ML20107L555

1995 Annual Environ Operating Rept

ML20107L555
Person / Time
Site:	Grand Gulf
Issue date:	12/31/1995
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GNRO-96-00046, GNRO-96-46, NUDOCS 9604290355
Download: ML20107L555 (66)

Text

n iY.

_ Entsrgy Opsratisne,inc.

ENTERGY eo sox 7se Pret Gibson.MS 39150 1 Tel 601437 2800 C. R. Hutchinson vce Presdent Opratuis Grand Gun Nutk'ai L W;r ,

April 23, 1996 l

1 U.S. Nuclear Regulatory Commission Mail Station P1-37 l

Washington, D.C. 20555 i Attention: Document Control Desk l

l

Subject:

Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 l 1995 Grand Gulf Nuclear Station (GGNS) Annual Environmental Operating Report (ABOR)

GNRO-96/00046 l

l Gentlemen l l

Attached is the Grand Gulf Nuclear Station (GGNU) Annual Environmental Ooeratina Reoort (AEOR) for the period l January 1, 1995 through December 31, 1995. This report is submitted in accordance with the Environmental Protection Plan, Appendix B to the GGNS Operating License (NPF-29),

Section 5.4, " Station Reporting Requirements".

l If you have any questions or require additional information concerning this report, please contact Michael J. Larson at (601) 437-6685, or this office.

Yours trul 1

( <

l CRH/ /

l attachment: 1995 Annual Environmental Operating Report l

cc: (See Next Page) i 9604290355 951231 PDR ADOCK 05000416 R PDR l 290130 .6 G9603251 /

1 April 23, 1996 l GNRO-96/ 00046 Page 2 of 3 cca Mr. J. E. Tedrow (w/a)

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

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

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

Mr. J. W. Yelverton (w/o)

Mr. L. J. Callan (w/a)

Regional Administrator '

U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 l Arlington, TX 76011 Mr. J. N. Donohaw, Project Manager (w/2)

Office of Nuclear Reactor Regulation ,

U.S. Nuclear Regulatory Connaission 1 Mail Stop 13H3 1 Washington, D.C. 20555 l i

I l

4 1

1 I

G9603251

April 23, 1996 i GNRO-96/ 00046  !

Page 3 of 3 bcc Mr. J. D. Barlow (w/a)

Mr. D. G.-Bost (w/a)

Mr. C. A. Bottamiller (w/a)

Mr. R. W. Byrd (w/a)

Mr. L. F. Daughtery (w/a)

Mr. L. F. Dale (w/a) i Mr. J. G. Dewease (w/a)

Mr. M. A. Dietrich (w/a)

Mr. C. M. Dugger (w/a)

Mr. J. J. Hagan (w/a)

Mr. C. C. Hayes, Jr. (w/a)

Mr. M. J. Larson (w/a)

Mr. M. J. Meisner (w/o)

Mr. R. L. Patterson (w/a)

Mr. T. L. Williamson (w/a)

File (LCTS/RPTS) (w/a)

File (Hard Copy) (w/a)

File (NSERA) (w/a)

File (Central) (w/a) ( 69 )

l l

l 1

G9603251 I

4a&-p=3- -,a+aJ> a m -dha- +w_4s -amr-- -

• d4-4"1 -JMd # A * - ----

l i

l -

1 f

i Li ,

i GRAND GULF NUCLEAR STATION l!

i a

. 1995 ANNUAL ENVIRONMENTAL OPERATING REPORT 3

l

l l

l 1

PREFACE l

l The Annual Environmental Operating Report (AEOR) provides information

~

and data obtained from implementation of Grand Gulf Nuclear Station's (GGNS) Environmental Protection Plan (EPP), Appendix B to the GGNS Operating License (NPF-29), which only requires terrestrial issues to be addressed, for the period January 1 through December 31,1995.

The GGNS Final Environment Statement did not identify any aquatic issues.

Consequently, the EPP does not address any. The GGNS National Pollutant Discharge Elimination System (NPDES) Permit issued by the Mississippi Department of Environmental Quality (MDEQ) contains effluent limitations and monitoring requirements for aquatic matters. The MDEQ regulates .

matters involving water quality and aquatic biota.

This report addresses only those issues required by the EPP. In the past, the AEOR included activities associated with the GGNS Constmetion Permit, and an Updated Final Safety Analysis Report (UFSAR) requirement which involved reporting regional and perched groundwater levels and precipitation data in the AEOR. However, the Nuclear Regulatory Commission approved cancellation of Construction Permit CPPR-119 for Unit 2 on August 21,1991 (GNRI-91/00176), and GGNS deleted the UFSAR AEOR reporting requirement in 1993 (GNRI-93/00025); therefore, GGNS terminated reporting activities associated with these items.

l l

l l

l j __ _ _

i l

TABLE OF CONTENTS l

PAGE PREFACE... . . . . . ii SECTION TOPIC

1.0 INTRODUCTION

.. . . . . . . . . . I 1.1 Impact Assessment and Summary.. 1 2.0 ENVIRONMENTAL SURVEILLANCE ACTIVITIES.. . 1 2.1 Transmission Line Surveys.. . . 1 2.2 Cooling Tower Drift Program.. . . . . . 1 2.3 Environmental Evaluations.. .. . 1 3.0 OBSERVATIONS AND DISCUSSIONS.. .. .. 2 3.1 Environmental Evaluations.. 2 4.0 ADMINISTRATIVE REQUIREhENTS.. .. 2 4.1 EPP Changes.. . . 2 l

4.2 EPP Noncompliances.. . 2 4.3 Nonroutine Reports.. . . .. . . 2 l 4.4 Potentially Significant Unreviewed Environmental Issues.. . . 2 l

l l

l 1

- lii -

i l- .

i

1.0 INTRODUCTION

1.1 Imoact Assessment and Summary GGNS personnel monitored the environmental impact of plant operational activities between January 1 and December 31,1995. The monitoring results contained in the following sections indicate no adverse impact on the environment due to operation of GGNS. In addition, GGNS personnel have not observed harmful effects or evidence of trends toward irreversible damage to the surrounding environment at GGNS.

2.0 ENVIRONMENTAL SURVEILLANCE ACTIVITIES 2.1 Transmission Line Surveys GGNS discontinued this program in 1988.

2.2 Cooling Tower Drift Program I

GGNS discontinued this program in 1992.

2.3 Environmental Evaluations The EPP permits changes in GGNS design or operation and performance of tests or experiments that affect the environment, provided they do not involve a change in the EPP or an unreviewed environmental question.

However, EPP requirements do not apply to changes, tests or experiments which do not affect the environment. Also, EPP requirements do not relieve GGNS of 10 CFR 50.59 requirements, " Changes, Tests and Experiments," which address the question of safety associated with proposed changes, tests and experiments.

1 I

i I

The EPP excludes changes, tests or experiments from the evaluation:

- If all measurable environmental effects confined to onsite areas previously disturbed during site preparation and plant constmetion, or

- If required to achieve compliance with other federal, state or local requirements.

3.0 OBSERVATIONS AND DISCUSSIONS 3.1 Environmental Evaluations Review of 1995 environmental evaluations indicate that none of the changes made involved an unreviewed environmental question per the EPP.

A review of evaluations conducted did not reveal any potentially significant unreviewed environmental issues. Table 4-1 provides a summary of evaluated changes which could have affected the environment. The evaluations are attached.

4.0 ADMINISTRATIVE REQUIREMENTS 4.1 EPP Changes GGNS made no changes to the EPP in 1995.

4.2 EPP Noncomoliances GGNS activities contained no EPP noncompliances during 1995.

4.3 Nonroutine Reports GGNS submitted no nonroutine reports in 1995.

4.4 Potentially Significant Unreviewed Environmental Issues Review of 1995 environmental evaluations indicated that none of the changes made involved any unreviewed environmental questions per the EPP. A review of evaluations conducted did not reveal any potentially significant unreviewed environmental issues. Table 4-1 provides a summary of evaluated changes which could have affected the environment. The evaluations are attached.

2

1 1

l l

TABLE 41 .

1 1995 ENVIRONMENTAL EVALUATION

SUMMARY

SAFETY AND I ENVIRONMENTAL DESCRIPTION I EVALUATION NUMBER 95-0054-R00 The activity involves changing the wording of UFSAR Section 18.1.34 so that minor leakage around vent and drain valves does not have to be eliminated. Instead, such leakage must be maintained as low as practical. I Leakage will receive treatment in the normal manner before leasing the plant. I Therefore, all releases will continue to be in compliance with established criteria and specifications for the plant. As a result, no unreviewed environmental question exists and a change to the EPP is not required.

95-0057-R00 This activity allowed performing the reactor vessel in-senice leak test with I the disc removed from IE12-F050B. The conditions and flow path for the test '

will remain basically the same. This activity will not result in a release to the l environment. As a result there is no unreviewed environmental question and I no need to change the Environmental Protection Plan (EPP). l 95-0059-R00 This activity revises the maximum allowable stroke time for various primary I and secondary containment isolation valves. Release from the plant must continue to comply with established criteria and specifications. This actisity is not increasing the probability, quantity or consequences of a steam release beyond that previously evaluated in the FES. As a result, no unresiewed l I

emironmental question exists and no change to the EPP is required.

95-0060-R00 This change removes the commitment to submit a summary startup report.

Since there will be no release to the environment or change in power level, there is no unreviewed environmental question and no change to the EPP is required.

95-4075-R00 This change deletes the requiremer.t to perform Type C local leak rate testmg on nine test connection valves. Functional operation of equipment will not be altered by deleting this testing requirement. As a result, there will be no l change in effluents or power level. Therefore the activity will not involve an l unreviewed environmental question o'r require a change to the EPP.

95-0072-R00 This activity allows control rod drive system drive water pressure to be temporarily increased up to 475 psi above reactor pressure during withdrawal.

The activity will not change operational design or monitoring and release of effluents so there can be no change in efiluents of power levels. As a result, there is no unreviewed environmental question or need to change the EPP.

95-0078-R01 This activity deletes requirement to have a pre-planned alternate method of monitoring when the AXM noble gas radiation monitor is inoperable.

Routine monitoring of plant effluents will still occur via established monitoring points and equipment. No effluent limitation is being changed or removed thus there will be no change in effluents or power level and no need to change the EPP.

l 3

l l

l 1

l l

1 I

l 1

I i I i  !

i i

)

i i

4 1995 ENVIRONMENTAL l i ,

i EVALUATIONS i 1

i <

i  !

i i l

4 b

a i

1 r

I i

i

}- .

i 4 .- 96- oG6 - FesE. i 1

GRAND GULF NUCLEAR STATION UNTT 1 l

CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM

{

k L Safety Evaluation Overview i

i A. Reference Data 1

l OaxHNATOn: Scott Kirby DErr/sECr: P&SE EVAI f: 95-0054-R00 DOCUMENT EVAWARD: UConsina Document Chance Reauest 95-036 9

REFERENCES:

UFSAR 18.1.34. Technical Soecification 5.5 . NUREG -0737 i

j FSAR CHANGE REQUULED7 E Yes O No Ca# 95-036 l i

l FSAR sECDONS M BE REVISED: FSAR Section 18.1.34 i

'IRM CHANGE REQUIRED 7 0 Yes o No TECH. SPEC. CHANGE REQUDLED7 0 Yes a No Ca# {al.al Is THE VAUDfrY OF THIs SAFETY EVAWADON DEFENDEKr ON ANY CHANOES OTHER DIAN O Yes i DE CHANGE BEINO EVALUATED (E.O. PROCEDURAL, OPERADONAL CONDITIONS)7 D No Ir VEs TO THE I.Arr QUE3 DON, HAVE THE ORGANIZADONS RESPONSIBt2 PORTHOSE CHANOES BEEN NorIFIED7 O Yes j

THE RESPONSIBUE ORGANIZ.ADONS MUST BE NanFIED FluOE TO IMPLIhGDrnNO f nas CHANGE.

{

j B. Executive Summarv (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT) saar DEaCairnON Or CHANOr., nrr On ExrEnzutwr: UFSAR Section 18.1.34 lists reauirements for the intearity of systems outside Cont-ininent likely to contain f

i radioactive material for Pressurtzed Water Reactors and Boilina Water Reactors.

