Text

Rept of Changes,Tests & Experiments Determined to Be Reportable Under Requirements of 10CFR50.59 for Jan-May 1987
	ML20236V219
Person / Time
Site:	Grand Gulf
Issue date:	11/27/1987
From:	Kingsley O; SYSTEM ENERGY RESOURCES, INC.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	AECM-87-0196, AECM-87-196, NUDOCS 8712040163
	Download: ML20236V219 (58)
	v • d • e

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-i l EYETEM ENERGY - 1 REEDUNCEE, INC.' lj Oue D KitGaIV..JR v a nes e nt November 27, 1987. '

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U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk Gentlemen:

SUBJECT:

. Grand Gulf Nuclear Station ,I Unit 1 Docket No. 50-416' j License No. NPF-29 j Report of 10 CFR 50.59 Safety Evaluations - January 1, 1987 through May 31, 1987 1 AECM-87/0196 In accordance with the requirement's of 10 CFR 50.59(b),-System Energy i Resources, Inc. is reporting those changes, tests, and experiments determined ,

to be reportable under the requirements of 10 CFR 50.59 for the period of . '

January 1, 1987 through May 31, 1987. A summary of these changes, tests, and experiments is contained in Attachment I.

Attachment II contains a brief description of those evaluations. performed -

under 10 CFR 50.59 that support Revision 2 to the Updated Final Safety )

Analysis Report but which have not yet been-included in a 10 CFR 50.59 summary report. The evaluations listed in Attachment II are provided here as required by 10 CFR 50.71(e) and will be summarized in the next 10 CFR 50.59 report.

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AECM-87/0196 Page 2 cc: Mr. T. H. Cloninger (w/a)

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

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

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

Mr. R. C. Butcher (w/a)

Dr. J. Nelson Grace, Regional Administrator (w/a)

U. S. Nuclear Regulatory Commission Region II 101 Marietta St., N. W., Suite 2900 Atlanta, Georgia 30323 Mr. L. L. Kintner, Project Manager (w/a)

Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission ,

7920 Norfolk Avenue Bethesda, Maryland 20814 I

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l ATTACHMENT I SAFETY' EVALUATION SUMMARIES t.

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Attac'mentlitolAECM-87/0196 1 4

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' SRASN: .PLS-87-001 DOC NO: TSTI-1J11-86.-002-0-S: _ SYSTEM: _ VAR

.g DESCRIPT'IO'N OF TEST: . .A Technical Special i Test Instruction.(TSTI) provi_ded for the collection'of data for use in; calculations totredefine the detect

.and suppress. region for thermal hydraulic. stability of the core.

REASON FOR TESTi This test was to obtain dataito support core stability.

. calculations. ,

SAFETY EVALUATION: This test was conducted to. collect. data:while operating the; plant in conditions ~which are. bounded by current transient and LOCA analyses. ~This test. consisted of operating the' plant'in accordance with existing license conditions,LTechnical' Specifications andi l operating directives. The equipment'used in this test was optically isolated from the plant safety systems. The only e'quipment affected directly by performance of this testLis the'.C88.GETARS' equipment which'is

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not important to safety, . Other. equipment manipulations may be necessary ;

in.the operation of the plant.but are performed per:previously approved 1 operating instructions. 'There is no increase in the probability of '

i' occurrence or in the consequences'of an accident or malfunction of equipment important'to safety'previously evaluated in-the Safety Analysis Report.

1 There is no creation _ of a possibility for an accident or' malfunction of.a 1 different type than any evaluated previously in the' Safety l Analysis .

Report as this test was for data collectio'n~only.and was performed within' normal operating procedures designed to' keep them plant within analyzed-:

conditions.

The plant was operated in accordance w'ith current Technical-Specifications. Therefore, the margin ofl safety for them was;not reduced.

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.SRASN: .PLS-87-002. DOC: NO: WP/P-42 Rev. 0" SiSTEM: .N/A y

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-DESCRIPTION OF CHANGE: This procedure provides-for; control of. Unit 2 .,

haza rdous: mater.ial s. This procedure: included an evaluation of-chemicalu 9 hazards in accordance with three: criteria, i~ e.', toxicity,cexplosivel hazard, and missile hazard,.for.those chemicals stored'and maintained on Unit 2. - >

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' REASON FOR CHANGE: This procedure was initiated to ensure that hazardousi l materials. stored at Unit 2 do not affect the safe operationiof Unit'1.-

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, j SAFETY EVALUATION: ProcedureWP/P-42wasinitiatedito'reu$eth'e.

probability of Unit 2 activities impacting Unit 1. The potentially- '

-q hazardous chemicals listed in the-procedure were evaluated against.

acceptable toxicity limits and no' adverse effects-were shoWn,to exist in:

regard.to these chemicals. The propane,. fuel gas and MAPP gascstorage-was analyzed and shown to be stored further than the limiting distance' O, required to obtain:1 PSI .on any safety related structure: should an-explosion occur. An inspection of the bottle storage. facility revealedi j -

that the bottles consisted of. standard compressed gas stora'ge. containers-similar to nitrogen bottles' referencedLin UFSAR 3.511.2.2.'tThere is,no 1

increase in the probability of. occurrence'or in thel consequences-of an accident or malfunction of equipment important to safety previously:

evaluated 'in the Safety Analysis; Report. . There is:no creation:of'a6 possibility for an accident or' malfunction of a different.typefthan:any l

evaluated previously in the Safety Analysis Report. Thereiis no reductionM ;

in the margin of safety as defined in the basis for any Technical' :j Specification.

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SRASN: .PLS-87-003 DOC NO: ;C/R 87-001' SYSTEM: :

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DESCRIPTION OF. CHANGE: This' change provided'for administrative' changes

-in the UFSAR for delineation of duties' between the' Shift-Supervisor and.-

Shift Superintendent,and delegation of Control' Room command authority '

to the Shift Superirmdent. .d

' REASON FOR. CHANGE: This change clarified the Control Room. command-function and reinforces the concept'that the' Shift Superintendent:has' the overall responsibility for the Control. Room command function.

'I SAFETY EVALUATION: These' changes were administrate've in. nature and.were made for clarification' purposes in the UFSAR 'No.unreviewed safety' question was identified and no change to the3T echnical Specifications.was. ;

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i _SRASN: PLS-87-004. ' 00C- NCu .C/R 87-003' SYSTEM: M41 DESCRIPTION'0F CHANGE: This change allows. containment purge'toibe-isolated as a normal mode of' operation. This' change willfnormally have the 20" and.6" purge systems both isolated, with provisions tolrun-the 20"'

Jor 6" purge system on an as needed basis to' maintain! habitability,-

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-pressure control or for required surveillance.

REASON'FOR CHANGE: This change reflects the? current mode of operation.

A temporary test provided for the. isolated' operation of containment-ventilation. A change to the UFSAR is.beingLmade to allowLnormally; isolated operation.

. I SAFETY EVALUATION: This change providesLa. conservative mode.of operation with. containment purge . isolated. -The vent dampers would be'closediat Lthe start of an accident and thus cannot fail "open."'_No new failure modes were created and the margin of safety was not reduced. NoJunreviewed

-safety questions were identified and no Technical Specification' change was required.

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l Attachment I to AECM-87'/0196 SRASN: PLS-87-015 DOC NO: Temp. Alt. 87-0008 SYSTEM: R61 DESCRIPTION OF CHANGE: This Temporary Alteration provided for the use of 1 the existing central communication system through the use of plant microwave to MP&L communications network.

REASON FOR CHANGE: This alteration provides a parallel path for paging j plant personnel- . The original paging system remains intact. ;

SAFETY EVALUATION: There is no increase in the' probability of occurrence or in the consequences of an accident or malfunction of equipment.

important to safety previously evaluated in the UFSAR. This Temp.. Alt.

does not affect any plant safety system. The communications system'is not connected to any safety system and this Temp. Alt. does not connect it to any safety system. No RF frequency is transmitted by this Temp. l Alt.; therefore it does not induce any RF signals into any plant system.

This Temp. Alt. ' stays within the design of the equipment and will not cause it to fail. Therefore, there is no creation of a possibility for ;

an accident or malfunction of a different type than any evaluated i previously in the Safety Analysis Report. '

The radio communications system is not addressed in the Technical Specifications and is not used as the basis of any Technical Specification. This Temp. Alt, will have no effect on the margin of safety as defined in the basis for any Technical Specification.

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J l e Attachment I'to AECM-87/0196-SRASN: NSP-87-001 D0C.NO: NPDP.3.1- '

SYSTEMi_.N/AL

. DESCRIPTION OF CHANGE: . Nuclear Production Department Proce'ure d 3.1 was1 revised to reflect a change in division of responsibility. :Specifically, Paragraph 4.1.3 transfers responsibility for transient and accident' analyses from the Manager, Nuclear. Services and. Fuels to the Principal Engineer, Operational Analysis Section-(OAS):in the Nuclear Plant Engineering (NPE) department; .this revised.UFSAR-Subsection 13.1.1.1.1.6.2.1.2.

REASON FOR CHANGE: This was.an administrative change to reflect a: chang'e; in organizational responsibility.

SAFETY EVALUATION: There was no increase:.in the probability'of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in.'the Safety Analysis. Report.

There was no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. There was no reduction in the margin of safety as defined in the basis for any Technical Specification. ,There were.no system or structural changes involved. The division of responsibili_ ties-changed, and the..

organization (NPE-0AS) assuming responsibility for transient'.and accident analyses has the manpower and experience to' perform this function.

NPE-0AS's knowledge in this area is comparable to that of Nuclear Fuels, which was responsible for transient and accident analyses.

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Attachment'I to AECM-87/0396 J

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SRASN: QA-87-001 DOC NO:.. Paps 01-5-02-2.& 01-S-02-3. . - SYSTEM: N/A ,

DESCRIPTION OF CHANGE: TheselQuality Programs procedure changes _(Plant. >

Administrative Procedures) allow section Supervisors or Superintendents' to:

approve Plant'.Section Procedures:-in 1ight of the 0perational Quality-Assurance Manual which currently states the-procedures'are to be appro'ved '

by the responsible: manager. '

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REASON FOR CHANGE: To_providethe'SupervisorsandSuperintendents'witN- !

procedure. approval' authority. 1 SAFETY EVALUATION: . No system or structural: changes-were involved. . The' '

I manner of controlling activities which could; affect systems remains- J equivalent'. Nn Technical Specification changeJwas: required for:this' change and no unreviewed safety question was identified.

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SRASN: QA-87-002 DOC NO: QAP: 5'10 Rev 15 ESY' STEM: N/A 1 i

R DESCRIPTION OF CHANGE: This QA' procedure change' allowed the Managers .

