of environmental procedures into volume 8 of the GGNS Operations Manual.

This change also deletes the second sentence of UFSAR Section 13.1.1.1.1.6.4. A requirement for maintenance, distribution and revision of the GGNS Final Environmental Report (FER) does not exist. Since the NRC has accepted the FER as complete and published the Final Environmental Statement (NUREG-0777), revisions to the FER are not necessary.

This change will not modify requirements associated with GGNS's Environmental Surveillance Program. Since environmental surveillance will continue as before the transfer, there will be no increase in consequence or probability of an accident or malfunction.

Furthermore, environmental surveillance activities occur outside and independent of plant operation.

Therefore, no adverse affect is possible.

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Attachment to GNRO-93/00001 i

SRASN: PLS-92-045 DOC NO.: MAEC-82/0093 DESCRIPTION OF CHANGE: The commitments made by -

Mr. C. K. McCoy to Mr. F. S. Cantrell of the Region II NRC office in a phone conversation on April 5, ~ 1982 as  !

documented in MAEC-82/0009, Section 8, Pages 5 & 6,  ;

I Paragraph 7, Parts a, b, & c are rescinded. These '

commitments were based on the lack of ANSI 18.1 '

Qualified Health Physics Technicians on plant staff at that time. Specifically, GGNS committed to perform and review surveys and radiation work permits (RWPs) in a manner that required the direct control of the few ANSI 18.1 qualified. individuals on plant staff who also had some commercial nuclear power experience. q REASON FOR CHANGE: Deletion of the commitments will  !

enable the Health Physics Section to utilize their [

experienced personnel in the most effective manner.

SAFETY EVALUATION: The majority of Health Physics f personnel on staff at this time are more experienced  !

than the Health Physics Supervisor was in 1982.  !

Program _ commitments to the FSAR, regulatory guides, and  !

INPO training programs provide assurance of acceptable  !

quality in these areas of the HP program. Requirements i listed in Technical Specification Sections 6.2, 6.3, t 6.4, 6.8, and 6.11; UFSAR Sections 12.1, 12.5, and 13.1; and the applicable regulatory guides will not be  :

changed by rescinding the commitments listed in Section 8 of MAEC-82/0093. This change will have no adverse affect on safety. 1 i

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Attachment to GNRO-93/00001 SRASN: PLS-92-046 DOC NO.: UFSAR CR NL-92-005 DESCRIPTION OF CHANGE: The positions of Superintendent, Operations Training and Superintendent, Technical Training were removed from the GGNS Training Organization.

REASON FOR CHANGE: The positions were removed due to company-wide organizational restructuring.

SAFETY EVALUATION: The Manager, Nuclear. Training will ensure the responsibilities previously held by the Superintendent, Operations Training and Superintendent, Technical Training are transferred to and performed by the Manager, Nuclear Training or the Supervisors of the GGNS Staff Training Program. The training organization change constitutes a purely administrative change only to the UFSAR and will not have a direct affect on safety or the operation of the plant.

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Attachment to GNRO-93/00001 i

SRASN: PLS-92-047 DOC NO.: LCTS 15692 f (AECM-89/0074)

DESCRIPTION OF CHANGE: This change modifies the commitment (LCTS 15692) to change plant procedures concerning_ closure of containment isolation valves during a station blackout in the event of imminent core '

damage. Specifically, the change removes certain valves from the list of valves to be closed. The ,

change is made in order'to prevent conflict with emergency procedures and to ensure that operator actions are not diverted from activities that would be more appropriate. This evaluation evaluated 1) Limited Station Blackout (SBO) Analysis for Grand Gulf, 2) ,

Assumptions underlying the SBO Rule, 3) Beyond Design / License Basis effects that may be considered pertinent to this issue and evaluates the potential i effects of the proposed changes on.those analyses.

REASON FOR CHANGE: Because of the safety implications  ;

and because a sound technical basis for the change  ;

exists, the commitment made to close selected  :

containment isolation valves during an SBO event will {

be eliminated as allowed under 10CFR50.59. I

, SAFETY EVALUATION: The proposed change does not j propose to utilize any systems, structures or t components in a different function or role than is i currently accounted for by GGNS Technical Specification. .

Exclusion of the specified valves from the closure l commitment does not alter the basis for evaluation of accident probabilities for any accident previously evaluated in the SAR.  !

The proposed change is only relevant to an SBO event and affects only the portion of that event just prior to core damage.

SBO is not evaluated in the FSAR. ,

I The proposed changes are applicable only to an SBO event which has proceeded to the point of imminent core damage.

This is strictly a condition which could only be postulated to occur after the evaluated SBO required .

coping duration and does not introduce failure l possibilities of equipment important to safety. l l

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i' Attachment to GWRO-93/00001 PLS-92-047 i Page 2  !

I The consequences of the assumed core damage event and I

any associated equipment malfunctions will be mitigated  ;

by actions consistent with the requirements of BWROG j Emergency Procedure Guidelines, Rev. 4, Appendix B.

The changes do not modify any equipment or alter the function of any equipment important to safety.  ;

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The response does not propose to utilize any systems, t structures, or components in a different manner which  !

could impact the license basis and reduce a margin of {

safety. j i

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