I The paracraoh in this section dealina with Water Leakaae recuires observable l leakaae oast vent and drain valves be eliminated. The paraaraon dealina with Gas Leakaae reauires any detected leakaae be eliminated, in each instance. the word

' eliminated' will be raciaced with ' reduced to as-low-as-oractical levels'.

REASON FOm CHANOE, ner On Exrnamrr Technical Specifications and the UFSAR est*N3sh release limits for water and aas leakaae to the environment. The reauirement to totally eliminate specrfic leakaoe oaths can cause sianificant system intrusions with little aooreciable affect to total leakaos rates.

SAFETY EVAWADON

SUMMARY

AND CONCW510Ns: This chance corrects an overiv-burdensome reauirement. and is not contrary to the letter or spirit of Technical Soecifications or aoolicable NUREGs. Other testina already in olace determines collective leakaoe for water and oss systems. with corrective achon roouired for leakaae above an acceotable amount. No unresolved safety ouestions are introduced. and this chance would not affect the imoact of oostulated accidents.

, - - - . - . - ,-g-

}- ,

4 .

=

h l

h IL Safety Evaluation O Not applicabic per Safety Evaluation Applicability Review

A. Technical Spa ifications l

! 1. Implementation or performance of the action described in the evaluated O Yes

] document will require a change to the GGNS Unit 1 Technical a No Specifications l

j ggis The Technical Soecifications. section 5.5. reauires controls and i orocedures to reaulate the release of radioactive materials. This l UFSAR chance does not alter any Tech Soec reauirements. Tech Soec

! section 5.5.2 reauires leakaae to be minimized to levels as low as j oractiemble. which standard is now beina used in this chance.

B. Unreviered Safety Ouestion IMPLEMENTATION OR PERFORMANCE OF DIE AC110N DESCRIBED IN THE EVALUA'IED DOCUMENT:

i 1. May increase the probability of occurrence of an accident previously evaluated O Yes m the SAR. o No 4

agig The oresence of minimal water and ans leakaae is adeaustely j included in current accident evaluations. These leakaos oeu, ways nose

' no new collective imonet to accident occurrence probability. and would

be tested collectively in other existina testina orocreins.

1 2. May increase the consequences of an accident previously evaluated in the SAR. O Yes a No i

j ggis The reevelina. treatment. or release of water and aas laekana is covered by current accident evaluations. The addition of leakane Dest

! vent and drain valves. or leakaae from aas systems. remains within

! Dostulated total leakaos amounts and chances no desian impact. No accident consecuences are imoacted by the oossibility of these leakaae i

Dathways.

3. May increase the probabdity of oxurrence of a malfunction of equipment O Yes imponent to safety previously evahlsted in the SAR. g No agig Leakaoe from vent and drain valves would be collected by liauid radwaste drains. to be treated with all other olant liauid waste. Gas leakaae from ces systems and associated oicina is treated with other i aas leakaae. Minimal leakaae in these areas will not adverselv imoact treatment system desian canabilities. nor will it imoact the orobability of an eauioment malfunction.

]

4. May increase the consequences of a malfunction of equipment wW4 to O Yes safety previously evaluated in the SAR. a No l

i i

4 i agig The consecuences of a malfunction of eauioment imoortant to safety remain the same as oreviousiv evaluated in the SAR. as any l leakaoe made oossible by this chance would fall within desian

{ soecifications of maximum olant leakaae. No new eauioment challenoes

) or malfunction consecuences are introduced l 5. May increase the possibility for an accident of a different type than any 0 Yes l previously evaluated in the SAR. [g] No

\

j ggig The oossibility of an accident of a different tvoe than oreviousiv i evaluated remains the same. as current accident evaluations consider l collective leakaos rates. which would include those oathways beina l l

l introduced in this chance. l

6. May create the possibility for a malfunction of equipment imped-4 to safety of O Yes a different type than any previously evaluated in the SAR. a No l

ggis The possibility of a malfunction of eauloment important to safety i remains the same as oreviousiv evaluated in the SAR. as current

! eauioment malfunction evaluations consider collective leakaae rates.

which would include any leakaoe from those oathways beina introduced

7. Will reduce the margm of safety as defined in the basis for any Technical O Yes j Specification. a No agis The marain of safety for Containment leakaae or oressure isolation l valve leakaoe as defined in the Technical Scecificetions Basis will not i be reduced. The testina orocrams which auantify the leakaae for the i soecified limits are not beino chanaed. Any leakaae detected by increasina the acceptance criteria from " eliminated" to "as low as j oractical levels" would be evaluated to determine if the limits werp, affected. Therefore. the marain of safety as defined in the Technical Soecifications will not be reduced 4

! IIL EnviroannestalEvaluation O Not apphcable per Eavkommensal Evaluatica

! Appucatmhty Review i

8 IMM2WDrrAT1oM OR PEaP0aMANCE OF THE ACTION DEaCalBED IN THE EVA111ATED DOCUMDrr*.

i

! A. Environmental Protection Plan i 1. Will require a change in the Enwonmental Protection Plan. O Yes

! o No I

i agig: The EPP does not address specific leakaoe oathways. and will not I reoutre chance, l

l 4

i 1

i l

]

i J

l B. Unreviewed Environmental Ouestion

1. Concerns a matter which may result in a significant increase in any adverse O Yes enwronmental impact previously evaluated in the Final Enwronmental a No

Statement (FES) as moddied by the NRC staffs tesumony to the Atomic Safety and Licensing Board (ASLB), supplements to the FES, environmental

) unpact appraisal, or in any decisions of the ASLB.

l Iggg The imonet oosed by emuent release is adeouately evaluated in i

the FES. and the oathways introduced by this chance will not affect the i criteria used in determinino any environmental imoact.

1

2. Concems a sqpuficant change in efBuents or power level. O Yes

a No l Igyg- Maintainino all aas and water leakaos to as low as oracticable i levels. as roouired by Tech Specs and NUREG - 0730. will not create a j sionificant chance in emuents; merely allow alternate oathways that

! orovide those emuents.

! 3. Concems a matter not pronously renewed and evaluated in the documents O Yes l spmAmt in H.B.1 above, which may have a sigmficant envirom=ntal g No j "

! B6313- Emuent leakaos is reviewed and evaluated in the aoolicable j documents.

I 1

i Signatures and Approvals l

i Eaua,ed %a caxMNAIOa/ DAM ddd i

Reviewed /Approvat

. / >'

l C' az@hvr.a/ DAR i

l Plant Safety Review Committee Review

/2,

! CHMaMAN, PSRC / DAu

/

i l

l 2

M - o 7 c2 - Pa, sE.

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM

l. Safety Evaluation Overview A Reference Data ORIGINATOR: D. R. Franklin DtPT/ SECT. P&SE EVAL. #: 97 oOS 7.Roo l DOCUMENT EVALUATED: 03-1-01-6. Reactor Vessel In-Service Leak Test. Also reference WO 144875. Wi&lR #2.

REFERENCES:

Technical Soecification Sections 3.4.6. 3.4.10. 3.10.1. 3.5.2. 6.4.2: FSAR Sections 5.2.5.2.k. 5.4.1.3. 5.4.7.1.2. 5.4.7.1.3. 5.4.7.2.7. 5.4.7.3.1. 5.4.5.4: GGNS-M 189.1. Pump and Valve Inservice Testina Procram: 10 CFR 50.55a(ct 10 CFR 50.

Aooendix A. GDC 55 FSAR CHANOi REQUIRED? O Yes Y No CR# N/A FS AR SECTIONS TO BE REVISED: N/A TRM CHANGE REQUIRED? O Yes [No l TECH. SPEC. CH ANGE REQUIRED? O Yes do CR# N/A IS THE VALIDITY OF THIS SAFETY EVALUATION DEPENDENT ON ANY CHANGES OTHER THAN OY I THE CHANGE BEING EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDITIONS)? O o IF YES TO THE LAST QUESTION. H AVE THE ORGANIZATIONS RESPONSIBLE FOR THOSE CH ANGES BEEN NOTIFIED?

Om THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED PRIOR TO IMPLEMENTING THIS CHANGE B. Executive Summary (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIPTION OF CHANGE, TEST OR EXPERIMENT: To allow oerformina the Reactor Vessel In-Service Leak Test with the disc removed from 1E12F050B. The disc will be removed and the retainina olate/cao assembly will be installed as normal. Valve 1E12F0538.

the other RCS PlV in series with 1E12F0500 in the 12"-DBB-68 line. will become a test boundary valve. The 3/4"088-66 test connection line from D88 68 will also be pressurized and valve 1E12F0588 will become a test boundary valve. Reference P&lD M-1085A.

REASON FOR CHANGE TEST OR EXPERIMENT: This will allow the Reactor Vessel in-Service Leak j Test to be performed while rework is beina performed on the 1E12F0508 disc cer 4 MNCR 0184-95. Reference WO 144875.

l

SAFETY EVAwATloN

SUMMARY

AND CONCWSIONS: Per Operability Revie; for MNCR 0184-95.

1E12F0500 is declared inoperable until reoaired and retested satisfactorily. This Safety Evaluation assumes 1E12F0508 is inocerable and reviews only oerformina the l

Reactor Vessel in-Service L eak Test with disc removed. With the disc removed but the retainer otate/cao assembly installed as normal. the cressure inteority of 1E12F050B is still intact. However. Without the disc. the normal test boundary must be moved upstream to the 1E12F0538 and 1E12F0588 valves. This chanae in test boundary does not create an unreviewed safety auestion as summarized in the Safety Evaluation Review.

II. Safety Evaluation O Not applicable per Safety Evaluation Applicability Review A. Technical Soecifications

1. Implementation or performance of the action described in the evaluated 0 Yes document will require a change to the GGNS Unit 1 Technical No Specificatior.s.

gasgi: The proposed chanae to perform the Reactor Vessel In-Service Leak Test

j. with the disc removed from check valve 1E12F0500 does not represent a chance to the Technical Specifications. Check valve 1E12F050B is used to meet reauired actions for Tech Soec 3.4.6. however. this Tech Soec is not

aoolicable in Mode 4. Tech Spec Sections 3.4.10. RHR Shutdown Coolina l Reauirements (also 3.10.11. and 3.5.2. ECCS Shutdown Limitations. are not I beina chanaed and still acolv. Tech Soec 6.4.2 also does not chance and 1E12F0508 will still be maintained in accordance with the Inservice Testina Proaram.

B. Unreviewed Safety Ouestion

[MPLEMENTATION OR PERFORMANCE OF UIE ACTION DESCRIBED IN THE EVALUATED DOCUMENT

1. Niay increase the probability of occurrence of an accident previously evaluateo O Yes in the SAR. [No DAs.gi Globe valve 1E12F053B will become a boundary valve for the Reactor Vessel In-Service Test. This valve alreadv serves as one of the two RCS PlVs. In series with 1E12F0508. Valve 1E12F0538 has been shown to be basically leak tiaht (3789 ml/ min allowable leak rate. 50 ml/ min actual) at a test differential pressure of 1050 osid oer LLRT WO 140719. Also. 3/4"-DBB-66 and valve 1E12F0588 will become cart of the test boundarv. The desian ratina of these components per MS-03 is more than adeauste for the Reactor Vessel in-Service Leak Test conditions. There is another closed valve

. downstream of 1E12F0588 which orovides an additional test isolation backuo to 1E12F0588. This is the 1E12F059B where the line terminates with a l oluaaed 1-1/2" hose connection. For the reasons stated above. this chance

! does not increase the probability of occurrence of an accident previously l evaluated in the SAR.

2 hiay increase the consequences of an accident previously evaluated in the S AR. Yes No

. _ _ _ . . . . . . _. - _ _ _ . _ _ _ _ _ _ _ _ _ __ _ _ _ . _ .m

,7 l .

I i-T 4

p_ jig Performina the Reactor Vessel In-Service Leak Test with the disc removed from check valve 1E12F0508 and the test boundary chanced to k

1E12F053B and 1E12F058B does not chanae the intent or method of the

Leak Test. The conditions (Dressure and temDerature) and flow cath for the j test remain basically the same. Valve 1E12F050B retains its pressure intearity I with the disc removed since the cao assembly is a desianed oressure boundary for the valve. The desian ratinas of the new test boundaries.