Program QA.and the Manager, Audits QA approval authority to issue Quality ,

Assurance Instructions.in the Quality' Assurance Inspection Manual.

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g REASON FOR CHANGE: This change reflected administrativeLchanges'in the-QA, Procedure..(QAP'5.10) to allow additional approva1 Land; issue-authority-

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for_the QA Inspection. Manual, i SAFETY EVALUATION: jThere is no increase,in:the probability'of-occurrence or in the consequenc'es of an' accident or malfunction of equipment.

important to safety previously evaluated in the Safety Analysis' Report.

There is no system or structural changes involved. Only division;of- .

responsibilities change, The QA mant.gers assuming responsibility for 4 2

approval and issue of Quality Assurance Instructions have the experience {

and background to perform this NnctionL Therefore,s there is no creation of a possibility for'an accident or malfunction of a different type than' .s any evaluated previously in the Safety-Analysis Report, . Also,.there is 'l no reduction in the margin of safety as defined in the' basis.for'any i Technical Specification. ' '-

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' Attachment.T to AECM-87/0196 SRASN: NPE-87-006 lDOCNO:. CN-251 SYSTEMi 251, DESCRIPTION :0F' CHANGE: This change rerouted the Control Room chloriner >

detector sample discharge line back into the supply ductf in order;to effectively obtain' chlorine / air samples.

To create'a zero pressure drop reference pointacrosst-REASON FOR' CHANGE:

.the. detector's ' blower allowing an' adequate sample to be drawn. <

,q, SAFETY EVALUATION: LThis change did not. result in any operational-for- 4 functional changes to the system. This change does not71ntroducelany'new . 1 failure modes or. increase the probability of an accident. No decrease in-the margin of safety was experienced. 'This change provided'for the l chlorine analyzer to operate as designed to ensure the capability' of ;

detecting chlorine in the concentrations required by the-Te'chnical I Specifications. No Technical Specification ~ change was. required,for th.is '

change and~no.unreviewed safety question.was identified.

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Attachment-1 to AECM-87/0196 ,

SRASN: NPE-87-007' DOC NO: . DCP 83-0554,- Rev. .' 0 SYSTEM: G17 .-

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DESCRIPTION OF CHANGE: : This design: change replaced the:3/8'? tubing. fromi i

valve NSP21-F362 to.the NSD17-J007 flush inlet valve;with on'e inch (1")'

stainless steel. pipe and1 installed a oneti nch (1") stainless steel. drain:

line, with isolation valve, connecting the. sample chamber' discharge line to existing' drain line l'!-HCD-325.

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REASON FOR CHANGE: .To increase,the flush flow rate"of demineralized-water through the NSD17-J007-radiation. monitor ~ sample manifold for liquidL 1 radwaste. The, previous' rate-was not adequate to flush the monitor sample.

chamber following a high radiation; level: detection.

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SAFETY EVALUATION: .This equipment is nonsafety and does not perform'anyL function.which mitigates the effects of any accident analyzed in the FSAR' and'therefore this change does not increase theLeonsequences of.any accident. This change does not: increase the probability.of an accident, This change does not decrease the margin of safety. This change does not create any new interface, or new failure mode which would affect any .

equipment, components or systems which are safety related or.important to safety. No Technical Specification change was required for this change and no unreviewed safety question was identified.

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' Attachment I tolAECM-87/0196

'SRASN: NPE-87-019 DOC NO: CN-P47-30' SYSTEM: P,47 V . . .

DESCRIPTION OF CHANGE: .This change provided for the installation.of a loinch sample tap on the Plant Service Water (PSW) system prior to entering the plant. '

REASON FOR CHANGE: .This change was.made.to provide a chemical-sampling.

connection on the PSW system' from.the radial wells priorJto'its Lentering the. plant.

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SAFETY EVALUATION: This' system has no safety function and is non-se'ismic. '

The piping meets all' ANSI.B31.1 code requirementsLand will function.in<its.

. intended manner. Therefore, there is no' increase.in the probability of?,

occurrence or in the consequences of.an accident'or malfunction of .

equipment important to safety previously evaluated in.the UFSAR. This1 change- also did not' create _the possibility. of an accident' or malfunction of equipment of a different' type than any evaluated previously<in.the UFSAR. There was no reduction-in the. margin'of safetyfas'defin'ed in'the-

. basis for the Technical Specifications. , .

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SRASN: NPE-87-023 00C h0: Ch-R61-029- ' SYSTEM: R61? l

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,- DESCRIPTION OF CHANGE: :This change provided for the: temporary. . .l"- :!

installation of.a public address desktop-station.and handset inside and- , 1 outside, respectively, off the'drywel1~ personne1' airlock. ,1 i

REASON FOR CHANGE: This change provided personnel'.with~ plant communication. capabilities at this location. , .

1 SAFETY EVALVATION: .This' change is to aid communication of personne1' working-in the drywell personnel airlock. LThere is no increase in the probability of" occurrence'or in the^ consequences of.an accident lor' .

j a malfunction of equipment important to: safety previously evaluated in.the safety analysis' report'. Nor.istthere the creation of a possibility for, an accident or. malfunction of: equipment.of a different. type than any- .

q evaluated p~reviously.in thel Safety' Analysis Report. There.is:no reduction j in theLmargin of- safety-as ' defined in the basis for any Technical- t i Specifications.