! 1E12F0538.1E12F0588. and associated oicina. are adeauate for the test I conditions. This chance does not increase the consecuences of an accident

! oreviously evaluated in the SAR.

, 3. May increase the probability of occurrence of a malfunction of equipment O Yes '

I important to safety previously evaluated in the S AR. 7 No i agg: Performina the Reactor Vessel in-Service Leak Test with the disc l

! removed from check valve 1E12F0508 and the test boundary chanced to i

1E12F0538 and 1E12F0588 does not chance the intent or method of the Leak Test. The 1E12F0538 and 1E12F050B which will be subiected to reactor oressure will be in a passive sta's isolated in the closed position durina the test. The 1E12F053B was local leak rate tested and was within the

! allowable leakaae rate of 1 aom oer Tech Soec 3.4.6.1. The function of the j 1E12F0538 valve is to orotect the low oressure RHR B oicina from beina over i pressurized. The desian ratinas of the new test boundaries exceed test I

conditions. For these reasons this chance does not increase the probability of j occurrence of a malfunction of eauioment.9'

] 4. May increase the consequences of a malfunction of equipment important to O Jes j

safety previously evaluated in the S AR. 7 No 3

ggg: Pgtformina the Reactor Vessel in Service Leak Test with the disc

! removed from check valve 1E12F050B and the test boundary chanaed to j 1E12F0538 and 1E12F0588 does not chance the intent or method of the j Leak Test. The conditions and flow cath for the test remain basically the j same. The low oressure oicina beina orotected via 1E12F0538 has not been

affected by this chance. therefore, the consecuences of the failure of

{ 1E12F053B to orotect the low oressurin oicina remains the same.

i 5 May increase the possibility for an accident of a different type than any 0 Yes l previously evaluated in the S AR. 7 ko BASIS: Performina the Reactor Vessel In Service Leak Test with the disc removed from check valve 1E12F0508 and the test boundary chanced to l 1E12F0538 and 1E12F0588 does not chance the intent or method of the Leak Test. The conditions (temoerature and oressure) and flow Dath for the test remain basically the same. This chance does not increase the oossibility for an accident of a different tvoe than any oreviousiv evaluated in the SAR.

6 May create the possibility for a malfunction of equipment important to safety of C Yes a different type than any previously evaluated in the S AR. y'No

. j l

R6gg Performina this Leak Test with the disc removed from check valve 1E12F050B and the test boundary chanced to 1E12F0538 and 1E12F058B does not chance the intent or method of the test. The conditions and flow path for the test remain basically the same. Subiectina the 1E12F0538 to test pressure is well within the desian and safety function of the valve to orotect the low oressure Dioina and removina the disc from 1E12F050B does not affect the cressure boundarv of the component. therefore.this chance does not create the possibility for a malfunction of eauioment lof a different tvoe than any previously evaluated in the SAR.

l 7 Will reduce the margin of safety as defined in the basis for any Technical O Yes Specification. do BAgg Performina this Leak Test with the disc removed from check valve 1E12F0508 and the test boundary chanced to 1E12F0538 and 1E12F058B i

does not chanae the intent or method of the test. The conditions and flow oath for the test remain basicady the same. The marain of safety as defined in Tech Spec 3.4.6 reauires RCS PlVs (1E12F053B & 1E12F0508) be operable in Modes 1. 2. and 3 and have leakaae s; 1 aom at 1050110 osia. The l

Reactor Vessel In-Service Leak Test is performed in Mode 4. In Modes 1. 2 l and 3. the loss of 1E12F050B would reauire isolation of the hiah oressure Dioina from the low oressure Dioina by isolatina 1E12F0538. This intent is beina met by the prooosed test boundaries. therefore this chanas does not l

reduce the marain of safety for any Tech Spec.

lit Environmental Evaluation O Not applicable per Environmental Evaluation Applicability Review IMPLEMENTAT!oN oR PERFORMANCE oF THE ACTloN DESCRIBED IN THE EVALUATED DOCUMENT:

A. Environmental Protection Plan I Will require a change in the Environmental Protection Plan. O Yes MO i gas _1g Performina the Reactor Vessel In-Service Leak Test with the disc l removed from check valve 1E12F050B and the test boundary chanced to  ;

1E12F0538 and 1E12F058B does not reauire a chance to the Environmental  ;

Protection Plan, B. Unreviewed Environmental Ouestion

1. Concems a maner which may result in a significant increase in any adverse O Yes environmental impact previously evaluated in the Final Environmental po Statement (FES) as modified by the NRC staffs testimony to the Atomic Safety and Licensing Board (ASLB), supplements to the FES, environmental j impact appraisal, or in any decisions of the ASLB.

gagg Performina this Leak Test with the disc removed from check valve 1E12F050B and the test boundary chanaed to 1E12F0538 and 1E12F0588 will not result in a sianificant increase in any adverse environmental imoact.

i j

r t

l l

l l l

l l 2. Concems a significant change in efBuents or power level.. - O Yes g/No m

B P_grformina this Leak Test with the disc removed from check valve j 1E12F050B and the test boundary chanaed to 1E12F0538 and 1E12F058B j l does not cha3Je the intent or method of the test. The conditions and flow l cath for the test remain basically the same. No chanae in effluents or power level will be reauired.

3. Concems a matter not previously reviewed and evaluated in the documents I specified in II.B.1 above, which may have a significant environmental O[Yes No impact.

BAEls Performina the Reactor Vessel in-Service Leak Test with the disc removed from check valve 1E12F0508 and the test boundary chanced to 1E12F0538 and 1E12F058B will not result in a sianificant increase in any adverse environmental imoact.

Signatures and Approvals l Evaluated: [hE f, kM DATE g-ga. 7 f l (ORio!NAT i

Reviewed / Approved:

A  % [

7 REVIEWER (dad

, Plant Safety Review Committee Review l

S/WM$$_

/ CHAIRMAN, PSRC / DATE rbzNr

• t f-

l 1

M - CM - F%St 5 6. -

GRAND GULF NUCLEAR STATION UNrr 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 1 of 16 I. Safety Evaluation Overview A. Reference Data ORIGINATOR: Alan J. Malone DEPT / SECT: P&SE EVAL. #: 95-0059-R00 DOCUMENT EVALUATED: Chance to Technical Reauirements Manual (TRM) Tables TR3.6.1.3.1-1 (Primary Containment isolation Valves) and TR3.6.4.21 (Secondarv Containment isolation Valves)

REFERENCES:

FSAR Section 6.2 & Chapter 15: EER 94/6182. with FFRR dated 5/12/95: AECM.

84/0330: MAEC-89/0124: Soecification GGNS-M-189.1: 10 CFR 50.55aff): ASME Code.Section XI FSAR CHANGE REQUIRED 7 E Yes O No CR#

1 FSAR SECTIONS TO BE REVISED: N/A l TRM CHANGE REQUIRED? E Yes O No i TECH. SPEC. CHANGE REQUIRED 7 0 Yes 3 No CR# N/A IS THE VALIDirY OF THIS SAFETY EVALUATION DEPENDENT ON ANY CHANGE 8 OTHER MN O Yes THE CHANGE BEING EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDirIONS)? E No IF YES TO THE LAST QUES 110N, HAVE THE ORGANIZATIONS RESPONSIBLE FOR THOSE CHANGES BEEN NOTIFIED 7 Om THE RESPONSIBLE ORGANIZATIONS MU3T BE NOTIFIED PRIOR TO IMPLEMENTING THIS CHANOE.

B. Executive Summarv (AIJO SERVES AS INPtfr TO NRC

SUMMARY

REPORT)

BRIEF DESCRIFTION OF CHANGE, TEST OR EXPERIMENT: This Channe revises the maximum allowable stroke times for various crimary and secondary containment isolation valves listed in Technical Reautrements Manual (TRM) Tables TR3.6.1.3.1-1 (Primary Containment Isolation Valves) and TR3.6.4.2-1 (maeandary Containment Isolation Valves). It does not revise times for drvwell isolation valves or any other valves.

REASON FOR CHANOE, TEST OR EXPERIMENT: Enaineerina Evaluation Recott (EER) No. 94/6182 was initiated beemu** SGedi"i&UGG GGNS-M-189.1. GGNS Unit 1 Pumn and Valve inservice Testina Proaram. A_aaandix A. *Ba==s for Maximum Stroke Times of Power Actuated Valves.'

imolies that numerous valves in the inservice testina (IST) crocram have analvtically based maximum stroke time limits. Althouah some of the valves listed in Anoendix A have caraoraohs in the Safety Analysis Reoort (SAR) listed. for which exclicit time limits are niven.

many of the valves are identified only as havina limits in either GGNS Technical Snecifications Table 3.6.4-1 for ortmary containment isolation valves or Table 3.6.6.2-1 for secondary containment isolation valves. Tables 3.6.4-1 and 3.6.6.21 have been relocated to the Technical Reauirements Manual (TRM) as Tables TR3.6.1.3.1 1 and TR3.6.4.21. resoectively.

H:\AMALONE\WINWORD\VLVTIMEl. DOC Revised: June 15,1995

GRAND GULF NUCLEAR STATION UNIT 1 l' CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 2 of 16 B. Executive Summary (Continued):

REASON FOR CHANGE. TEST OR EXPERIMENT (CONTINUED): In " Detailed Descriotion of Problem.' the EER noted that many of the valves listed in TRM Tables TR3.6.1.3.1-1 and TR3.6.4.2-1 do not ,

have analvtical safety bases for the listed maximum allowable stroke times. In many cases. the  !

listed maximum allowable stroke times for these valves had been arbitrarily determined based on either orevious performance data or on information sunolled by the manufacturer based on oerformance of tvoical valves of the same model/ size.

Due to maintenance. desian chances. and/or normal wear. Derformance of some valves has mooroached or exceeded limits in the TRM tables. resultina in additional man-hours and man-rem exDosure to adjust their oerformance to aaain be within TRM limits.

In a resoonse to EER 94/6182. Nuclear Plant Enaineerina (NPE) aareed that there were no technical bases for the stroke times for many of the valves in the TRM tables.

l I

In the resoonse. NPE identified. In an attached Table 1. valves with soecific stroke time limits which were identified in the SAR. NPE also identified. In an attached Table 2. valves withond specific limits dearly identified in the SAR but for which certain imolied limits could be identifled from the SAR haead on certain analyzed events in which their coaltions were important. An examole is the function of the Standbv Gas Treatment Svstem (SGTS) in drawinn down the Auxiliary Buildina atmosohere. which reauires that secondary containment isolation valves must I be closed within the time limit for the SGTS to perform its function.

NPE's Resoonse identified numerous valves with orimary containment and secondarv i containment isolation functions for which the maximum stroke time limits in the TRM were shorter than could be iustified by their ortmarv or secondary containment isolation function. The valves considered in the TRM chances for which this Safetv Evaluation is written are crimary containmerit isolation valves with indirect oathways for leakane. for which the NPE Resoonse identified a 60-second dosino time limit. and secondary containment isolation valves. for which -

the NPE Resoonse identified a 120-second dosina time limit. I NPE'i_Resoonse also identified drvwell isolation valves whose closure time limits were not suooorted by the safety analysis. Additional analysis is recuired to identify the time limits for drvwell isolation: therefore. these valves are not induded in these TRM chances.

SAFETY EVAWATION

SUMMARY

AND CONCWSIONS: The Safety Evaluation condudes that increasina the maximum allowable isolation times for the primary and secondary containment isolation valves identified in th::: TRM chanaes to the isolation times enaaarted by the SAR safety analyses does not incr**** the Drobability or consecuences of an accident or malfunction of eauioment important to safety. The times currently listed for these valves in the TRM tables Mpntified above do not have any safety sionificance to the olant.

Performance of inservice stroke testina of these valves will continue in accordance with the GGNS Unit 1 Pumo & Valve Inservice Testina Procram (Soecification GGNS-M-189.1). as reauired by ASME Boiler and Pressure Vessel Code (ASME Code).Section XI ' Inservice insoection." which is referenced in and recuired by GGNS Technical Snecifications 5.5.6 and IBM 7.6.3.3. This oractice will minimize the 0055ibi;ity of operatina w4h dearaded components. which will prevent an increase in the probability of occurreice and the consecuences of malfunctionino eauioment.

H:\AMALONE\WINWORD\VLVTIMEl. DOC Revised: Jure 8,1995

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 3 of 16 II. Safety Evaluation O Not applicable per Safety Evaluation Applicability Review A. Technical Soecifications

1. Implementation or performance of the action described in the evaluated O Yes document will require a change to the GGNS Unit 1 Technical Specifications. g No EMH: These chances do not chance the reauirements in GGNS Technical Snecifications SR 3.6.1.3.4 and SR 3.6.4.2.2 to perform Inservice Testina (IST) exercisina tests on the valves listed in Technical Reauirements Manual (TRM) Tables TR3.6.1.3.1-1 (Primary Containment Isolation Valves) and TR3.6.4.2-1 (Secondary Containment isolation Valves). nor do they chance the actions to be takan if a valve fails to meet the maximum allowable stroke time listed in the TRM tables siitified above.

In many e===s. the listed maximum allowable stroke times for these valves had been artWtrartiv determined hamad on either orevious cerformance data or on information supolled bv the manufacturer ha*=d on cerformance of tvalca! valves of the same madaValva.

These chances do not chance any maximum allowable stroke times exolicitiv stated in any GGNS Technical Specification.

Neither the IST orocram raa@ements stated in GGNS Technical maaefflemelan 5.5.6 and TRM 7 6 3.3. nor the underivina IST Dissisiii soecified in ASME Boiler and Pressure Vessel Code (ASME Code).Section XI. " Inservice insoection.' Subsection IWV. " Inservice Testina of Valves in Nuclear Power Plants." 1960 Edition with Winter.1960. Addenda. are beino chanced.

The maximum allowable stroke times currentiv in the TRM tables id->U,sd above were oreviousiv listed in GGNS Technical Soecification Tables 3.6.4-1 and 3.6.6.21. but were relocated to the TRM in Amendment 102: therefore. chances to these tables do not reauire a Technical Soecification chanoe.

B. Unreviewed Safety Ouestion IMPLEMENTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT:

1. May increase the probability of occurrence of an accident previously evaluated O Yes in the S AR. g No ggg: These chances do not affect the orobability of occurrence of any accident evaluated in the SAR because they do not make any chances to the ohvsical clant or any oceratino orocedures identified in the SAR or Technical Soecifications. The chanaes do not chance the actual ooeratino characteristics of any valve. nor do they affect the valve's nerformance in any way that could lead to an accident occurrina. No ohvsical or orocedural chances in the olant are beino made. These valves are crimary and secondary containment isolation valves which are reauired to close to isolate primarv or secondary containment under certain accident conditions.

(Some of these valves may also have safety functions to open. but these TRM chances do not affect the openina time limits.)

H:\AMALONE\WINWORD\VLVITMEl. DOC Revised: June 8,1995 I

1 .

i i

l GRAND GULF NUCLEAR STATION UNIT 1

! CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM i

! PAGE 4 of 14 1

i 1. May increase the probability of occurrence of an accident previously evaluated in the SAR (Continued):

?

I BASIS (CONTINUED): Most of these valves are normally open durina oower operation and are

, reauired to close on orimary and/or secondary containment isolation sianals: however. some of

) them may be closed or opened infrecuently during power operation.

To affect the orobeNiity of occurrence of an accMeat would reauire that one or more of these i valves were in a an=dion different from that analyzed in the accident scenario at the onset of

the accident seauence. These TRM chanaes do not make any chances to the operatino orocedures or ohvsical chances in the olant: therefore. thev will not affect the initial cosition of any valve at the onset of any accident.

(Failure of a valve to raeaand to an isolation sianal nonerated durina the accident secuence or j failure of the valve to do** within the maximum closure time analyzed for the accident could j certainly affect the consecuences of the accident. but thev do not affect the probabilltv of

occurrence. In eddMion. a valve which eL4a-!=alatad due to a malfunce_lm of its control dr=#

j could affed the orahaNimv of neemtence of an meddaest: however. such malfunctions are bevond the menaa of this Safety Eve!"*'lan. since the crocosed TRM channes do not affed

!. such malfunctions.)

l The orah=N8mv of occurrence of an accident is affected bv extemal factors not within the control i Af the Owner or Operator. as well as bv intamal desian factors within the control of the Owner or j Ooerator. However. In order to affect the oreh=Ni#v of accurrence of an acddent. the Owner l would have to chance the initial cosition of one or more of the valves identified in the TRM tables identified above. No chances to the operatino orocedures to chance the operatina

oositions of any valve are beina made. and no chvsical chances in the olant are beina made.

It is possible that the orah=N!ity of occurrence of an accident could be increased if a valve's j stroke time were decreenad to the coint that severe water hammer was aenerated by the

closure of the valve. None of these TRM chances reoresent decreases in the stroke time limits.

l they are all increases in the limits.

t

- Since the orooosed chances do not chance the operatina nosition of any valve. they do not j affect the orobability of occurrence of any accident analyzed in the SAR.

I a

2. May increase the consequences of an accident previously evaluated in the SAR. Yes

! r No i

gis,Is:

s These chances do not affect the consecuences of any accident evaluated in the SAR.

- unless they would cause a reauired valve to fail to be in its safety position when it was reouired

! to be. or unless their soeed of closure caused additional oroblems.

) These chances do not make any chances to the ohvsical olant or any operatino or accident

! resoonse orocedures identified in the SAR or Technical Soecifications: therefore. the ONLY j way these chances could affect the consecuences of an accident oreviousiv evaluated in the i

SAR is by increasina its allowable i<a!*' ion time to the extent that the valve could still be j closina when it is reauired to be fully cloaad or by causina damane to the olant or oloina system i by its closina speed.

H

\AMALONE\WINWORD\VLVTIMEl. DOC Revised: June 8,1995 i

0 GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 5 of 16

2. May increase the consequences of an accident previously evaluated in the S AR (Continued):

BASIS (CONTINUED): All of the valves addressed in these channes are either orimary containment isolation valves listed in TRM Table TR3.6.1.3.1-1 or secondary containment isolation valves listed in TRM Table TR3.6.4.21. No drywell isolation valves are included in the proposed TRM Chances.

All of the valves addressed in these chances are desianed to automatically isolate (move to the fully closed oosition) in response to an isolation sianal. None of the isolation sianals are affected by these chanaes. nor are the actions of the valves to dose in resconse to the sianals beina chanced. Althouah some of the valves may be canable of beina recoened. or even stooned from fully closina by Ooerator action. such actions are not within the scoes of the chances beina made to the TRM tables listed in Section 1.A.

The ONLY oossible wavs that these chanaes could affect the consecuences of any accident l would be if one or more valves failed to isolate to its/their safety position within the time allowed I by the accident analysis. or if the the dosure of the valves in.too fast a time caused water hammer that damaaed the osoina or succorts further.

It is possible that the orobability of occurrence of an accadent could be increased if a valve's stroke time were decreased to the point that severe water hammer was aenerated by the closure of the valve. None of these TRM chances represent decreases in the stroke time limits:

they are all increases in the limits.

The accident analyses in the SAR that reauire crimarv and secondary containment as oart of the accident response are Loss of Coolant Accidents (LOCAs). which are described in SAR l Section 6 2. The ament analyses assume that the primary and secondary containments are )

isolated within soecafled times after the start of the acddents in order to Drevent radioactive  !

releases to the environment in excess of the auidelines established in 10 CFR 100. Althouah some other accidents analyzed in SAR Chapter 15 expect or reauire isolation of some ortmary and/or secondary containment isolation valves (such as Main Steam isolation Valves or other air-ooerated valves with fail-safe isolation functions). LOCAs are the limitina accidents for orimary and secondary containment isolation valve closure.

As discussed in the Nuclear Plant Enaineerina (NPE) Response to Enaineerina Evaluation '

Reauest 94/6182. orimary containment aM secondary containment isolation valves may have either exolicit or imolicit analvtical closure time limits based on the LOCA analyses. as well as for a variety of more limited events which recuire crimary and/or secondary containmerit isolation. which are descnbod in Chanter 15.

The NPE Response provided two lists of valves: Thon with exoticit analytical closure time limits (soecific times stated in the SAR). which were described in Table 1 in the NPE Resoonse.

and those without exDlicit limits but for which isolation time limits are imolied (imolidt limits) from other events (e a.. SGTS drawdown) described in the SAR. which were described in Table 2 in the NPE Resoonse.

None of the valves included in this TRM chance have exolicit limits listed in Table 1. They are all specifically identified in Table 2 as either odmary containment isolation valves with indirect leak oaths (for which the maximum analytical closure time is 60 seconds. as soecified on Pace 8 of the NPE Resoonse) or secondary containment isolation valves (for which the maximum analvtical closure time is 120 seconds. as also soecified on Paae 8 of the NPE Resconse).

H:\AMALONE\WTNWORD\VLVIIMEl. DOC Revised: June 8,1995

F l GRAND GULF NUCLEAR STATION UNIT 1  ;

CHANGEsfTESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM l 4

PAGE 6 of 16 )

2. May increase the consequences of an accident previously evaluated in the S AR (Continued):

BASIS (CONTINUED): The oostulated radioactive release from the olant will be limit 6d to less than the cuidelines established in 10 CFR 100 if all orimary and secondary oenetrations are isolated in less than these analyzed times.

None of the valves induded in this TRM chanae are drvwell isolation valves. for which Table 2 of the NPE Response indicates additional analysis is reauired.

The conseauences of an accident would be adverselv affected if a valve malfunction caused by these TRM chances orevented the valve from closina within its reouired time. The orobability of occurrence and consecuences of eauinment malfunctions are addressed in Sections ll.B.3 and ll.B.4 of this Safetv Evaluation. These analyses condude that neither the probability of occurrence nor consecuences of eauioment malfunctions would be increased by these TRM chances.

The EER Response from NPE specifies that primary containment isolation velves rqqst dose within 60 seconds to isolate oenetrations which do not Dmvide a direct Dathway betWeen containment and anvinarv buildina atmosoheres. unless lonner time limits have been soscifically accroved by the Nuclear Reaulatory Commission (NRC). All of the crimary containment valves identified in these TRM chances currentiv close in less than 30 seconds i (half of the 60 second limiti. excent for four valves: 1E51F031.1G41F028.1G41F029.and 1G41F044. all four of which close within 45 seconds.

These four valves are motor-ocerated valves. which are hiahly credictable in their operatino-time performance and have been shown over many vaars of ooerational performance to close  ;

reliably within a very narrow time band. Therefore. all of the ortmary containment isolatig.g valves in these TRM chanaes can be exnected to continue to dose within the 60 second limit.

The EER Resoonse from NPE also snedfies that secondary containment isolation valves must close within 120 seconds unless lonner time limits have been soecifically accroved by the Nuclear Reaulatory Commission (NRC). The 120 second isolation time is soecified in SAB Section 6.2.3.1.1. which enstan that the secondary containment. In coniunction with the Stacq,g Gas Treatment System (SGTS). is reauired to maintain a 1/4-inch water anuoe (w.a.) neosqyg oressure in the Auvillary Buildina within 120 seconds after actuation. As described in FSA3 Section 6 2.3.2. the SGTS is caaahia of maintainina the secondary containment neaative ora =*"re in =ama of the failure of all nonaualified lines 2 ina.:hes and smaller or with the failurej a_sJnnle_norlianlated_line_as_ lame as 4 inches.

All of the secondary containment valves identified in these TRM channes currentiv close in less than 60 seconds excent for eioht Plant Service Water (PSW) System valves: 1P44F116 throuah 1P44F123.

The secondary containment valves (other than the siaht PSW valves listed above) would not sianificantiv affect the drawdown rate of the SGTS unless their operatina times ware Increased to near the orocosed limit of 120 seconds. They should not affect the ability of the SGTS to draw down the Auxiliary Buildina atmosohere for the followino reasons:

1. They are easily caoable of clotna in less than 60 seconds and there are no olans to increase their normal closina times, H:\AMALONE\WINWORDWLVTIMEl. DOC Revised: June 8.1995

Q GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMEKIS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 7 of 16

2. May increase the consequences of an accident previously eveluated in the S AR (Continued):

BASIS (CONTINUED):

2. The 120 second limit is intended to be used oniv for determinina whether or not a valve should be dedared Inocerable. It will not be the sole criterion for deturminina whether or not valve performance is Acceptable in accordance with ASME Code.Section XI.