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, A$techinent"I 9,o AECM-8Y/0196 SRASN: NPE-87-040' DOC NO: ~ Misc - A/C'in3Centrol Building- .SYSTEMi .Z17 DESCRIPTION OF' CHANGE: Three self-contained air conditioning units inLthe J Unit 1 Computer and Control Panel Room' were. temporarilycinstalled.
REASON FOR CHANGE: -The three air conditioning units were-installed to! "
provide additional cooling tothe Unit L1 Computer and' Control Panel Room, within the GGNS Control. Building. '*
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SAFETY EVALUATION: .There,is no' increase in the probability,ofEoccurrence ,
or in the consequences of an accident or malfunction of equipment-important to safety previously evaluated ini.the. Safety Analysis Report.
The three self contained A/C units'are temporarily inttalled in.a non-i safety.related application. The A/C units and the'ir-_ associated piping and_.
pipe supports were evaluated and:found_to conform to.,all. applicable codes; and standards. The A/C units are powered from non-safety.related power; supplies. ' The three A/C units were reviewed:against. potential. hazards.~ .
required to be considered in'designiincluding fire, flooding, water let impingement, missile generation 'and' seismic:II/ILconcernsiand11t wast determined that no safe shutdown equipment'will be:affected nor would the failure of the A/C units in any way' affect systems or components hich' are required to mitigate.the consequences of an accident. . Therefore , thi s-change does not create the possibility fo'r-.an accident or malfunction'of a.
different type than any evaluated previously in.the Safety' Analysis Report.
The addition of the three A/C units in the' Computer Room and Control' Pan'el:
will not reduce the margin of safety as defined in the basis.for any ->
Technical Specification. The A/C units serve no safety related function and their failure will not compromise'the safety function of" safety. .' )
related systems or prevent a safe reactor shutdown.
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.Aitachment'I'toAECM-87/0196 SRASN: NFE-87-088~ DOC NO: NPE'FSAR'87-0004 SYSTEM: N/A'
. DESCRIPTION ~0F. CHANGE: This change permits theluse of1the criteria:of Code _ Case N-411LinLthe piping analysis'of new.and' existing' systems including' optimization of seismic restraints (snubbers) in existing systems. . Code, Case N-411= criteria.will only be used for, response spectra '
type . analyses,l including OBE, . SSE and hydrodynamic. loads.:
REASON:FOR CHANGE: This Code Case' permits.the use of-higher damping for; response spectra. analysis as an' alternative to the va'ues specif,ied in
' Regulatory Guide :1.6; for Class 1,. 2,' and 3 piping' .
SAFETY. EVALUATION! Code'Ca'e.N-411.is s in compliance'with ASME.Section III-requirements, has been' approved by the NRC and in Regulatory Guidef1.84 Revision 24. :If piping' supports'are moved,tmodified, orfeliminated, .
any increased piping l displacements'due-to the. greater. piping flexibility -
will be" checked to assure thatLthey can.be accommodated'and that therel will be no adverseLinteraction with adjacent-structures, components, or equipment. Therefore, no new failure modes were~ created or waslthe margin.
of safety reduced. No Technical Specification change was required and no' unreviewed safety question was identified.
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Attachment 2 to AECM-87/0196 1
SRASN: NPE-87-097 DOC NO: MNCR S7-0051 SYSTEM: N/A DESCRIPTION OF CHANGE: This change reclassified the section of the Standby Liquid Control (SLC) system piping between the outboard isolation check valve (Q1C41F006), up to, and including the' explosive actuated valves (Q1C41F004A and Q1C41F004B) from ASME Class I to Class-2.
REASON FOR CHANGE: The change in ASME Class 1 boundary to the' outboard check valves allows the portion of the piping outside the RCPB including the explosive valves F004A and F004B to be exempt from Section XI system '
pressure test following the opening and reclosing of a component in that portion of the piping system.
SAFETY EVALUATION: The outboard check valve (F006) is the second of the two normally closed valves and the reactor coolant pressure boundary (Class 1) can therefore be terminated at this valve in accordance with the criteria of 10CFR50.2.
The reclassification of ASME Class 1 to Class 2 boundary will-not affect the operation of the SLC system nor affect the design function of the F004A and F004B valves nor the reclassified piping. Valve qualification remains unaltered. Therefore, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. There is no creation of a possibility for an accident or l malfunction of equipment of a different type than any evaluated previously l in the Safety Analysis Report.
l Implementation of this change does not change the limiting conditions for operations, applicability, or surveillance requirements as defined in the l
basis for Technical Specifications.
All piping and valves meet all applicable codes and necessary qualifications. There is no reduction in the margin of safety nor were any new failure modes created. No unreviewed safety question was t
identified and no Technical Specification change was required.
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1 Attachment I to'AECM-87/0196 l SRASN: NPE-87-104 DOC NO: TDI.DR/QR 02-540A & C SYSTEM: P75 i l
e DESCRIPTION OF CHANGE: SERI performed an alternate calculation to that j recommended by the TDI Owners Group for the lube oil piping connecting the- ,
j sump tank and engine mounted lube oil pump for the'Div. I and II standby: 1 diesel generators. This evaluation included the effects of engine thermal expansion. l I
REASON FOR CHANGE: The TDI Owners' Group DWQR item 02-540A & C- i recommended that engine, tank and pipe displacements be measured in the j vicinity of the lube oil pump, and an evaluation be performed to determine if modifications or maintenance upgrades are required. However, SERI performed an alternate evaluation / calculation which included thermal expansion considerations showing that the maximum stresses in the lube oil piping and sump tank shell are within allowable valves and that the. engine thermal expansion is being relieved in an acceptable manner. This document is being reported as required by Operating License Condition 2.C.(25).
SAFETY EVALUATION: The evaluation concluded that the existing design is' i adequate and no modifications or actions are necessary, therefore the d operational and functional basis of the standby diesels are not affected.
As a result, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.
Based on an evaluation of the lube oil piping and sump tank, the existing .J design is adequate to perform its safety function. No modifications of j
l maintenance upgrades are required, therefore leaving the existing design condition as is will not create the possibility of a malfunction of-equipment important to safety different than previously evaluated in the i UFSAR nor is there a design modification to be performed that will create a new type of accident. Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification. ;
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Attachment I to AECM-87/0196 j l
SRASN: NPE-87-105 DOC NO: TDI DR/QR Report GG-119 -SYSTEM: P75. i i
l DESCRIPTION OF CHAN3E: This document provides an. evaluation of the j recommendation concerning replacement of integrated. circuits contained in: a the SERI response to TDI Owner Group Report GG-119. Specifically, this recommendation is to replace any integrated circuits, in a T0-5 package-with those cf the dual-in-line pin type. It is SERI's_ position that this.
recommendation is not required, and that the intent of the recommendation j]
is already met with the existing configuration.
REASON FOR CHANGE: This document evaluated TDI Owner Group Report GG-119 and determined that the intent of the subject report had already been met in regard to the integrated circuits with.TO-5 packages. This document is being reported in accordance with License Condition 2.C.(25).
l SAFETY EVALUATION: Existing GGNS arrangements for the. voltage regulator circuit boards comply with the' intent of the TDI recommendation. The-
! qualification established for these components is adequate to ensure i proper performance during events for which they are called upon to >
function. Therefore, no improvement in reliability of these. components would be expected to occur as a result of implementation of the TDI Owner Group recommendation, and it has been established that the current GGNS designs provide adequate reliability without modifications to voltage regulator circuit board integrated circuits. No change in function or performance characteristics would result from implementation of the recommendation, therefore accident probabilities and consequences will remain unchanged by not performing the prescribed recommendation.
Reliability for the subject components as established by qualification. l and the above analysis demonstrates that equipment malfunction '
probabilities and consequences are acceptable. Therefore, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. Not performing the recommended change does not create the possibility for an accident of malfunction of a j different type than any evaluated previously in the Safety Analysis j Report. Also, there is no reduction in the margin of safety as defined in ;
the basis for any Technical Specification. i i
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.SRASN: NPE-87-106' DOC NO: .NPEI-86-09264 SYSTEM: P75
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,3 DESCRIPTION OF CHANGE: This evaluation took' exception-in. implementing the- '
following changes in' regard to the TDI.0wners Group. recommendation on.thef .
Power Driven Potentiometer (PDPs) for the GGNS Standby Diesel. Generators:
1)' ' Replace the existing 500 ohm /25W resistor;w'it Oa 500 ohm /50W 6 resistor.
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2) The resistor should be mounted such that it'is.not directly below. <
,, I the' circuit board. '
3)' The bottom section of the PDP enclosure should be m'odifiedLt o provide
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slots'or. vents which support convection and air exchange.
l The first recommended change is based on calcula'tions.t'at h state' the. ,
500 ohm resistor will~be forced to. dissipate 26W.when'the DClpowerisystem l voltage is at its. maximum al.lowable voltage- (140V), .and. that this value?
exceeds the. rating -of' the existing r' sistor e (25W). (This . recommendation, 4 however, does not take into considerationiseveral factors which ~ negate the d necessity to replace the 25W resistor with one of a: higher rating. First, the PDP is an intermittently energized " Low Duty Cycle"Jdevice called'uponL j
to operate for several. seconds (usually less than~25 seconds,.with a- 'I 60 second full travel time) during surveillance land periodic testingof' the diesel generator. Also, the DC' power' system voltage does not.normally.
I attain the maximum value of 140V (usually floated at 132V) unless 'the. o battery chargers are set to the " equalize" position. /Thus, the only time- 1 that the resistor is forced to dissipate the stated 26W is when diesel. h surveillance and battery equalization occur concurrently,'and even-then. l the resistor sees this dissipation level for' only-brief. intervals as- -!
l discussed above. H
y The last two recommended changes are related to potential-thermal problems-with the PDP assembly. Tests conducted by NEW PEEBLES-ELECTRIC PRODUCTS '
for PDP qualification operated th& PDP' continuously for 8-hours.'at.a ,
temperature above 120 F with no failures. .These test.results, coupledt I with the fact that the control panels where the PDPs are, mounted'are ;
maintained below 120 F, and that the PDPs are only. used on an intermittent-I
' basis, allow a reasonable conclusion that there are no thermaltproblems with the PDP package. ~
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.1 REASON FOR CHANGE: It was determined that'no modifications were required {'
to assure adequate life expectancy and reliability of. the PDP package. l Therefore, the TDI Owners Group, recommendations for the PDP were not 1 implemented.. This_ document is being reported in accordance with; License i Condition 2.C.(25).
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..l J14 MISC 87102101 - 19 1
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Attachment 3 to AECM-87/0196-SRASN: NPE-87-106(cont) DOC NO: NPEI-86-0926 SYSTEM: P75 SAFETY EVALUATION: Tha intent of the TDI Diesel Generator DR/QR Report recommendation has been met as shown by the' analysis discussed under
" Description of Change." Therefore, there is no increase.in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the-Safety Analysis Report. The use of the existing design as justified above insures that probability of accident occurrences will.not be increased, as the previously tested and qualified PDP package demonstrates adequate reliability and service life, and this is. consistent with the present i UFSAR analyses, and the. intent'of the TDI Diesel Generator DR/QR. Report. !
Qualification testing as described in NPEI 86/0926 insures that malfunction probabilities of the PDP without performance of the described modifications are acceptable and remain within previously evaluated limits.
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There is no creation of a possibility for an accident or malfunction of equipment of a different type than evaluated previously in'the Safety i Analysis Report. Also there is no reduction in.the margin of. safety as '
defined in the basis for any Technical Specification.
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l J14 MISC 87102101 - 20
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Attachment 1 to AECM-87/0196 SRASN: NPE-87-107- DOC NO: TDI DR 02-380A SYSTEM: P75 DESCRIPTION OF CHANGE: This evaluation justifies the use of the existing.
GGNS Div. I and Div. II standby diesel generator design to that recommended by the TDI Owners Group DR/QR Report subsection 02-380A for removing one of the two 6" slip joints in each-D/G exhaust line and replacing each with a 150 lb. slip-on flange.
REASON FOR CHANGE: To take exception to TDI Owners Group DR/QR Report subsection 02-380A for the GGNS standby diesel generators. -This document is being reported in accordance with Operating License Condition 2.C.(25).
SAFETY EVALUATION: A supplement report No. 1, Rev. I and calculation prepared by Duke Power Company for GGNS Unit 1 evaluated the exhaust piping, considering the effects of the slip joints and sliding spur. The analysis reflected that the exhaust manifold piping is adequate to perform i
its function without modification. Since there is no design change, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously' evaluated in the Safety Analysis, Report. Nor is there the creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Also, there is no I reduction in the margin of safety as defined in the basis for any Technical . Specification.
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J14 MISC 87102101 - 21
Attachment I to AECM-87/0196 SRASN: NPE-87-105 DOC NO: TDI QR 02-371A & B SYSTEM: P75 1 DESCRIPTION OF CHANGE: This document evaluated TDI QR 02-371A & B which required GGNS to determine the hardness and material of the fuel pump control shaft for the GGNS Div. I and II standby diesel generatori. As documented in a letter from Duke Power Company, representing the TDI Owner's Group, dated August 29, 1986, to SERI, it was determined that QR 02-371A & B requirement was no longer applicable and is being removed from the GGNS DR/QR report.
REASON FOR CHANGE: SERI is taking exception to QR 02-371A & B in the j GGNS EDG DR/QR report as recommended by the TDI Owners Group. This document is being reported in accordance with GGNS Operating License ,
Condition 2.C.(25). 'I i
SAFETY EVALUATION: Deleting the 02-371A & B requirement to determine the fuel pump control shaft hardness and material does not affect diesel l
generator design or performance adversely. There is no increase in the probability of occurrence or in the consequences of an accident or ,
malfunction of equipment important to safety previously evaluated in the '
Safety Analysis Report. There is no creation of a possibility for an accident or malfunction of a different-type than any evaluated previously in the Safety Analysis Report. Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
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J14 MISC 87102101 - 22 1
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.o Attachment ILto'AECM-87/0196-
'SRASN: .NPE-87-109 DOC /NO: eTDI QR 02-360B' SYSTEM: P75 N
DESCRIPTION OF CHANGE: .ThisJdocumentLevaluated'TDI Owners' Group. item QR 02-3608/8 which required'GGNSito determine the material of,the valve and. valve' rings of the Div. I?and II: standby: diesel' generators. As documented.in a letter dated August 26,l1986 fromLDuke Power Company,
' representing the(TDI Owner's. Group, to SERI~ .it'was: determined that QR.
02-360B/8' requirement was no longer. applicable and

should be~ removed from.
.the GGNS DR/QR.
REASON.FOR CHANGE: ,SERIistakinge'xceptionLtoQR'02-360 Bin:theiGGNS-standby,D/G DR/QR report aszrecommended:by the TDI Owners Group.';This document is being' reported'in accordance with GGNS Operating License.
Condition 2;C.(25).
~ SAFETY. EVALUATION: Deleting.the 02-360Bl requirement;to' determine the' material of the valvef and valve; rings; does- not affect ' diesel: generatori design or performance adversely, :There is.no. increase'in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to . safety previously evaluated Lin the Safety Analysis Report. There is no creation:of-a possibi.lity foi an accident or ,
malfunction'of a different type than anyl evaluated lpreviously in the Safety Analysis Report. Also, there is.no reduction in the margin of safety as defined in the basis for'any' Technical Specification.
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Attachment'1.to1AECM-87/0196c SRASN: .NPE-57-110 'D'OC NO: TDI DR102-317A & B SYSTEM: P75 DESCRIPTION OF CHANGE: This. evaluation justifies not implementing the recommended TDI Owners' Group DR/QR subsection 02-317A & B which states' that 2-inch 90' degree dresser' elbows replace the ' elbow and. flange located on both water Jacket. lines between the 3-inch header'a'nd water. Jacket' shroud to mitigate the thermal expansion loading and stresses on the diesel engine.
\
REASON FOR' CHANGE: SERI analyzed the TDI recommendation. and ger.erated '
-Calculation MC-Q1P75-85022 Rev. O which concluded that the TDI recommendation intent had been met. This document is being reported in accordance with GGNS Operating License Condition.2.C.(25) -
SAFETY EVALUATION: The implementation of'02-317A & B DR/QR recommendation.-
as described above is unnecessary based- on MC-Q1P75-85022. Rev. O. This:
plant unique analysis for Division I.and II standby diesel gen'erator jacket water discharge piping has confirmed that-stresses are within-ASME
~Section III Subsection NC Code allowables-and-therefore are~ acceptable.
There is no increase in the probability of occurrence or in the-consequences of an accident or malfunction of equipment-important to=-
safety previously evaluated in the Safety Analysis Report. There is no creation of a possibility for.an accident or malfunction of equipment of a different type than any evaluated previously in the Safety Ar.alysis-Report. There is no reduction in the margin of safety as. defined.in the basis for any Technical Specification.
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J14 MISC 8/102101 - 24 l
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Attechment I to AECM-87/0196
SRASN