3. ASME Code.Section XI. soecifies that corrective adion is reouired if a valve exceeds its limitina value of full stroke time. In accordance with the GGNS Unit 1 Pumo and Valve IST Program. Specification GGNS-M-1891. that limitina value of full-stroke time is established as a multiole of the averaae stroke time performance of the valve. Currentiv. the multiole is 1.35 for valves with full stroke times areater than 10 seconds.

4. Since they close in less than 60 seconds. their limitina values of closina time are limited to less than 1.35 x 60 seconds. or less than 81 seconds. This is well below the 120 second Ilmt j for Auxiliary Buildina drawdown.

Therefore. all of the secondary containment isolation valves identified in these TRM chances.

> except oossibiv the eloht PSW valves identifled above. will continue to dose well within the 120

{' second limit. even with tilase TRM channes. and will not affed the consecuences of an accident reauinna secondary containment intearity.

l The eloht PSW valves identified above are the oniv .acondary containment isolation valves j which take lonaer than 60 seconds to close. These valves currentiv close within a relativelv

( narrow band of 90 to 100 seconds. The 90 second limit at the low and of the band is soecified j to minimize water hammer when thev da=. These valves are aaniaad with air aduators j which fait closed on loss of air pressure. Dependina on their conditions. one or more of these j valves could fail to close in time to normit the Standbv Gas Treatment Svstem (SGTS) to draw down the Auxiliary Buildina atmosphere to 1/4 6nch w.a. within 120 seconds.

l It is unlikely that inctsasina the dosina time limit for one or more of the elaht PSW Valves identified above to 170 seconds would cause the SGTS to fail to draw down the Aux 13[y j Buildina atmosohere to 1/4 inch w,0, within 120 seconds for the followina reasons:

f 1. They are 611 in water-filled penetrations: therefore. thev are not evnaud to Auxjggy Buildino atmosphoe at any time. except as a possible result of nine ruoture inside the Auxiliarv l

Buildina.

l I 2. Thev all have air-ocerated fail-safe aduators. which cause the valves to automatically

{ close on loss of air oressure or electrical oower. They do not reauire oower to dose and.

! therefore. except due to component malfunction. are costulated to dose on a secondary containment isolation sianal. Since more than one indeoendent comoonent failure in addition to

) the initiatina event is not a credible occurrence. all but one of these siaht valves can be i costulated to close durina the accident.

3. All of the PSW Dioina isolated by thr.se valves have redundant isolation valves: that is.

{ there are two of these valves on each_PSW Dioe which nonetrates the Auxiliary Buildina. In l j accordance with Point 2 above. wNch mstulates no more than one indeoendent comDonent j failure in addition to the initiatino event. each PSW Diot can b6 reliablV eXDected to ta isolated i by at least one valve durina an accident reauirina secondary containment. l

\ l l

i H:\AMALONE\WINWORD\VLVTIMEl. DOC Revisert June 8,1995 i____-_.____----___-_---______._ _ - -- - -__. - -

\

l*

• i

! l i  !

I- l

! GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 8 of 16 i .

j 2. May increase the consequences of an accident previously evaluated in the SAR (Continued):

1 j sASIS (CONTINUEDh j s  !

4. The PSW oinina ou+ side the Auxiliary Buildina is buried several feet underoround.

l Althouah the cioina outside the Auxiliary Buildina isolation boundaN is not safetv related. the

( interior of the nice would nr1 be exoosed to the atmosohere. even if it ructured. until the water j in the Dios drainglqpt gnd a cath was opened up to the atmosohere.

l 5. Since it is buried. the water cannot readily drain out of the oice. Followina an accident.

{ the water would be most likely to drain out the ends (at the radial wells or inside the Auxiliary Buildinal This drainina would be a slow orocess for the reasons discussed in the followina l

! O l 6. At the radial wells. check valves at the oumos would retard drainaae back throuah the i oumo. even if the pumos were not runnino. Even if the oicina in the Auxiliary Buildina ructured. .

I

[ it would take a lona time (several minutes) for the DIDina between the Auxiliary Building and the j radial well oumos (over a mile of Dece) to drain so that the atmoscheres would be exoosed.

l 7. Even if the osoina outside the Auxiliary Buildina ructured. the oscina inside the Auxiliary i Buildina would also have to ructure in order to expose the isolation valve discs to the Auxiliarv l Buildina atmosDhare ...While this Dininoja_noLaafety-reta;9d. It is constructed to the Power i Ploina Code. ASME B31.1. and. therefore. can be exneced to remain reasonably intact.

j Althouah nortions raay crack durina a seismic event. it can reascnably be awaesari. based on industry experience during other seismic events with cloina constructed to the same Code. to

! remain in one piece and sunoorted. This means the water would not instantaneousiv run out of

! the oice and exoose the isolation valve discs. It would. Instead. Drobably maintain the Dioina water-filled for most of the time that the valves were closina.

8. Even if the oicina outside the Auxiliary Buildina ructured and the oinina inside the j Auxiliary Buildina ructured. the valve discs would not be exDosed to atmospheres inside and

outside the Auxiliary Buildina until water drained out of the nice sufficientiv to expose a leakaae oathway. Durina the first few minutes after an accident. there is no sianificant oressure inside Dnmary containment. therefore. assumina a containment oenetration were open to atmosphere.

there would not be any significant oressure in the Auxiliary Buildina. Some sinnificant drivina oressure inside the Auxiliary Buildina would be reouired in order to push the water In the PSW j oicina old of the underaround oicina. Without a drivina oressure head. the water would drain j out slowlv. but it would take sionificantiv lonaer that two minutes to drain the cioina completely.

l Based on the above. even if the maximum allowable closina times for the elaht PSW valves

! listed above were increased to 120 seconds. it would not affect the ability of the SGTS to draw j down the Auxiliary Buildina atmosohere to 1/4-inch w.a.

t

! Since the orooosed chances do not exceed the limits in Table 2 of the NPE Resoonse and thev 1

would not increase the DrObabilitV of occurrence or consecuences of Gouioment malfunClions.

j they do not affect the consecuences of any accident analyzed in the SARs i

i I H:\AMALONE\WINWORDWLVTIMEl. DOC Revised: June 8,1995 i

i

-- . -. _ - .- - _..- - .-.- - - .~_..- . ___ - _ - ._

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 9 of 16

3. May increase the probability of occurrence of a malfunction of equipment O Yes important to safety previously evaluated in the S AR. g No ugls: These chanaes do not affect the probability of occurrence of a mayunction of any eauipment important to safelv Dreviousiv evaluated in the SAR because the'? do not make any chances to the ohvsical plant or any operatino or accident response nocedures identified in the SAR or Technical Scecinc&tions and thev do not affed omorams aNadv in olace to orevent the valves from operatina under dearaded conditions.

The probeNiity of a malfunction of eauioment important to safety is affeded by channes in the  !

conditions under which it ocerates (environment. nower sunolv inouts. mechanical and ePgggg[

condit $n. etc.). none of which are beina chanced by these channes. No chvsical or procedural chances in the olant are beina made.

The orohaN!ity of malfunction of eauloment imoortant to safety may also be increased by operatina the eauipment under daaraded conditions for lona ceriods. Such dearaded conditions may inc8ude under voMaam. under-lubrie* Hon. rMued air *"aaay and/or oressure. etc. One of gig svmotoms of operation under dearaded conditions. at least for some valves. is an increase s in the me==ured stroke times. To the extent that increasina the allowable maximum stroke time limit may allow a valve to be ocerated under dearaded conditions for extended oeriods. these incree=== in the limits miaht increase the probability of a malfunction of the affected valves, if they were the only aoolicable limits.

However. the TRM maximum allowable stroke time limit is not 1}g.oniv stroke time limit for 8

these valves. All of these valves are also reauired to be exerci? and stroke time tested in accordance with ASME Code.Section XI. Subsection IWV. as reauired by and soecified in GGNS Technical Soecification 5.5.6. TRM 7.6.3.3. and 10 CFR 50.55am.

The Nuclear Reaulatory Commission (NRC) orovided auidance and clarification on the performance of IST on oumos and valves in accordance with Section XI with the issuance of Generic Letter (GL) 89-04. " Guidance on Develocina Accentable Inservice Testina Procrams" (MAEC-89/0124). In Attachment 1 to GL 89-04. Section 5. entitled "Limitina Values of Full-Stroke Times for Power Ooerated Valves.' the NRC stated the followina auidance:

"The ournaam of the limitina value of full-stroke time is to establish a value for takina corrective adion on a dearaded valve before the valve reaches the point where there is a hiah probability of failure to perform its safety function if called upon. The NRC has. therefore. established the auidelines described below reaardina limitina values of full-stroke time for power ocerated valves.

'The limitina va!ue of full-stroke time should be based on the valve reference or averaae strake time of a valve when it is known to be in nood condition and operatino Dronerly. The limitina value should be a reasonable deviation from this referents stroke time based on the valve size. valve tvos. and aduator tVoe. The deviation should not be so restridive that it results in a valve beina declared inonerable due to reasonable stroke time variations. However. the deviation used to establish the limit should be such that corrective action would be taken for a valve that may not perform its intended fundion."

H:\AMALONE\WINWORD\VLVTIME1. DOC Revised: June 8,1995

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUAMON FORM PAGE 10 of 16 I

3. May increase the probability of occurrence of a malfunction of equipment importmt to safety previously evaluated in the S AR (Continued):

BASIS (CONTINUED): The GGNS Unit 1 Pumo and Valve Inservice Testina Procram. Soecification GGNS-M-189.1. "GGNS Unit 1 Pumo and Valve Inservice Testina Proaram." has been wntten to comolv with ASME Code.Section XI. reouirements. includina scadfic reauirements in GGNS  !

Technical Soecifications and the TRM. The NRC auideline stated above is imolemented in Soecification GGNS-M-189.1. Anoendix C. entitled ' Calculation of IST Maximum Stroke Times.* Anoendix C describes the method of calculatina a maximum stroke time based on the performance of the valve. Paraaraoh 7.1.3 of Accendix C incoroorstes the auldance in the section of GL 89-04 auoted above.

In estance. Aooendix C reauires that. if a limit calculated based on the nerformance of the i valve is less than an analvtical limit in the SAR. Technical Scecinc Gons or TRM. the lower limit shall be the limit for takina corrective action. before the valve performance dearades to the 1 Doint that a malfunction could occur. If the analvtical limit is less. the analvtical limit is the limit for takina corredive action. Either limit minimizes the orobability of operatina the olant with valves in dearaded condition.

In addition. it is na==Ne that the orobability of a component malfundion could be increased it a valve's stroke time were decraa*M to the coint that severe water hammer was aenerated by the closure of the valve. None of these TRM channes reoresent decreases in the stroke time limits: thev are all increases in the limits.

Therefore. the orobability of occurrence of a malfundion of eauipment important to safety previousiv evaluated in the SAR will not be increaseq

4. May increase the consequences of a malfunction of equipment important to O Yes safety previously evaluated in the S AR. g No

) aASIS: These chances do not affect the consecuences of a malfunction of any Gauloment j important to safety oreviousiv evaluated in the SAR because thev do not make any chances to i the ohvelem! olant or any coeratino or accident resoonse procedures identified in the SAR of

! Technical S.+d0

canons. and berauem existina inservice testina (IST) oroarams in accordanca

with Am8 CMa.Section XI. will minimize the likelihood of ooeratina valves under dearaded i conditions.

i The consecuences of a malfunction of eauipment important to safety could be affected by l

j chances to the conditions under which it normally operates (environment. oower inputs.

d mechanical and electrical condition. etc.). none of which are beina chanced bv these chances.

< No ohvsical channes to the olant or its components are beina made. and no orocedumi chanaes in oDeration of the olant or response to any accident are beina made.

j The consecuences of a malfunction of eauioment important to safety may also be increased by l

i oDeratina the eauioment under dearaded conditions for lona ceriods of time. Such dearaded l conditions may include under voltaae. under-lubrication. reduced air sucolv and/or pressure.

1 i

! H.\AMALONE\WINWORD\VLVrIMEl. DOC Revised: June 8,1995 j

l

. l l

l GRAND GULF NUct EAR STATION UNrr 1 l CHANGES', TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 11 of 16

4. May increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR (Continued):

sAsis (CONTINUED): One of the svmotoms of ooeration under dearaded conditions. at least fgI some valves. is an increase in the measured stroke times. To the extent that increasina the allo =eNe maximum stroke time limit may allow a valve to be operated under dearaded conditions for extended Deriods. these increases in the limits miaht increase the consecuences of a malfunction of the affected valves. if they were the only maanicaNa limits.

However. the TRM maximum allowable stroke time limit is not the oniv stroke time limit for these valves. All of these valves are also reauired to be exercised and stroke time tested in accordance with ASME Code. Sedion XI. Subsection IWV. as reauired and specified in GGNS Technical Soecification 5.5.6. TRM 7.6.3.3. and 10 CFR 50.55a(a).

The NRC orovided auidance and clarification on the performance of IST on numos and valves in merardance Section XI bv i=*nMa Generic Letter 89 04.

• Guidance on Develonina Accentable (

Inservice Testina Prooramt (MEC-89/0124). In Attachment 1 to Generic Letter 89-04.

Sect!on 5. entitled "Limitina Values of Full-Stroke Times for Power Ocerated Valves.' the NRC )

orovided anMance. which is onated in Ametion ll.B.3 of this Safety Evein=elag.  ;

The diaen**lon in Section ll.B.3 of this Safetv Evaln=*ian **Med that =Mi'ional limits imaa*M by ASME Code.Section XI. and the NRC auldance in Genesic Letter 89-04. which gg imolemented in Soecification GGNS-M-189.1. would minimize oceratina valves undet dearaded conditions.

In essence. Accendix C of Specification GGNS-M-189.1 roouires that. if a limit calculated be**d on the nerformance of the valve is less than an analvtical limit in the SAR. Technical Soecifications or TRM. the lower limit shall be the limit for takina corrective achon. before the valva Derformance dearades to the coint that a malfunction could occur. If the analvtical limit is a less. the analvtical limit is the limit for takina corrective action Either limit minimizes the orobability of operatina the plant with valves in dearaded condition.

]

In addition. it is na**iNa that the consecuences of a component malfunction could be increased l

e it a valve's stroke time were decreased to the coint that severe water hammer was aenerated by the closure of the valve, None of these TRM chances represent decreases in the stroke time limits. they are all increases in the limits.

! Therefore. the conseauences of a malfunction of eauipment imoortant to safety oreviousiv evaluated in the SAR will not be increased, i

i j 5. May increase the possibility for an accident of a different type than any 0 Yes j previously evaluated in the S AR. g No i

} ggg: These chanaes do not increase the possibility for an accident of a different tvos than any j oreviousiv evaluated in the SAR because thev do not make any chances to the ohvsical clant i

or any operatino or accident response orocedures identified in the SAR or Technical l Specifications. and because existina inservice testina (IST) orocrams in accortlance with ASME Code.Section XI. will minimize the likelihood of operatina valves under dearaded conditions.

1 i

j H:\AMALONE\WINWORD\VLVrIMEl. DOC Revised: June 8,1995 3 _ ._

k

j. .

j

! GRAND GULF NUCLEAR STATION UNIT 1 CHANCES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 12 of 16 i

1 5. May increase the possibility for an accident of a different type than any previously evaluated in

! the SAR (Continued):

BASIS (CONTINUEDh To affect the aassiNiity of a new accident would reauire that one or more of j these valves were in a cosition different from its analyzed oosition at the onset of the gg,gident secuence or that the valve failed to stroke to its safety position durina its response to an isolation sianal.

The aassibility a new ereMent could be affected by extemal factors not within the control of the Owner or Operator. as well as by intamal dasian factors within the control of the Owner or Operator. However. In order to affect the oossibility of such an acc; dent the Owner would have i to chance the initial na*Mion of one or more of the valves identified in the TRM tables identified above. No chances to the operatino orocedures to chance the coeratina positions of any valve sre beina made. and no chvalemi chanaes in the olant are beina made.

The ONLY aaseible ways that these chanses cauM incr-- the aeWidv of any new meeMant would be if the valve failed to !*alete to its safety aa*Hian within the time allowed by cunent accMent analyses or if chances in the valve stroke times intradnead new instabilities. such as water hammer.

The proben!ity of valves failina to move to their safety cositions has been discussed in Section ll S.2 of this Safety Eve!"a'lan. The discussion concluded that 'Nse TRM chances would not increase the likelihood that the valves wauM not move to their safety positions. In addition. the l discussion in Section ll.B.3 of this Safety Eve!"a' ion exclained that additional stroke time limits l Imnatad by ASME Code.Section XI. and Generic Letter 89-04 would minimize operatina valves under dearaded conditions. which could also orevent the valves from closina when reauired.

The oossibility for an accMent of a different tvos could be increased by introducino instabilities into the system when the valves cle==. such as due to water hammer. Water hammer occurs when water flow is anddanly altered by stoaaina. startino or chanoina the direction of the water flow. The only way water hammer or other flow inataNiklan could be intmduced by these TRM chanaes would be if a valve's clonina stroke time were shortened to the coint that severe water ,

hammer or other flow iria*=Ninies were Generated by the closure of the valve. l None of the TRM chaname analvred in this Safety Evel"a' ion reoresent decreases in the stroke time limits: all of the channes are increases in the limits.

Therefore. the aa==iNidv for an accMent of a different tvos than any previously evaluated in the SAR will not be increased.

1

6. May create the possibility for a malfunction of equipment important to safety of Yes a different type than any previously evaluated in the SAR. g No ggp: These chances do not create the nassibility for a malfunction of eauioment imoortant to safety of a different tvos than any previousiv evaluated in the SAR because thev do not make any chanaes to the ohvsical plant or any operatina or accident resoonse Drocedures identified in the SAR or Technical Soecifications.

H:\AMALONE\WINWORD\VLVTIMEl. DOC Revised: June 8,1995

i i

i i

l GRAND GULF NUCLEAR STATION UNIT 1 l CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM i

PAGE 13 of 16 j 6. May create the possibility for a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR(Continued):

l

) BASIS (CONTINUED): The oossibility of a new malfunction of eauioment important to safety is i affected by chances to its operatino conditions (environment. oower inouts. mechanical and l electrical condition. etc.). none of which are beina chanaed by these chances. No ohvsical or i orocedural chances in the olant are beino made.

4 i The an==iN!iev of a new malfunction of eauioment imoortant to safety may also be increased by

! ooeratina the eauioment under dearaded conditions for lona ceriods. Such dearaded conditions i may inc!"de under voltaae. under-lubrication. reduced air anaalv and/or or==="re. etc. One of

! the svmotoms of operation under daaraded conditions. at least for some valves. is an increase j in the mes=" red stroke times. To the extent that increasina the allowable maximum stroke time limit may allow a valve to be operated under dearaded cGietiens for extended neriods. these i incr=a=*s in the .imits may increase the probability of a malfunction of the affected valves.

However. the TRM maximum allowable stroke time limit is not the oniv stroke time limit for j these valves. All of these valves are also reauired to be exercised and stroke time tested in 1 accordance with ASME Code.Section XI. Subsection IWV. as required and specified in GGN8 i Technical Soeciflem' inn 5.5.6. TRM 7.6.3.3. and 10 CFR 50.55aff).

The NRC orovided auidance and clarification on the performance of IST on Demos and valves in accardance with Section XI bv issuina Generic Letter 89 04.

• Guidance on Develocina Acceptable Inservice Testina Proarams" (MAEC-89/01 i 89-04. Section 5. entitled "Limitina Values of Full-Stroke Times for Power Operated Valves.*

i the NRC orovided auidance. which is auoted in Section ll.B.3 of this Safetv Evaluation. The l dimensaion in Section ll.B.3 of this Safety Evaluation exclained that additional limits imoosed by 3

ASME Code.Section XI. and the NRC auldance in Generic Letter 89-04. which are i imolemented in Sascification GGNS-M 189.1. would minimize ooeratina valves under i dearaded conditions. which could also prevent the valves fmm closina when reauired.

l In addition. the oossibility for an eauioment malfunction could be increased by introducina 4

l instabilities into the system when the valves close. such as due to water hammer. Water hammer occurs when water flow is suddenly altered by stoonino. startino or chanoina the f direction of the water flow. The only way water hammer or other flow instabilities could be l introdnead by these TRM chances would be if a valve's closina stroke time were shortened to the coint that severe water hammer or other flow instabilities were cenerated by the closure of the valve, j None of the TRM chances analyzed in this Safety Evaluation represent decreases in the stroke time limits: all of the chances are increases in the limits.

l Therefore. the nassiN!ity of a malfunction of eauioment lincortant to safety different from any

] oreviousiv evaluated in the SAR will not be increased.

t 1

i i

I H

\AMALONE\WINWORD\VLVTIMELDOC Revised: June 8,1995 I

l- .

l 4

i*

1 i

{ GRAND GULF NUCLEAR STATION UNIT 1

CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 14 of 16 1

l 7. Will reduce the margin of safety as defined in the basis for any Technical O Yes l Specification. g No i

sASIS

These chanaes do not reduce th' mamin of safety as defined in the basis for any i Technical Specification because they d i not make any channes to the ohvsical plant or any operatino or accident resoonse crocedU'es identified in the SAR or Technical Soecifications.

l Mamins of safety are associated Wfth the redundancy of two indeoendent crimary containment

' isolation valves. for crimary contair nent isolation valves. and with the redundancy of the u

Standbv Gas Treatment System (SQTS) as a backun for the secondary containment isolation valves.

In order to orevent reducina the rmrnins of safety associated with both crimafY and secondary containment isolation valves. t~e valves' capability of closina within specified maximum time limits when reouired must not t,e dearaded. Such dearaded conditions may include under-voltaae. under lubncation. reduced air sucoiv and/or pressure. etc.

One of the svmotoms of operation under dea 1ded conditions. at least for some valves. is an increase in the measured stroke times. To the extent that increasina the allowable maximum stroke time limit may allow a valve to be operated under dearaded conditions for extended oeriods. the mamins of safety may be reduced.

However. the TRM maximum allowable stroke time limit is not the oniv stroke time limit for these valves. All of these valves are also reouired to be exercised and stroke time tested in accordance with ASME Code.Section XI. Subsection IWV. as reautred and soecified in GGNS Technical Specification 5.5.6. TRM 7.6.3.3. and 10 CFR 50.55aff).

The NRC orovided auidance and clarification on the performance of IST on oumos and valves in accordance with Section XI bv issuina Generic Letter 89-04. " Guidance on Develonina Acceotable Inservice Testina Proarams" (MAEC-89/01241.

In Attachment i to Generic Letter 89-04. Section 5. entitled "Limitina Values of Full-Stroke Times for Power Ooerated Valves." the NRC orovided auidance. which is auoted in Section ll.8,3 of this Safety Evaluation. The discussion in Section 11.8.3 of this Safety Evaluation exclained that additional limits imoosed by ASME Code.Section XI. and the NRC auidance in Genenc Letter 89-04. which are implemented in Soecification GGNS M-189.1. would minimize ooeratina valves under deereded conditions. which could also prevent the valves from closina when reauired.

In addition. the mamin of safety could be reduced by introducino instabilities into the system when the valves close. such as due to water hammer. Water hammer occurs when water flow is suddenly altered by stoccina. startino or chanoina the direction of the water flow. The only way water hammer or other flow instabilities could be introduced by these TRM chances would be if a valve's closina stroke time were shortened to the coint that severe water hammer or other flow instabilities were cenerated by the closure of the valve.

None of the TRM chances analyzed in this Safety Evaluation represent decreases in the stroke time limits: all of the chances are increases in the limits.

Therefore. the mamin of safety as imolicitly defined in the bases for Technical Soecifications 3.6.13.1 and 3.6.4.2 will not be reduced.

It\AMALONE\WINWORD\VLVTIMEl. DOC Revised: June 8,1995

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPER1MENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 15 of 16 III. Environmental Evaluation .