NPE-87-111 DOC NO: TDI DR-07-02-688A-0 ; G30A, B, C-0 SYSTEM: P75 l
DESCRIPTION OF CHANGE: This evaluation justifies not implementing the TDI Owners Group items DR-07-02-688A-0 & DR-07-02-630A, B, C-0 which recommended supports for flexible conduit every three feet between the
! junction boxes and the Div. I'and II standby diesel generators. Analyses l was performed using a conservative method to justify not supporting.the flexible conduit.
! REASON FOR CHANGE: SERI is taking exception to TDI Owners Group recommendation items DR-07-02-688A-0 and DR-07-02-630A, B, C-0 for implementation of GGNS for.the standby diesel generators. This document ;
l is being reported in accordance with Operating License Condition 2.C.(25). '
SAFETY EVALUATION: The flexible conduit and connections have been analyzed for dead load and loadings imposed by a Safe Shutdown Earthquake-(SSE) and found to remain structurally adequate. The analysis performed modeled the flex conduit as a catenary (conservative) resulting in higher loads than will actually occur. The flex conduit is routed such that the 1 flex is draped over the diesel support channels (beams) and confined horizontally such that the flex will not create a hazard to the diesel and associated components. The DR/QR report requirement for 3' 0" maximum spacing between supports parallels the National Electric Code (NEC). The 3' 0" spacing of the NEC is for neatness and to prevent the creation of a tripping hazard which is not a concern in this case due to the location of the flex conduits. The flexible conduit and the connections have been analyzed for all applicable loadings, including SSE and found to remain structurally adequate in the unsupported configuration. Therefore, the '
ability of the Diesel Generator to function as designed will not be affected by the unsupported flexible conduit. Elongation of the flexible conduit is less than 1% of the total length of the flex. The additional length of cable wrapped inside the junction boxes will prevent tension being placed on the connections by elongation of the flex. The actual tension loads on the flexible: conduits and junction box connections are considerably less than the allowable tension loads as documented in the referenced calculation. The omission of flexible conduit intermittent supports will not adversely affect the operation of any safety related equipment. Therefore, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report. _ There was no creation of a possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report. Also, there is no reduction in the margin of safety as defined in the basis for any Technical Specification.
J14 MISC 87102101 - 25
5g' Attachment I toLAECM-87/0196 SRASN: NPE-87-112 . DOC NO: .CFRMISC001R00' .SYSTEMi P75 DESCRIPTION OF' CHANGE': This. evaluation reviewed the deletion of-two pipe supports from the. scope of the modifications' implemented by DCPJ85-4037 on:the Div.:II' Diesel Generator starting air piping'and supports.
. REASON FOR CHANGE: To justify an exception to a GGNS modification being performed for'the TDI Owners Group item 02-441A regarding two supports which caused binding of the' fuel racks and were subsequentlyLremoved.
This document is being reported in .accordance'with Operating License Condition 2.C.(25).
SAFETY EVALUATION: .The operation and function of the; Standby Diesel '
~ Generator starting air system is not-affected by the deletion of these supports. Deleting:these' two supports'does not change ~the limiting conditions for operation applicability. or surveillance requirements. 'The starting air piping has been shown by . stress analysis to meet all applicable. codes.and design bases as stated in the UFSAR. Therefore,
.there is no incr' ease 'in the probability of occurrence or in the consequences of an accident or malfunction of equipment important.to safety previously evaluated 'in the Safety- Analysis Report. Nor is there any creation of a possibility for an accident or malfunction of-a different. type than any evaluatec previously in the Safety-Analysis-Report. Also, there is no. reduction in the margin lof safety as defined in the basis for any Technical Specification.
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1 J14 MISC 87102101 - 26 y i
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Attachment I to AECM-87/0196 .j l
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.SRASk NPE-87-113 DOC N0: TDI DR-GG-119 SYSTEM: P75 i
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DESCRIPTION OF CHANGE: This document evaluated TDI Owners' Group DR/QR -j
. Report, item DR-GG-119 which identified the field flashing relay as being i inadequately rated for' operation within a 90 to 140 VDC. range j (item 66-119).
1 REASON FOR CHANGE: The reason for not making any changes is-that this
~
situation had been previously' recognized for GGNS ~and appropriate testing I completed. Misleading paragraph. Batteries were not installed per contactor voltage requirements, but rather the Contactors were selected to match the batteries. This document is being reported in accordance with Operating License Condition 2.C.(25).
SAFETY EVALUATION: The Technical Specifications do not address the field flashing relay as a specific component. The reliability of the 125 iDC f station battery system which provides the basis for not requirir.g relay replacement is addressed in the Technical Specifications. The battery surveillance requirements and the diesel generator system design are unchanged.
The field flashing relay is a subcomponent of-the Division I and II Emergency Diesel Generators. There is no credible scenario where this relay is relevant to an FSAR evaluated accident precursor. The implementation of the action recommended in DR-GG-119 could increase the ;
probability of failure of the field flashing relay for the following j reasons:
a. No contactor coil suitable for operation at 90 VDC as well as 140 VDC is available off the shelf. A custom made coil that physically fits an NEMA-2 contactor would generate roughly 150 percent of the i operating temperature of a standard coil and its thermal life would be appropriately shorter. This is due to the fact that the losses i at 140 VDC are more than triple those at 90 volts. '
b. If a 64 VDC coil was employed in conjunction with a Zener-type
" voltage clamp" circuit, an additional component which can fail is ;
introduced.
The problem that initiated the generic fix (relay failures at 95 volts' DC or less) is not a problem at Grand Gulf. Even with the 125 VDC batteries at their limiting design point (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) the minimum voltage at the field !
flashing relay will be sufficient to insure diesel generator operation.
Therefore, there is no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment-important to safety previously evaluated in the Safety Analysis Report.
This action does not create the possibility of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report because component failure probabilities have not been increased.
Even if postulated, failures are enveloped by single failure criteria.
This component is not related to any UFSAR evaluated accident precursors.
J14 MISC 87102101 - 27
, o l Attachment f to AECMS 87/0196" 1'
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a L,
l SRASN: NPE-87-113 (cont)3 DOC NO: .'DR-GG-119 SYSTEM: P75 l
SAFETY EVALUATION (Cont'd): ,
. Page'B 3/4 8-2'of the Technical Specification bases deal'iwith'.
s. l surveillance of the battery systems, which have been shown.tofjustify the- '
1 adequacy of the. existing field flashing relay. .The Technical.
,1 Specification bases are acceptable and do not require revision. -No c_hange ,
is required to the. existing. installation.
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'J14 MISC 87102101 ,
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-'SR SNi NPE-87-116 f ' DOC'NO: ~DCP.85-583,RO,' 1, 2, 3, 4-501-R00:
LSYSTEM: - P47
. DESCRIPTION OF CHANGE: This. change' modified the.contfoi of th'-radial. e wells to' bypass the balance of' plant' computer cnd utilize'the microwave /
telemetry system to operate Lthe radial' well stop/ start" circuits.-
M REASON FOR CHANGE: This change gives the' Control lkoom direct' control of-the. radial we'lls without the. BOP computer. U
' SAFETY EVALVATIO': N iThePSWradialwell] system'has'nosafety-related' f 7 function. The radial well system provides. makeup _toithe.SSW' system.
. cooling ~ tower basins through the PSW system, but.this. capability;.is not -b required to safely shut down the;reactorsfollowing a LOCA. Therefore, there is no. increase lin the probability.of'occurrencefor in tne t consequences of an accident'or malfunction of equipment important tos safety previously evaluated in'the' Safety Analysis Report. This change" does not. create the possibility of an accident or. malfunction of a -
different type than any evaluated previously in-the'UFSAR. "
The'SP47 and SC91 systems are not addressed by the Technicali
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Specifications. Implementation of this. change will-have_no effect.on'the-Technical Specifications or on the operation.of the PSW' radial well system. There is no reduction in the margin'of safety as defin'ed in the basis for any Technical Specification.
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'J14 MISC 87102101 - 29 m.____.m.u.-. m. _.._m ____-_.- _-.-..-__._.____m__-_m.__m.__ . . . _ . _ _ _ _ _ ___m _- ____ _ .a._--..__m -:.__m..-u._____m____ _.._hau_.----
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Att$chmentItoLAECM-87/0196' i
SRASN: 'NPE-87-124 DOC NO: DCP-85-0068 SYSTEM: B33 I
L DESCRIPTION OF CHANGE: This Desigr. Change disabled lthe flow l controller automatic mode of operation of.the master controller.for'the' recirculation flow control system.
d
' REASON.FOR CHANGE: This change prevents operation in the Automatic Load: -l Following Mode which was. removed from startup. testing activities as-discussed in AECM-8UO131.
SAFETY EVALUATION: The automatic / manual mode of operation of the flux controller and the manualfmode of operation of the flow controller are !
still operational. This design' change has.no impact <on'the interface between the ApRM!s and the-flux. controller. This design change removed- -
the Automatic Load Following (ALF) mode of operation of the' recirculation flow control. The ALF mode of the< recirculation flow control system is :
an enhancement.'for operations during normal operation. The' flor 1 controller and master controller.are non-safety-related. There is no f increase in the probability of occurrenee or in.the consequences of an 1 accident or malfunction of equipment important' to safety previously evaluated in the Safety Analysis Report.
This change only allows for the manual control'of the recirculation flow ',
control, which is the mode:of operation.specified in the UFSAR should there be a failure to the recirculation. flow control (increasing or ]
decreasing flow). There is no creation of a, possibility for an accident 1 or malfunction of a different type:than any evaluated previously in the
. Safety' Analysis Report, q
The ALF mode of operation of the' recirculation flow system is 'an enhancement for operation of the recirculation flow system and the .
deletion of.this mode'will not reduce the mergin'of safety as defined in the basis for any Technical Specification.
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Attachment I to AECM-87/0196 SRASN: NPE-87-150 DOC N0: DCP 83-0538-R00 SYSTEM: GA1 DESCRIPTION OF CHANGE: This change provided for the temporary spacer in the restricting orifice (R0 01G41D010) to be. replaced with a permanent spacer in the Fuel Pool cooling and cleanup system. ' Figure 9.1-26 in the FSAR was revised to reflect this change.
REASON'FOR CHANGE: R0 Q1G410010 was improperly sized for the trimmed impellers installed on the fuel pool pumps (G41C001A and B). The flow is now controlled by motor operated throttle valves Q1G41F021 and.F043.
l Therefore, the temporary spacer in RO Q1G410010 was no longer required.
l SAFETY EVALUATION: Replacement of a temporary spacer in the restricting orifice (R0 01G41D010) with a permanent spacer meeting the requirements of ASME Section III, Class 3 piping did not affect the fue1~ pool cooling and cleanup system safety function of removing decay heat from the fuel assemblies, maintain pool water level, and removing radioactive fuel from l the pool. Therefore, no increase in the probability of occurrence or in t
the consequences of an accident or malfunction of equipment important to f
1 safety previously evaluated in the Safety Analysis leport was experienced.
This change did not create the possibility for an accident or malfunction of equipment of a different type than any evaluated previously in the
, Safety Analysis Report was experienced for the same reason as stated in i
the previous paragraph.
There was no reduction in the margin of safety as defined in the basis for Technical Specifications as' replacement of the temporary spacer in the restricting orifice (R0 Q1G410010) with a permanent spacer did not change the limiting conditions for operation applicability or surveillance requirements.
No Technical Specification change was required by this DCP. Nor was any unreviewed so/ety question identified.
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J14 MISC 87102101 - 30
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, ' Attachment L to AECM-87/0196 SRASN: NPE-87-164 000 NO: DCP 86-4028 Rev. 0 & Rev.;1., SYSTEM: Nil, N35- j y ,
)
DESCRIPTION OF CHANGE! This DCP'made'the following modificationsLfor 1st and'2nd stage.' reheater excess steam lines in the Moisture Separator /
Reheater' Vent: ,
s , .
Ic Removed 1st and 2nd stage reheater excess steam" orifices andJreplaced them by line size ~ orifices. '
2. Modified the e_xisting Ist stage. reheater excess steam localicontrol. ..
station'of control' valves to throttle desired' excess. steam flow rate.
at any power level.
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3. Modifiedtheexisting2ndistagetreheater.excesssteam' local [ control-station of control valves-to: throttle desired excess steam flow rate:
at any power level. ^
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REASON FOR CHANGE: Thi-s design change.is to' prevent level > fluctuations in. 1 both 1st and 2nd stage: reheater drain' tanks.
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1 SAFETY EVALUATION: ..There.is no increase:in the'probSbil'ity of' occurrence !
or in the consequences of an accident or malfunction'of equipment important to safety previously evaluated.-in'the Safety Analysis Report'as-
.]
q this design change did not affect the N11 and:N35 systems as' currently- -
described in UFSAR Section 10.3 text, since:it removed existing orifices-to provide desired excess steam flow at any. power level. This design change did not affect the occurrence.of an accide'nt as described in.UFSAR Section 15.6.4 (Steam System Piping Break Outside' Containment), since the steam piping has not been modified. Since N11 and N35 Systems serve no safety function,.=the failure of these ' systems would not compromise any safety-related systems, k
There is no creation of a possibility for an accident or malfunction of.
equipment of a different type than any evaluated previously in the Safety Analyses Report as the N11 and N35 systems have no~ safety-related.
function. These systems are not required to support the safe shutdown of-the reactor. This design change does'not affect.any safety-related.'
equipment but only enhances operability of the N35~ system.
There is no Technical Specification which established ma'rgin' of safety for excess steam flow requirements. Therefore, the margin of safety-is; not reduced by this design change.
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AttachnientII to? ECM-87/0196 7 ,
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NPE-87-175' SRASN: DOC N0: DCP 83 4025-506-R01' , SYSTEM:: 017.
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DESCRIPTION OF CHANGE: - This DCP installed redundant stask flow monitors-E for all six HVAC stack flow radiation monitors;as1 discussed in '
UFSAR=11.5.2.2.4.1.
REASON FOR CHANGE: This' change provides for' independent? stack monitoringj for accident range monitoring as' required by Reg.. Guide 1.97 Cat. 2-variable; 3
SAFETY EVALUATION: There was.no increa'se.in the probability of occurrence; or;in the' consequences 'of fan accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis: Report, The changes made were to a'monitoringisystem only. ' The ' installation off h
[
new, instrument. tubing is; supported per MP&LL JS-05, which provides adequate ;
guidance for two over one hazards. :ConduitLand tray supports have been seismically designed in areas where two:over one; consideration. exists, ,
i There .is no creation of a possibility for anJaccidentf or malfunction of a different. type than any evaluated previously:in-thelSafety Analysis:
Report. The installation of redundant'. flow monitors i improved system reliability by providing an independent stack; flow' monitoring capability' '
for the accident range monitor. .The new flow monitors are environmentally, qualified for those systems which are inza: post accident' harsh environment. This ensures that total. stack discharge can be read,on containment purge, fuel handling area,. stand-by gas, treatment'. It also provides an independent means of stack flow measurement for all' systems-and does not affect system operation. '
.There is no reduction in the margin of. safety.as'de' fined'init heibasis for any Technical Specification. The addition of the redundant' flow monitors increased the system's ability to monitor' post' accident discharge accurately. -
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, JAttachment,I;to.AECM-87/0196 ,
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SRASN: NPE-87-176 DOC NO: 'DCP-83-5021, Rev. 1 SYSTEM: C71. O j
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, DESCRIPTION OF CHANGE: This DCP revised power supplies of RPS sensors and . trip units', Neutron Monitoring System, Nuclear Steam Supply Shut-Off
']'I System, Leak Detection System and Process Radiation Mo'nitoring System from RPS Bus to Class 1E Vninterruptable-Power Supply (UPS).
REASON FOR~ CHANGE: This change.has been implemented to significantly reduce spurious scram signals and; prevent-overchallenge of ESF systems. .;
This will 'also improve the functional reliability of the subject systems. q l
SAFETY EVALUATION: There'.is-no increase in the probability of tccurrence i or. in the consequences of an accident or malfunction' of ' equipment '
important to safety previously. evaluated 'in the. Safety ~ Analysis Report. '
The switching of power supplies ~ from 'the RPS Bus to'UPS : increases the plant ability to operate more. effectively. . The. use of UPS power.in no:
way reduces the capability of the RPS to perform its intended. safety ,
function in mitigating postulated accident' conditions. This design. i j
change, in compliance.with UFSAR Section 8.3.1.2.1C' considerations,:is an enhancement which ensures the availability of UPS to the equipment. RPS =
power system is not an engineering safety feature, component or system'.~
The system itself fails in a fail safe mode, hence, providing UPS powe'r .
to the above described equipment in the RPS system does not increase:the j
probability of any malfunction. i
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This change does not create the possibility of an accident'of a different type than any evaluated previously in. the Safety Analysis Report. No new failure modes have been added per this DCP. Switching of the power supplies from RPS to UPS is an enhancement of the original design and the equipment is able to perform its intended function more effectively.
The addition of load to UPS affects the surveillance requirement for IE- -
batteries (providing alternative feed to. inverter). However, it has been determined that the batteries are capable of handling the. additional load l
requirement. This. change does not affect the limiting conditions for -(
'i
. operation or surveillance requirements for RPS instrument channels as 1
defined in Technical Specification 3/4.3.1. Hence, the margin of safety is not reduced.
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m Attachment I to'AECM-87/0196 SRASN: NPE-87-177- DOC NO: DCP 83-154-R0 & R1 SYSTEM: .C11 DESCRIPTION OF. CHANGE: This change added isolation. valves for the filter (C11-D006) in the instrument air line and bypass line filter (C11-D026) for the Control Rod Drive (CRD) system as shown in UFSAR Figure .
.4.6-7.
REASON FOR CHANGE: .This change allows any required maintenance to b'e performed on Filter C11-D026. or C11-D006 without affecting plant operation. v
. SAFETY EVALUATION: There is no. increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important.to-safety analysis repo_rt. The.C11 System operation'and.
. function'will.not change. . The piping and pipe supports supplied'by this change meet all applicable design requirements and.will function'in their intended manner.
This change does not create a possibility for an accident or malfunction of equipment;of a different type than any evaluated previously..in the safety analysis report as.the piping and pipe support changes meet.all l applicable design. requirements, r i The modified piping and pipe supports d not change the limiting conditions for operation applicability,or' surveillance requirements. The piping and pipe supports supplied by this change meet all applicable code requirements and will function in their. intended manner. Therefore, this change does not affect the margin of safety.
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J14 MISC 87102101 - 34 h
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Attachment i=to AECM-87/0196" J
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SRASN