O Not applicable per Environmental Evaluation Applicability Review IMPLEMENTATION oR PERFORMANCE oF THE ACTION DESCRIBED IN THE EVALUATED DOCt' MENT:

A. Environmental Protection Plan

1. Will require a change in the Environmental Protection Plan. O Yes g No ggis: Valve maximum allowable stroke times are not addressed in the Environmental Protection Ela!L B. Unreviewed Environmental Ouestion

1. Concems a matter which may result in a significant increase in any adverse O Yes environmental impact previously evaluated in the Final Environmental g No Statement (FES) as modified by the NRC staff's testimony to the Atomic Safety and Licensing Board (ASLB), supplements to the FES, environmental impact appraisal, or in any decisions of the ASLB.

gggg: As lona as the orimarv and secondarv isolation valves close within the maximum allowable stroke times addressed in the SAR. notential radioactive releases to the environment as a result of accidents will not exceed the auidelines of 10 CFR 100. and there will be no neaative effect on the environment.

The orooosed maximum allowable stroke tirres are in accordance with the times analyzed oer NPE Resnanse to 88R 94/6182. which are intended to orovide primary and secondary containment isolation sufficientiv aulckly after an acckisnt that radioactive releases to the environment as a result of accidents will not exceed the auidelines of 10 CFR E These chances do not chance any chvsical plant system or component. nor do they chance any operatino or set nresconse orocedures described in the SAR or Technical Specifications.

Since the maximum allowable stroke times addressed in these chanaes arc within those addressed in the SAR. there will be no effect on the environment. i l

2. Coneems a significant change in effluents or power level.. O Yes g No sAsis: The orooosed chances to the TRM maximum allowable stroke times of orimarv and secondary containment isolation valves do not represent channes in effluents or oower level.

Thev do not represent any ohvsical chanaes in the olant or in any olant operatino orocedure that could affect effluents or oower level. Discussion under Sections 11.B.1 throuah II.B.4 above cleativ exclain why the probability of occurrence and consecuences of accidents and eauioment malfunctions. which could increase effluents. will not be increased.

H:\AMALONE\WINWORD\VLVTIMEl. DOC Revised: June 8,1995

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 16 of 16

3. Concems a matter not previously reviewed and evaluated in the documents O Yes specified in II.B.1 above, which may have a significant environmental g No impact.

gg,[s: As iona as the crimarv and secondary isolation valves close within the maximum allowable stroke times aMrasaM in the SAR. notential radioactive releases to the environrnent as a ras"# of acrMents will not avW the auidelines of 10 CFR 100. and there will be no neoative effect on the environment.

The oroaW maximum allowable stroke times are in accordance with the times analyzed oer l NPE Raarwinse to FFR 94/6182. which are intended to orovide ortmary and secondarv l containment !aa!aHon sufficientiv auickly after an accident that radioactive releases to the environment as a raenk of wMants will not exceed the auidelines of 10 CFR 100. l These chances do no chance any chwsical clant system or comoonent. nor do they chance any I operatina or accident resconse procedures described in the SAR or Technical Soecifications.

Since the maximum allowahle stroke times aMeessed in these channes are within those addressed in the SAR. there will be no effect on the environment.

Signatures and Approvals Evaluated. 78 ORJOINATOR/ DATE

/

Reviewed / Approved:

//[] d 9 ff

/

Rzv!zwsR/ DAp /

Plant Safety Review Committee Review

[,

j/ ' CHAIRMAN, PSRC / DATE H:\AMALONE\W )RD\VLVTIMEl. DOC Revised: June 8,1995

) -

s j i

97 - o 15 - P45.6-l l

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM

1. Safety Evaluat'.on Overview A. Reference 'Ata ORIGINATOR: K.L. Walker DEFT / SECT: P&SE/RE EVAL. s: 95-0060 R00 I DOCUMENT EVALUATED: Technical Reauirements Manual 7.7.

1.1 REFERENCES

OQAM 3.4.2 i FSAR CHANGE REQUIRED 7 E Yes O No CR# CR 95-043 I FS AR SECTIONS TO BE REVISED. Accendix 3A. Daae 1.16-1 .

l TRM CHANGE REQUIRED? E Yes O No TECH. SPEC. CHANGE REQUIRED 7 O Yes E No CR# falal IS THE VAUDITY OF THIS SAFETY EVALUAllON DEPENDENT ON ANY CHANGES OTHER WAN O Yes THE CHANGE BEINO EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDITIONS)? E No

IF YES TO THE LAST QUESTION, HAVE THE ORGANIZATIONS RESPONSIBLE FOR THOSE CHANGES BEEN NOTIFIED 7 O Yes THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED l'RIOR TO IMPLEMENTING THIS CilANGE.

1 B. Executive Summarv (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT)

]

BRIEF DESCRIPTION OF CHANGE, TEST OR EXPERIMENT: This chance removes the commitment to submit a summary Startuo Reoort to the NRC under certain conditions. This I reauirement is contained in Reaulatory Guide 1.16. C.1.a. and acoears in the Technical Reauirements Manual 7.7.1.1 as well as in the UFSAR Rea. Guide commitments listina. The reoort has tvoically been submitted followina each reload.

< and this would no lonaer be reauired.

REASON FOR CHANOE. TEST OR EXPERIMENT: The reoort Contains only basic information and references other oroarams and reports available on site. NRC acoroval of startuo tests is not reauired. and comolete information is readily available in olant records of the various tests performed after each reload. Adecuate oroarams are in olace to ensure that testina is done and that results are analyzed and screened for non-

conformances or other oroblems. Summarizina this information for review and filina

by the NRC does not enhance olant safety. It does, however. reauire use of valuable plant time and resources.

~ _ .. _ ._ _ .. _ . _ . _ __ _ __.-_ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ ___

Q SAFETY EVALUAT!oN

SUMMARY

AND CoNCLUS!oNS: This proDosed Chanae serves only to remove an administrative reauirement associated with commitment to RG 1.16 C.1.a. Elimination of the summary Startuo Test Reoort currently sent to the NRC after each reload will in no way increase the orobability or conseauences of accidents or malfunctions oreviously evaluated. NRC aooroval of startuo test results is not recuired by RG 1.16. and the reoort is submitted for information ourposes oniv. No evidence could be found that the Commission in any way based the GGNS Safety Evaluation Report on a reauirement to submit startuo test results followina each reload. No modifications to the facility or to the conduct or review of startuo testina will be done under this chance. No new tvoes of events could be created by elimination of this reoort nor is any marain of safety affected. Thus elimination of the TRM recuirement for the summary Startuo Test Reoort and removal of the UFSAR commitment to comolv with this asoect of RG 1.16 do not oresent an unreviewed safety auestion.

IL Safety Evaluation O Not applicable per Safety Evaluation Applicability Review A. Technical Soecifications

1. Implementation or performance of the action described in the evaluated O Yes document will require a change to the GGNS Unit 1 Technical g No Specifications.

BAEi F*"ar'i"* r=yirenwata are enntnined in the TRM S~+ina 7.7. and are not desenbed m the Techmcal Specifications.

B. Unreviewed Safety Ouestion IMPLEMENTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT

1. May increase the probability of occurrence of an accident previously evaluated O Yes in the S AR. g No BAgi The proposed change makes no physical modifications to the facility or any operating, mamtanance, or testmg practices. Startup testmg will contmue to be enadacted in accordance with applicable plant programs and Technical Specification requuvanenen Reportmg or not reporting these results is unrelated to accident Probabdity.

2. May increase the consequences of an accident previously evaluated in the SAR. O Yes g No Basi: The proposed change makes no physical modi 6 cations to the facility or any operating, maintenance, or testmg practices. Startup testmg will contmue to be conducted in accordance with applicable plant programs and Technical Specification requirements. Reporting or not reporting these results is unrelated to accident consequences.

e

3. May increase the probability of occurrence of a malfunction of equipment O Yes important to safety previously evaluated in the SAR. g No 1

gagi No modifications to any systems, structures, or components are being made by the proposed change. The change only removes the requirement to summarue startup test results to the NRC. Stanup tests will continue to be performed and evaluated in accordance with applicable programs and requirements, and results will be available on-site for review at any time by the Comnussion. Herefore, there is no increase in the probability of malfunction of plant equipment.

4. May increase the consequences of a malfunction of equipment important to O Yes safety previously evaluated in the SAR. g No i g6gi: No modifications to any systems, structures, or t-t 3---- .ts are being made by the l proposed change. De change only removes the requirement to summarue stanup test l results to the NRC. Startup tests will continue to be performed and evaluated in l accordance with applicable programs and requirements, and results will be available on-l site for review at any time by the Commission. Herefore, there is no increase in the

! consequences of malfunction of plant equipment.

l 5. May increase the possibility for an accident of a different type than any 0 Yes previously evaluated in the SAR. g No BASIS No new accident possibilities are created since there are no physical modifications l being made by the proposed change, nor are there any changes to the way testmg is l conducted or evaluated. Adequate plant programs and procedures, combined with Technical Specification surveillance requirements are in place to ensure that thorough stanup testing is performed to confirm design predictions. NRC approval of startup tests is not required. He proposed change only affects reporting requirements for test results provided for NRC information purposes only.

6. May create the possibility for a malfunction of equipment important to safety of O Yes a different type than any previously evaluated in the SAR. g No DAgi: No new equipment malfunction possibilities are created since there are no physical modifications being made by the proposed change, nor are there any changes to the way testing is conducted or evaluated. Adequate plant programs and procedures, combined with Technical Specification surveillance requirements are in place to ensure that thorough startup testing is performed to confirm design predictions. De proposed change only affects reporting requirements for test results provided for-NRC information purposes only.

7. Will reduce the margin of safety as defined in the basis for any Technical O Yes Specification. g No gas.l.t There is no margin of safety related to reporting of information to the NRC.

Commission approval of test results is not required, and adequate review of startup tests i

is provided for under existmg plant programs and Technical Specifications. No change is being made to the stanup test procedures or review processes There is no mention in the Safety Evaluation Repon of requiring stanup test reporting to the NRC following reloads. Thus, removal of information-only reporting of startup tests results does not reduce the margin of safety.

a l

IIL Environmental Evaluation O Not applicable per Environmental Evaluation Applicability Review l IMPt.EMENTATloN oR PERFORMANCE oF THE ACTION DESCRIBED IN THE EVAlt!ATED Doctl MENT:

A. Environmental Protection Plan

1. Will require a change in the Environmental Protection Plen. O Yes !

g No l ,

m: Reportmg requirements for startup testmg are not addressed in the EPP.

B. Unreviewed Environmental Ouestion

1. Concems a matter which may result in a significant increase in any adverse O Yes environmental impact previously evaluated in the Final Environmental g No l Statement (FES) as modified by the NRC staff's testimony to the Atomic l Safety and Licensing Board (ASLB), supplements to the FES, environmental impact appraisal, or in any decisions of the ASLB.

m: Reporting requirements for startup testmg are not addressed in the FES.

2. Concems a significant change in effluents or power level., O Yes g No  ;

m: The proposed change deals only with NRC *inierative reporting requirements .

for startup test results and has no impact on effluents or power level. Any proposed I changes in effluents or power level are processed in accordance with applicable plant programs which are not being changed

3. Concems a matter not previously reviewed and evaluated in the documents O Yes specified in II.B.1 above, which may have a significant environmental g No impact.

m The proposed change deals only with NRC *inierative reporting requirements for startup test results and has no impact on the environment.