NPE-87-179 DOC.NO: DCP 84-4GS0-501-R00 SYSTEM: P41 DESCRIPTION OF CHANGE: This change supplied the pipe support information required for modifying Unit I and Unit 2 shared pipe supports for the Standby Service Water-(SSW) Basin A as a result of the soil structure-interaction analysis.
REASON FOR CHANGEi Prior to restart after the first refueling.outageEper
' Unit 1 Operating. License, Condition 2.C.-(6) any structural . modification required as a; result of the soil structure interaction analysis'would be completed.
' ~
~ SAFETY EVALUATION: There is no increase in.the' probability.of. occurrence or in the consequences' of an accident or- malfunction.of equipment important'to safety previously evaluated-in the Safety. Analysis Report.
The consequence of a seismic event evaluated in the FSAR has not changed. The. seismic design criteria of the SSW basins is. upgraded; The.
SSW piping with modified pipe' supports meet all design requirements and will function in their intended manner.
This change does not create the possibility of an-accident or: malfunction of equipment of a different type than any evaluated previouslyLin the Safety Analysis Report. The consequences of.a seismic event evaluated in the FSAR has not cnanged. The seismic design criteria of the SSW basins-is upgraded. The SSW piping with modified pipe supports meet all design requirements and wi11 function in their intended manner. No new failure modes are introduced by this DCP.
There is no reduction in the margin of. safety as-defined in the basis for.
any technical specification. This DCP-does not change the_ limiting
' condition for operation, applicability, or surveillance requirements as defined in the Basis for Technical Specifications. The SSW piping with modified pipe supports meet all design requirements and will, function in their intended manner. Therefore, the margin of safety is not reduced.
J14 MISC 87102101 - 35 L
,j Attachment I to'AECM-87/0196 , ,-
5 1
. , 1 SRASN: NPE-87-180 DOC:.NO: LDCP 83-0229,'Rev. 0 SYSTEM: G18 H d
DESCRIPTION OF CHANGE: This change relocated'.th'e~ solid radwaste.
hydraulic compactor (NSG18D007) as identified in UFSAR,section 11.4.2.1 from elevation 118'-O, of the Radwaste' Building to' elevation' 93'-0",' in ' ~l the Radwaste Building. ,
REASON.FOR CHANGE: .This change was implemented since-the compactor' prevented efficient access to certain stairs, rooms and an elevator ini its previous location.
SAFETY EVALUATION: There is no increase in the probability of.' occurrence or in the consequences of.'an accident :or malfunction of equipment -
important to safety previously evaluated.in the Safety Analysis Report.
'I This equipment is non-safety related. l This change, does not create the ' possibility of an accident or malfunction-of equipment of a different' type than any. evaluated:previously in=the.
Safety Analysis Report. Radioactive' releases from a subsystem and-
' component, analyzed in FSAR Section 15.7, envelope postulated accidents a due'.to the design change.
There is no reduction'in the margin of safety as defined in the. basis for l any technical specification. .The design change'does not change the limiting conditions for operations, applicability, actions, or surveillance requirements as defined in the basis for Technical Specification 3/4.11.3 and 6.15.-
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1 J14 MISC 87102101 -~36 ]
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Attachment i to~AECM-87/0196 k
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NPE-87-181 DOC NO: DCP 83-4074-500-R00 & ROI . SYSTEM: G17 ,
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1 DCP 83-4074-501-R00= J.
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. DESCRIPTION OF CHANGE: These changes modified the liquid radwaste (Gli) system by. installing a Funda Filter bypass .for transferring liquid radwaste. from the floor drain collector ' tank,: the equipment ' drain collector tanks and the waste. surge tanks' to a mobile / portable vendor:
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' filtration processing station located in the Radwaste Building Rail / Truck Bay.
j-4 l REASON.FOR CHANGE: The modification will provide a _ permanent: alternate
'fil_tration capability and wi.11 facilitate maintenance outages.and .,
i emergency' operations'for liquid waste processing.
SAFETY EVALUATION: There is:no increase'in'the probability of occurrence-or in the conseque_nces 'of an accident'or malfunction of equipment- 'I
impo.rtant to safety previouslyLevaluated in the Safety Analysis Report.
The postulated worst case failures (radwaste tank' rupture and piping leaks) analyzed in UFSAR Sections 15.7.2 and 15.7.3 enveloped the occurrence and consequence of~ postulated accidents due to the-design- -1 change.
This change does not create the' possibility of an _ accident or malfunction of equipment of a different type than any previously in the Safety, Analysis Report. All potentially radioactive portions' of the transfer.
line systems are located within the Radwaste Building and designs were developed in accordance with the related guidance in Branch Technical Position ETSB 11-1 of NRC Standard Review Plan 11-2. Therefore, use of the radwaste transfer'line will not result in releases which differ from those previously predicted in UFSAR Section 15.7.'2 and 15.7.3 nor will-there be a change to an individual's exposure in the. unrestricted area.
l There is no' reduction in the margin _of safety as defined in the basis'for any technical specification. The design change does not change the limiting conditions for operation, applicability, actions, or surveillance requirements as defined in the basis for Technical-Specifications 3/4.11.1, 6;12 and 6.15.
0 J14 MISC 87102101 - 37
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i SRASN: NPE-87-182 . DOC N02 DCP 83-0567-R00 'SiSTEM: 'X57 l
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. DESCRIPTION.0F CHANGE: .- This. change relocated the fresh air intale for the counting room from.the east wall of the Water Treatment Building to
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the south wal1 away from the Radwaste Building exhaust. The counting room is located within the Water Treatment Building !
i REASON FOR CHANGE: The fresh air intake was previously located near -the: .j radwaste building exhaust.which; increased the chances for the i introduction of~ contaminants into the counting. room. 1 i
SAFETY EVALUATION: There'is no increase in the probability of occurrence i or in the consequences of an accident' or. malfunction of equipment :
important to safety previously evaluated in the safety. analysis. report. ,
The relocation of the fresh air intake allows the system to function as-desig nd. This component does-not affect any system important to safety. q This change does not create the' possibility of an accident or malfunction I
of equipment of-a different type than any evaluated previously.in.the.
Safety Analysis Report. Failure of this component or system.will not-compromise any safety-related component or system.
1 There is no reduction in the margin of safety as defined in the basis < for l any Technical Specification ~ No new failure modes are introduced since the-system operates as designed. l J
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J14 MISC 87102101 - 38 e
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L i Attechment.I to' AECM-87/0196 SRASN: NPE-87-183 DOC NO: .DCP 86-4003-R00 SYSTEM: N19 l -DESCRIPTION.0F CHANGE: This change. replaced previous component's on 1.
.six control valves'in the Feedwater and Condensate systems with Fisher Model 546 I/P Converters separately mounted from the' valve and a Bailey.
Model AP-4-12100' valve positioner for.each. A Fisher Model 2625 Booster Relay was also installed for each of.the subject' valves.
REASON FOR CHANGEi The_ previous I/P converters were.failing due to line i
l' vibration and were replaced with units mounted' separately from the valves.
. Booster Releys were added to help the valves.open fast enough to meet the design parameters of the control' loops.
SAFETY' EVALUATION: There is no increase in the probability of occurrence or in.the consequences of an accident or malfunction of equipment important to' safety previously evaluated in the safety analysis report.
E The modification of the valve controls will decrease the probability of valve failure. Therefore,.the probability of a loss of feedwater as-evaluated in Chapter.15.2.7 of the-UFSAR is reduced. 'The modified valves only affect the probability of an. occurrence"of'a loss of feedwater as evaluated in'15.2.7. -They perform no function ~which affects the consequences of the accident evaluated in Chapter 15.2.7 of the UFSAR.
The modification made increases valve reliability therefore, the probability of a malfunction of equipment important to safety is decreased.
There is no creation of a possibility for an accident or malfunction of ~
equipment of.a'different type than any. evaluated previously in the FSAR.
The original control scheme remains unchanged therefore, no now failure modes are created. There is no reduction.in the margin of safety as defined in the basis for~any technical specification as these valves are i not addressed in the GGNS Technical' Specifications.
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Attechmenti I to AECM-87/0196 l
' '. l y SRASN: 'NPE-87-184 DOC !C .- NPE FSAR 87-0013 SYSTEM: :B21 i 1
l' DESCRIPTION OF-CHANGE: FSAR Section 3.9.3 was revised to reflect more l accurate environmental qualification testing data.for the main steam L safety relief. valves .(SRVs). This change'more accurately! reflected the s l GGNS Environmental Qualification Central File based on the~changeout;of.
l the SRV solenoids for qualified components. .The physical plant changes-were performed under a previously implemented design change. No component. configuration changes were required based on the, solenoid-replacement.
~
REASON'FOR CHANGE: The FSAR,re' vision was performed to reflect accurats environmental qualification testing data for the mainsteam SRVs based on solenoid replacements.to meet 10CFR50.49.
SAFETY EVALUATION: :There is'no increase in the probabil'ity.ofioccurrence ' "
or in the' consequences of an accident or malfunction of. equipment l .important to safety previously evaluated in the safety. analysis. report.
l This change replaced non qualified solenoids with environmentally s
qualified solenoids.and-did not change'the system function.
There is no creation of a possibility for an accident or-malfunction of a different: type than any evaluated previously in the safety analysis report. No additional factors, with regard to plant safety or system function, were introduced by this change that have not been.previously evaluated.
1 The replacement of non qualified' solenoids with modified solenoids did not reduce the margin of safety.for any technical specification.
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-J14 MISC 87102101 - 40
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Attachment'I/to AECM-87/0196
, l 3 y ,s SRASNi NPE-87-185 DOC NO: ;DCP 86-3008-R011 0 ,
SYSTEM: E12 i
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DESCRIPTION'0F-CHANGEi This change installed an additionalikeylock switch on each-of the remote shutdown panels'(1H22-P150' and P151)' to disable / enable the open function: of. the F042A and B valves..in the RHR-system.
REASON FOR CHANGE: This change ensures that the. opening of these' valves; from the remote shutdown panels will be done'by deliberate Operator action only to prevent inadvertent overpressurization of the LPCI system. piping, SAFETY EVALUATION: There is no increase.in the probability of occurrence L
or in the consequences of an accident or malfunction-of equipment important to safety previously evaluated in<the Safety Analysis Report.
Control Room operation of-valves E12-F042A and B is unaffected by this change. The changes itoplemented by this change decrease ~ the. probability.
of over pressuri::ing RHR low pressure systems by inadvertent connection' to the higher pressure reactor coolant system at the remote shutdown panels. Therefore,-the probability of an! accident by control' room operation is unchanged; the probability of an accident by remote shutdown panel operation is reduced, due to the additional " enable"' function and administrative control of the keys. The use of' Class'1E switches with seismic qualification located'in mild environment ensures' proper circuit operation during design basis events. The consequences of LPCI failure s
are unchanged as a result of'this change, as no effect will be incurred upon control room operation under normal conditions and the only impact upon operation of the remote shutdown panels is that : instances where the' F042 valve switches would be, employed will be subject to'more' deliberate operator decision before any action may occur. Reg. Guide 1.75 separation employed by this design ensures that multiple ESF divisions l
will not be. impacted by a single failure.
There is no creation of a possibility for an accident-or malfunction of a different type than any evaluated previously in'the Safety Analysis-Report. The use of seismically qualified Class IE equipment in mild environment with Reg. Guide 1,75 separation will' limit accident possibilities to those previously evaluated.
There is no reduction of the margin of safety as there is no change to the bases of the Technical Specification regarding LPCI, Section 3/4.5.
p J14 MISC 87102101- 41 a i- t l.
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.Attschment I'to AECM-87/0196 SRASN: I NPE-87-186 . DOC NO: DCP 85-3111 . SY3 TEM: C61 DESCRIPTION 0F CHANGEi This DCP changed the previous.ESF Div.-I power-supply to'Div.I. Class.1E uninterruptible power for the: Remote Shutdown Panel RCIC-flow-control loop and indicating lights.
REASON FOR CHANGE: This design was modified.to comply with G.E. RCIC
. system design specifications.for D.C. power supply to be provided'on the RCIC system.
SAFETY EVALUATION: There is no increaselin the probability of occurrence or in the consequences of an accident or' malfunction of equipment: ..
important to safety previously evaluated in'the UFSAR. .'This. change did-not affect- any system function as. previously. evaluated tin the UFSAR. The.
change of the power: supply from ESF Div I power to Class 1E uninterruptible power'did not increase the consequences-of a malfunction'
'of any: kind previously evaluated in thelUFSAR.
There is no creation of 'a. possibility for an! accident Lor malfunction ofi equipment of a .di.fferent type than any evaluated previously in the Safety Analysis Report. No additional factors with' regard to plant safety or system function, were introduced by this change that were not'previously-evaluated.
The change of the power supply from ESF Div I power to Class IE uninterruptible power does'not reduce the' margin of safety for any.
technical specification.
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-J14 MISC 87102101 - 42 V .
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~ Attachment 7 to-AECM-87/0196; -t SRASN: 'N'PE-87-188 ' DOC N0': DCP 83-4100/ 'Rev 0 , : SYSTEM: B211 DESCRIPTION OF CHANGE: .This change removed energy' absorbing material. in' the +Y (vertical-up). direction for pipe whip ~ restraints (8A & B and.10AV
&.B) on:Feedwater. Piping.
-REASON FOR CHANGE: 'This change resolved therthermal expansion problem for feedwater piping during nuclear heatup by1 removal.of the pipe whip I
restraints.
SAFETY EVALUATION: There'is no increase in.the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report.
Pipe whip protection 'in .this direction is no longer' required.
Reference:
Bechtel Report " Protection Against Dynamic Effects Associated with Postulated Ruptures of Piping"... Appendix 0; and.UFSAR' Figures 3.6A/B.
There is no' creation of a possibility.for an accident or malfunction of-a different type than any evaluated previously in the Safety Analysis Report.
There is no reduction in the margin of safety as defined in the basis for any technical specification as pipe whip. restraints are not addressed in' the Technical Specifications.
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-SRASN: NPE-87-192- DOC NO FSAR CN 3762 2
SYSTEM: E'31! '
DESCRIPTION OF. CHANGE:..This Change' Notice updated UFSAR-Figures:
7.6-2b; 7.6-2d; 7.6-16'and 7'6-17 to reflect the design change (DCP 82-4178)~which installed' sight glasses to replace seven turbine meters (1E31FTN023A, B, C, D and 1E31FTN033A, B~,
. C) and associated instrumentation.
REASON,FOR CHANGE: Th'e existing. turbine meter instrumentation was replaced since the meters.are norma 11y ' empty.and, if leakage flow l occurred, rust-and scale could clog the~ meters.
' SAFETY: EVALUATION: No accident evaluated in'the UFSAR postulates a failure of the existing flow monitoring subsystem. lThis' subsystem
. performs'no automatic safety function. The. piping and pipe supports
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added by this cesign change meet all; applicable design. requirements and-L will function:in-their intended manner. ~There'is no-increase in the-probability of occurrence or' in' the -consequences of an accident or.
malfunction of equipment important to: safety previously evaluated in.the Safety Analysis Report.
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There is no creation of a possibility for an accident or malfunction of a' different type than any evaluated previously,in the Safety Analysis-Report. Only one: of the' turbine' meters to' be_ replaced monitors drywell leakage. This-meter, 1E31FTN023A, monitors leakage from the refueling
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i. bellows from which leakage could be expected only when= the drywell head l cavity is flooded during refueling. Since the 25.gpm limit is not applicable in modes 4 and 5,' control, room, indication. and alarm ,is not -
reouired. Failure of ~ this flow monitoring subsystem could result in an initially undetected leak in:the refueling bellows,or pool. liners.
However, any leakage in excess of 25GpM'wouldfultimately be detected and-l annunciated by sump' fill or pump out. timers. _'This flow monitoring subsystem performs no automatic safety function.
No margin of safety as . defined. in _the basis' for' any technical specification is affected by this flow monitoring ' subsystem. There-is no.
change in the GGNS Unit-One Technical Specification since the-limit of 25 gpm for identified leakage (30 gpm total leakage less 5 gpm unidentified leakage) is not affected.
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'J14 MISC 87102101'--44 , ,
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Attachment I to:AECM-87/0196;!
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.:SRASN! NLS-87-001' DOC NO: 'GGNS, Emergency PlanfR14L
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' DESCRIPTION 0F CHANGE: This documentLevaluated-changes.to the GGNS ,1 Emergency Plan, for-Title Changes,L EPZ Population Updates, Clarification - ..
of Responsibi.lity," Organizational Charts, Transition Plan, NRC J '
'a Commitments, and Training.
1
REASON.FOR CHANGE: The reason.for the changes wasito update.the 1 Emergency Plan by updating Titles, Organizational. Charts, Training
' Criteria, Fulfilling NRC Commitments,: for -10CFR50 Appendix E.
SAFETY. EVALUATION: :There is no increase in the' probability'of occu'rrence i or.in.the consequences of an accident or malfunction of equipment important to safety previously evaluated irr the Safety. Analysis; Report.
The revisions are mostly administrative and will' enhance the capability. ;
of the Emergency Organization with the -increased clarification of
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Emergency Plan. criteria. These Emergency. Plan revisionsodo not. involve l equipment. operation and adds conservatism to the plan; Therefore,'there is no creation Hof a possibility for an accident uor malfunction of a different type than any evaluated previously in the Safety, Analysis.
Report.
l These revisions will enhance the margin of-safety by providing the most' current update of administrative information required for the successful 1 overall operation of the GGNS Emergency Preparedness' program.
This evaluation is for GGNS Emergency Plan Rev. 14.
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J14 MISC 87102101 - 45 '
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Attachment I to AECM-87/0196' SRASN: NLS-87-002 DOC NO: FSAR CN NLS-87-004 SYSTEM: E12 f 1
i DESCRIPTION OF CHANGE: This change provided proper identification of ~the j Suppression Pool Cooling as a mode of RHR operation for UFSAR Subsection i 1.2.2.4.9.2.
REASON FOR CHANGE: To provide consistency with UFSAR Section.6.2.2.1.
SAFETY-EVALUATION: 'This UFSAR change was editorial only and has no .
effect on the operation or design of the Suppression Pool Cooling mode of RHR or the Technical Specification requirements of 3/4.6.3.3. No ,
unreviewed safety question was identified. l 1
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J14 MISC 87102101 - 46 i
Attachment I to AECM-87/0196 ,
i SRASN: NLS-87-003 DOC N0: FSAR CR NLS-86-051 SYSTEM: E61 DESCRIPTION OF CHANGE: The UFSAR was changed from manually initiating the hydrogen recombiners when the RPV water level reaches the top of active fuel to the design basis value of 3.5 percent hydrogen concentration.
REASON FOR CHANGE: This change corrected an' imposed commitment for early actuation of the hydrogen recombiners that is more conservative than the FSAR design basis analysis.
SAFETY EVALUATION: The FSAR design criteria in UFSAR section 6.2.5.2 states that the hydrogen recombiner system will be manually initiated when the hydrogen concentration in the containment has reached 3.5 volume percent, so that the hydrogen concentration in the containment is maintained at or below 4 volume percent. This change therefore, reflects the design criteria by changing the manual initiation criteria back to the design basis value of 3.5 percent hydrogen concentration.
Therefore, there was no increase in the probability of occurrence or in the consequences of an accident or malfunction of equipment important to safety previously evaluated in the Safety Analysis Report.
There is no creation of a possibility for an accident'or malfunction of a different type than any evaluated previously in the Safety Analysis Report.
Since this change does not involve a relaxation of the criteria used to<
establish the safety limits, the bases for limiting safety system settings, the bases for limiting conditions of operation, a change to technical specifications or a change in plant operation, the change will not reduce the margin of safety as defined in the basis for any technical specification.
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I J14 MISC 87102101 - 47
ATTACHMENT-'II 10 CFR50.59 EVALUATED CHANGES TO THE UFSAR (REV 2) BUT'NOT PREVIOUSLY REPORTED'TO THE NRC' 1
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l J14 MISC 87102701 - 1 L_______________________.______ _ . _ _ . . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _
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Attachment II to AECM-87/0196' 10CFR50.59 EVALUATED CHANGES T0 THE UFSAR-(REV. 2) BUT.NOT PREVIOUSLY REPORTED TO'THE NRC DOCUMENT FSAR EVALUATED SECTIONS SUBJECT CR NLS-87-003 .1.2,2 : This change provides needed clarification and 6.0 identification'of plant ESF. systems in thel 7.3.1 FSAR.
8.1.3 9.2.1 CR NLS-87-057 8.1.2 This FSAR change provided reference'to the 8.2.1 SERI Switchyard agreement with MP&L to. assure i 8.2.2 continued compliance with GDC 17. I I
1 CR NLS-87-064 8.2.2 This FSAR change updates outage data for'the Tables offsite 115 & 500KV transmission lines.
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CR NLS-87-065 Figures This FSAR change updates the System Maps for l 8.2-1 Mississippi Power &-Light and Middle South l
&2 Utilities to reflect the latest system grid revisions.
CR NLS-87-067 13.1 This' change request is for various '
organizational changes over the past year.' ;
. 1 CR NLS-87-068 13A The revision to' Appendix 13A. reflects' changes to update the. resumes of key personnel ,
associated with the operation of GGNS. j l
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j CR NLS-87-074 9.5.2 .The UFSAR Subsection 9.5.2 is.being modified to reflect changes in'the communicate _ons system. ,
CR NLS-87-075 5.4.8 This document provided a summary description .'
of the pre pump'and post pump' modes of Reactor Water Cleanup (RWCU) system operation.-
a CR NPE-86-0034 3C.3 This software change to,the UFSAR is'.to exempt' the moderate energy piping inside the Diesel Ei Generator building from postulation of pipe'.
cracks per NRC guidance.
i J14 MISC 87110503 - 1
Attachment II to AECM-87/0196 '
10CFR50.59 EVALUATED CHANGES TO THE UFSAR (REV. 2) BUT NOT PREVIOUSLY REPORTED TO THE NRC
- DOCUMENT FSAR EVALUATED SECTIONS SUBJECT CR NPE-86-0069 3.5.1 These changes exempt all pressure vessels and-Table pressurized bottles containing non-condensible 3.5-5 gases with. operating pressures at or below 275 psig from potential missile source .i evaluations.
l Material 3.7.4 This SE was for MNCR-740-83 3rd Submittal Nonconformance which stated that peak recording Report accelerographs (PRA),-1C85-XR-R011, R012, 1 MNCR 83-740 R013, & R014, were being abandoned, but would 3rd Sub remain installed.
(CR NPE-87-003)
CR NPE-87-009 7.5.1 Subsection 7.5 of the FSAR is being revised to incorporate the GGNS Reg. Guide 1.97 position !
which has been approved by the NRC per the SER j dated 1-12-87. I l
CR NPE-87-0021 Various CR 87-0021 incorporated later revisions to I Figures several drawings into the UFSAR.
i CR NPE-87-0026 3.11 Question & Response 040.32 which deals with 1 submergence of equipment during a.LOCA, is being modified and incorporated into Section 3.11 of the UFSAR. ;
CR NPE-87-0029 3B.10.1 This CR incorporated changes relating to HCU
&2 floor loads due to a LOCA and the corresponding drag and impact loads on structures below the HCU floor and above the pool surface.
CR NPE-87-0031 5.4.9, FSAR inaccuracies identified in the review of 6.2.4 & the GGNS LLRT program performed per a Table commitment of the response to Licensee Event 6.2-49 Report (LER) 86-002-0 were corrected by this change.
l J14 MISC 87110503 - 2 L_______.____._m_. .____._.__-._m_____ . _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . . _ . _ . . _ . _ _ _ _ . -_ _ _ _ - ___....mm_. _._ . . . . _ _ _ _ . __ _ . _ _ . _ . _ _ ._ _ _ _ . _ _ _ . _ .
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Attachment.II'to AECM-87/0196 I
'10CFR50.59 EVALUATED CHANGES T0'THE'UFSAR.
(REV. 2) BUT NOT PREVIOUSLY REPORTED TO THE NRC DOCUMENT FSAR EVALUATED' SECTIONS SUBJECT MNCR 0106-86 Fig. This change added oil demister not shown on. l 2nd Sub. 9.3-2f, P&ID M-1067G and changed equipment numbers on.. l (CR NPE-87-038) 9.3-1, oil demisters located on.the instrument and- J 9.3-3 service air compressors. j sh. 2 l of 3- )
l CR NPE-87-045 Tables This change request revised the tables to show. -)
6.2-44 one (1) missing valve.and to correct General i 6.2-49 Design Criteria Applicability. .I
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MNCR 92-87 11.4.2 MNCR 92-87' reported a discrepancy.between the (CR NPE-87-046) statement in FSAR'11.4.2.4.7 (second paragraph), which says that "De-energize-to-open solenoid valves provide a source of flushwater to the pump casing upon loss of.
electrical power" and the 'f as-built" condition of the plant in which the solenoid valves,are de-energize-to-close models. The'FSAR.was revised to reflect that~a different method of .
flushing is utilized & that the de-energize- l to-open solenoid valve feature is not )
required.
CR NPE-87-052 8.3.1 This change deletes item (m) which incorrectly !
includes " Fuel Oil day tank low-low or empty" as a condition where the diesel generator is ;
rendered incapable of responding to an' l emergency auto start., signal.
CR NPE-87-054 3.5 & This revision replaces existing'UFSAR Figures 3.6 for high energy line breaks and internal missiles with controlled engineering drawings.
Adds UFSAR figures for 2" & under high energy lines based on controlled engineering drawings. Removed existing UFSAR Tables for.
high energy line stresses. Incorporated the evaluation of hign energy line breaks'in the vicinity of RWCU drywell isolation valve F251 (RWCU heat. exchanger room).
!- J14 MISC 87110503 - 3 b__ _ _ _ _ _ _ _ _ _ _ _ _ _
1 Attachment II to AECM-87/0196 l 4
10CFR50.59 EVALUATED CHANGES T0'.THE-UFSAR -)
(REV. 2) BUT NOT PREVIOUSLY REPORTED TO THE NRC- ;
DOCUMENT FSAR EVALUATED SECTIONS SUBJECT CR'NPE-87-056 .15~7.6
. These changes clarify the discussion of a fuel-handling accident'(FHA) inside containment,' so-that it more clearly describes the assumptions-and results of:the' engineering calculation and incorporates editorial corrections.
MNCR 0592-86 Fig. . Figures were revised based on an. engirieering (CR NPE-87-058) . 9.4-1 review which identified " engineering mark-up" .!
& 8b discrepancies. This. included revising the {
location of' supply and discharge lines (1"-HCC-56 & 1"-HCC-57) and showing 2 capped lines (1"-HCC-85).
CR NPE-87-0059 3.10 This change addresses (1) the impact of the relocating a large portion of.the body of-seismic qualification details to the' Seismic Qualification Central File (SQCF),
(2) the addition of,UFSAR text-detailing GGNS operational stage seismic qualification.
criteria and practice,.and'(3)-the' general update of UFSAR statements which address GGNS
~
compliance with IEEE qualification standards.
CR NPE-87-0066 2.5.4 This change increased the; allowable interval-for surveying the Turbine Generator Pedestals due to ALARA concerns andl allow for missed' data due to' marks becoming inaccessible.
NPE ADMIN Multiple This evaluation'was performed.to address-l PROCEDURE Figure certain minor revisions to drawings'(based on
~
l NPEAP 807 Changes non-hardware deficiencies) which do not (CR NPE-87-081) involve facility changes. Any physical work-(CR NPE-87-064) to be accomplished in the plant will be.
1 (CR NPE-87-087) performed and controlled under other programs /
L procedures, 1
CR NPE-86-0089 Table This change request is to revise Table 6.2-49 6.2-49 to reflect that testing fexible-wedge gate-valves inboard disc at low pressure in the reverse direction is considered an' equivalent' test method.'
'J14 MISC 87110503 4
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Attachmest 1I to AECM-87/0196 >
10CFR50.59 EVALUATED l CHANGES TO THE UFSAR-
-(REV. 2) BUT NOT,PREVIOUSLY REPORTED TO THE NRC f
DOCUMENT- FSAR EVALUATED SECTIONS SUBJECT CR NPE-86-0098 6.2.3 LThe UFSAR' description of the 2 1/2' inch.or-larger. lines which penetrate the secondary containment boundary with only one secondary containmsr.t isolation' valve. requires revision..
to add two four inch fire protection' system (P64) lines.
MNCR 1055 Table This MNCR establishes minimum acceptable (CR NPE-87-051) 9.4-7 design flow of 23.5 GPM for cooler Q1T46B003B-B.
l MNCR 1197-86 Figures: MNCR.1197-86 documents the deficiencies noted (CR NPE-87-057) 9.5-11 on NRC' Inspection-Reports 50-416/86-26 and 9.5-11a 50-417/86-04 paragraphs'7a,-7b, and 7d. _ These:
9.5-12 deficiencies consist of incorrect P75 valve l 9.5-12a numbering and omission of root valve's on-P& ids.- These drawing nonconformances will be alleviate'd by implementing EAR'87/00042 and-NPESC 87/0046.
CR NPE-87-0058 Figure Drawing M-0035B (UFSAR Figure 9'.5-2))will be' i 9.5-2 revised to accurately reflect the as-built i plant in regard to. Secondary Containment. l Isolation By-Pass piping connections.
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CR NPE-87-0067 9.2.1 & The UFSAR section and table are being revised !
Table to be consistent with NRC guidance published !
9.2-1 in NUREG-0138 and SECY-77-439 which state that i credible passive SSW failures that can result- j in a loss of fluid post-accident are' limited j l
to pump _or valve seal' leakage, not. ruptures of' .!
I SSW system piping.
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1 Attachment II.to AECM-87/0196 10CFR50.59 EVALUATED CHANGES TO THE UFSAR-(REV. 2)'BUT NOT PREVIOUSLY REPORTED TO THE NRC DOCUMENT FSAR EVALUATED SECTIONS SUBJECT l
CR NPE-87-0070 9.2.1 This change request addresses proposed changes 9.2.5- to the UFSAR Section 9.2.1 and 9.2.5 UHS ;
Appendix cooling. tower. performance analysis. These- l 3A changes (1) clarify the discussions of the i Regulatory Guide 1.27 UHS cooling tower performance analysis, so that it reflects !
system capabilities assuming single unit 1 operation, (2) incorporate the peak heat load j for the Unit I high density spent fuel storage i racks (HDSFSR), and (3) incorporate the UHS cooling tower performance analysis results assuming single unit operation and peak HDSFSR heat load. Proposed UFSAR figure.. changes update figures to. reflect analysis results.
CR NPE-87-0071 5.4.6 This change addressed the maximum differential pressures expected for the RCIC system motor-operat2d valves which were changed to reflect- ;
i the conditions stated in the GE RCIC Design l l
Specification or as determined by IEB 85-03.
CR NPE-87-073 9.2.10 This change reflects the addition of PSW Table Radial Well No. 4.
9.2-13 9.4.9 l Table 9.4-12 Quality Various Various Fire Protection System P & ID's were Deficiency Figures revised to more clearly reflect those portions l Report in of the fire loops which are required for )
QDR 010-87 Section Unit 1 operation and those which are for (CR NPE-87-074) 9. 5 Unit 2. Per procedures, those valves >
providing an interface between Unit 1 and 2- I are locked closed. The appropriate P & ID's were revised to reflect the locked closed valve positions. Other valve position discrepancies between the procedures and the P & ID's were corrected.-
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CR PLS-87-006 6.4.2 This change request revises the description of the recharging capability for the Control Room portable self-contained air breathing units.
J14 MISC 87110503 - 6
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Attachment. 11 to AECM-87/0196 10CFR50.59 EVALUATED CHANGES TO THE UFSAR ]
(REV, 2) BUT NOT PREVIOUSLY REPORTED TO THE NRC j I
DOCUMENT FSAR EVALUATED SECTIONS SUBJECT !
CR NPE-87-0051 Tables This change updated the flow rates of i i
9.2-3, components in the SSW A and B loops.
16, 17
9. 4- 1, 7 l l
CR NPE-86-12 7.1.2 This change permitted RPS cables outside the .i Main Protection System cabinets to be run with .]
other ESF wiring of the same division in the same raceway system.
CR NPE-87-001 Table This change provides isolation between 8.3-10 Class IE power distribution panel 15P61 and remote shutdown panel room heat pump N1Z77B003-A. Isolation is achieved with redundant Class 1E overcurrent protection devices (Breaker & Fuse).
CR NPE-87-0079 9.2.1 Subsection 9.2.1.2 requires revision to indicate that " suitable isolation capability" has been provided at SSW/ nonessential system interfaces to assure that the SSW system I safety function can be accomplished assuming a j single failure. l J
QDR 448-86 F gure This revision was made to accurately reflect-(CR NPE-87-087) s.3-24 stop check valve Q1C41F007 on P&ID M-1082.
i QDR 63-87 Figure This revision was made to accurately reflect (CR NPE-87 's7) 6.2-81 on P&ID M-1091 that flow control valve NIE61F515 does not exist but rather a restricting orifice had been installed.
i FSAR nange Table This change notice reflects changes resulting Noti e 3785 7.5-2 from incorporating the GGNS R.G.1.97 position into the UFSAR.
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1 J14 MISC 87110503 - 7 I
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AECM-87-0196, Rept of Changes,Tests & Experiments Determined to Be Reportable Under Requirements of 10CFR50.59 for Jan-May 1987

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AECM-87-0196, Rept of Changes,Tests & Experiments Determined to Be Reportable Under Requirements of 10CFR50.59 for Jan-May 1987

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