I Signatures and Approvals Evaluated . JO/9[

oRio!NAToR / DAT[ /

Reviewed / Approved: d l k. I. d> -lo-1'f RzviEwER/yTE Plant Safety Review Committee Review l '

Af1

?_~ 9

/ CHAIRMAN, PSRC / DATE

C 6 - o'M- Ps r GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, tests OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 1 of 14

1. Safety Evaluation Overview  :

A. Reference Data J IPT/ SECT; P&SE EVAL #: 95-0075-R00 ORIGINATOR: Alan J. Malone DOCUMENT EVALUATED: Chances to UFSAR Tables 6.2-44. 6.2-49 and 168-3.6.41. and Technical Reauirements Manual (TRM) Table TR3.6.1.3-1

REFERENCES:

10 CFR 50. A_naandix J: UFSAR Section 6.2 and Chapter 15: ANSI /ANS 56.8-1987:

UFSAR Chance RaaeM PL 93-007: GIN-95/01815: NUMARC. T. E. Tioton. letter. dated January 17.1992. SuMM: *NRC ProaaM Revision to 10 CFR 50. Accendix J. Containment Leak Rate Testina*: Safety Evaluation 93 0049 R00 FSAR CHANGE REQLTRED? E Yes O No CR # 95-065 FSAR SECTIONS TO BE REVISED: Table 6.2-44: Table 6.2-49: and Table 168-3.6.4 1 I TRM CHANGE REQUIRED? E Yes O No TECH. SPEC, CHANGE REQUIRED? O Yes E No CR# N/A 15 THE VALIDirY OF THIS SAFETY EVALUATION DEPENDENT ON ANY CHANOES OTHER THAN O Yes THE CHANGE BEINO EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDITIONS)? E No IF YEs TO THE LAST QUESTION, HAVE THE ORGANIZAT10NS RESPONSIBLE POR THOSE O Yes CHANGES BEEN NOTIFIED? ,

i THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED PRIOR TO IMPLEMENTING THIS CHANOE.

1 B. Executive Summarv (A1.30 SERVES AS INIVT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIIFT10N OF CHANGE, TEST OR EXPERIMENT: This chance deletes the reautrements to perform Tvoe C lam MM rate tdna on nine test connection valves listed in UFSAR Tables 6.2-44 and 6 2-49 and in Technical Reauirements Manual (TRM) Table TR3.6.1.3-1 (formeriv Tech Soec Table 3.8.4-1). The nine valves are the followina:

Valve No. Penetration No. Penetration Service 1821F025A 5 Main Steam A 1B21F0258 6 Main Steam B 1821F025C 7 Main Steam C 1821F025D 8 Main Steam D 1821F030A and F063A 9 Feedwater A 1821F030B and F0638 10 Feedwater B .

1E51F072 17 RCIC Steam Sucoiv Various directives associated with the local leak rate testina oroaram will also be revised as a result of this chanoe.

H:\AMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995

_ _ . _ m. _ _ _ . _ _ _ _ . . _ - _ _ _ _ . _ _ _ _ . _ . . _ _

I t

1 s

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 2 of 14 4

i j B. Executive Summary (Continued):

1 REASON FOR CHANGE, TEST OR EXPERIMENT: The test connection valves are not reauired to be Tvoe C

! local leak rate tested because thev do not conform to the characteristics of valves that are j reauired to be Tvos C tested under the definition of "Twee C Test

• as defined in 10 CFR 50.

3 Accendix J. Definition ll.H. Thev are small manual valves. are locked in closed position durina j oower operation. are operated irifrecuentiv. and are not canable of remote or automatic j

operation. In mMMion. hac="a= these test connection ala== mMar*i to their oroe*** oices a between inboard and outhaard main isolation valves and have additional valves and Dios caos in

! series. these oenetrations cresent multiple indeoendent barriers to leakana throuch the

} Denetration.

Eliminatina the Tvoe C tests of these test connection valves will save sinnificant outaae time.

man-hours and man-rem exposure.

l

! SAFETY EVALUATION

SUMMARY

AND CONCLUSIONS: The Safety Evaluation concludes that neither the i orohahi!My not the consecuences of an accident or malfundion of eauinment will be increased by exemotina the local leak rate testina. Local leak rate testino is not effedive in detectiner f

! miaaami+!oned valves. which is the oniv likelv reason that the orobabilltv of an accident or i malfunction of eautoment would be increased. The consecuences of an accident or malfunction I of aaniament are minimized by the valves' construction. their infreauent oceration. and administrative controls on the valves' disk positions.

Althouah these valves seem to be soecifically identified in ALwadix J. Definition ll.H. as reauirina Tvoe C testina. this evaluation lustifies why thev should be exemoted. In addition.

GIN 95/01815 documents Nuclear Safety & Reaulatory Affairs (NS&RA) oosition that these chanaes may be made under the orovisions of the 10 CFR 50.59 orocram.

i IL Safety Evaluation Not applicable per Safety Evaluauon Applicabdity Review A. Technical Soecifications l

1. Implementanon or performance of the action described in the evaluated O Yes document will require a change to the GONS Unit 1 Technical Specifications. g No gg,13: Valves and other containment oenetrations that must be Tvoe B or C tested Der 10 CFR l 50. Anoendix J. are listed in Technical Reauirements Manual (TRM) Table TR3.6.1.3-1. This j

table was oreviousiv Tech Soec Table 3.6.41 but was removed from the Technical

Soecifications (TS) in Amendinent 102. Part of the basis for removal of this table from the i Technical Specifications. as stated in Generic Letter 91-08 and the Nuclear Reaulatorv j Commission's (NRC's) Safety Evaluation of Tech Soec Amendment 102. is to " allow lise=; to i make corrections and undates to the list of components for which these TS reauirements acolv.

] under the Drovisions that control channes to olant orocedures as specified in the Administrative i Cnntrols Section of the TS.* Chances to TRM tables do not reauire a chance to the Technical i Soecifications. althouah thev do reauire comoliance with the Administrative Controls Section of j the Technical Soecifications.

HAAMALONE\WINWORD\TVADDEL. DOC Revised: August 28,1995

I J

i i'

i

GRAND GULF NUCLEAR STATION UNIT 1 i CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM i

i PAGE 3 of 14 i

i IL Safety Evaluation Not applicable per Safety Evaluation Applicability Review l

i 1

B. Unreviewed Safety Ouestion l

! IMPLEMENTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT:

1 4 1. May increase the probability of occurrence of an accident previously evaluated Yes i j in the S AR. a No

! ggg: The Anoendix J lankaaa testina reauirements are intended only to minimize leakana of

}

radicadivity from the containment durina and followina an accident Neither these channes nor i

the underivino A_naandix J testina reauirements affect the probability that an accident wlit occur.

The chanaes do not affect the Dhysical desion. or operatina cosition or status of any olant l

i comoonents or svstems.

t l

j 2. May increase the consequences of an accident previously evaluated in the SAR. Yes-

! g No l

j g[g: The resoonse of the olant and its eauioment to an accident reauirl.ea containment isolation

is basad on meetina the dadan and testina reauirements in reaulatory documents such as 10

{ CFR 50. Anoendices A and J. Accendix J roouires a oroaram to be develooed for conductina Tvos i A. B and C tests.

i l

10 CFR 50. Accendix J. Definition ll.H. defines Tvos C test as followe

. "H. 'Tvoe C Tests' means tests intended to measure containment isolation valve

leakaae rates. The containment isolation valves included are those that:

i "1. Provide a direct connection betwesn the inside and outside atmosoheres of j the ortmary reador containment undar normal ooeration such as ourne and l ventilation. vacuum relief emi iDstrument valves:

"2. Are raanired to close automatically uoon recelot of a containment isolation l sianal in resconae to controls intended to effect containmern isolation:

4 "3. Are renuired to ocerate intermittent!v under oost accident conditions: and '

5 "4. Are in main steam and feedwater oinino and other systems which penetrate i containment of direct cycle boihna water oower reactors."

l Containment lealmelan valves are defined in Accendix J. Definition ll.B.. which reads as follows:

i "B. ' Containment isolation valve' means any valve which is rolled upon to

{ oerform a containment isolation function."

j For the rest of this indMcation I will refer to the four conditions. alven in Definition ll.H.. that reauire i containment isolation valves to be Tvoe C leak rate tadad as & "".- e 1. 2. 3 and 4.

" Containment isolation funcbon" is not soecifica!iv defined in Anoendix J: however.10 CFR 50. .

Anoendix A. General Desson Crtteria for Nuclear Power Plants. General Desen Criteria (GDC) 54 l throuah 57 contain the dssion reauirements for Dioina systems oenetratino orimary containment.

GDC 55. 56 and 57. in oarticular. descnbe the number. locahons and arranaemert of valves which j Derform the containment isolation funcbon i

H:\AMALONE\WINWORD\TVADDEL. DOC Revised: August 28,1995 i

a

i .

i I

GRAND GULF NUCLEAR STATION UNrr 1 i

l CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM t

l PAGE 4 of 14

{ 2. May increase the consequences of an accident previously evaluated in the SAR (Continued):

i

sAsis (covitNtto)

Definition II.H. reouires Tvoe C testina to be oerformed on a containment isolation i valve if it meets at least one of Qualifiers 1. 2. 3 and 4. However. as diern==ad below. these valves

! do not meet the cualifiers in Anoendix J. Definition ll H.. that would reouire them to be subied to Tvoe C testina, l

A larne number of small (1" nominal oice size or lessi manual alobe valves. Indudina the nine valves l which are the subeed of this evaluation. are located on test connection. vent. and drain oscino which

! conned with orne=== oscina between the inboard and outboard isolation vanes of orimary contain-i ment oenetrations On some cenetrabons which have both contanment isolatiq[Lyjives outside the j containment wall. the test connedion. vent. or drain oscina connects with the orneman oscina between the contanment wall and the first unboarti) isolation vaNo. On some cenetrations. the cenetrabon

!, itself is the test ccnr+t -i and the valves are installed accordinoiv All of these valves meet the I daeaan r=anirements of GDC 55 or 56. as noted in UFSAR Table 6.2-44. for desionation as f contenment isolation valves.

1 I These test connad_ian. vent. and drain valves are maintained dosed at all times exceot when thev are i I

beina us=d for fillina. ventina. drainina or oerformina teshna whidt reouire them to be onen. The

! naadians of these valves are ccnuvi;.d and they are vertfled to be in their domed naeelans durina l l

valve lineun venfkatians after mmor system evolutions and orlor to restartina the olant. On most of

these cenetrations. induttion all of the cenetrabons that are the anhiace of this safety evaluation. there j is a =meand i$^wian valve in senes whidi is also controlled. There is also oenerally a nice can at the I end of a test co6r+R-n or vent line, in adreian. as noted in UFSAR Table 6.2 44. all of the test connection vent. and drain valves liatart in UFSAR Tables 6.2-44 and 6.2-49 are reauired to be i locked in the dosed oosshon durino normal ooershon Azek=NW of m "% 1

Since hv are cenic ":-i and lacerari &=-1 durina normal oceration

, these valves do not orovide a direce connection between 'he inside and outside atmoscheres of the orimary randar containment under normal ooerabon Therefore they do not meet Qualifier 1 of l Accendix J. Definition ll.H.. for Tvos C testina.

Ass!kehiH*v of O""lar 2: Since thev are manualiv-ocer=8ad vanes with no orovision for omer

! operation. these valves cannot r*wa automatically uoon recelot of a contenment isolation sianal.

! Therefore. thev do not meet Qualifier 2 for Tvoe C testina.

A--#F" of A " __- 3: Many of these valves are I -MW1 in the drvwell or other oarts of the f

! olant C. unmid not be :- - % ureer oost acewiant c6r r^ s. (The nine vaNas that are the i suhaar* of this safety eval"=*ian are in the Amalierv Builrtina Steam Tunnel.) All of these valves are 1 on systems whidi either are rar=_ared to ooerate under cost accident c66dn;0rs or are reouired to i ic^ late fmm the containment on recaia* of a containment leaMian sional. In either case these valves j must remain elaaad and =rwdd not be opened durina cost aceirient conditions Therefore. thev do not

meet Qualifier 3 for Tvoe C testina.

1 l Anolicability of QualPler 4: These valves are all on systems which oenetrate containment of Grand Gulf 1. which is a dired cvde boilina water oower reador therefore. on casual insoection. they all a meet Qualifier 4 for Tvoo C testina raauirements. However. we believe that Qualifier 4 raouires j interoretation to determine its intent.

4 I

i 0

1 H:\AMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995

)-

4

.g 4

4 i

i GRAND GULF NUCLEAR STATION UNIT 1

)

i CHANGES, TEsrs OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM 3

) PAGE 5 of 14 i

)

2. May increase the consequences pf an accident previously evaluated in the S AR (Continued):

3 sAsts (coNTINt'roh Qualifier 4 seeminoiv roouires that all containment isolation valves on any system i that oenetrates containment of a direct-evcie boilina water oower reador (BWR) must be Tvoe C J

ta*iad. In amian. Onnii8Ier 4 maammes to avranda fmm Tvos C ta=*ino all enntanment isolation valves on indirect-cvcie crassurized water oower readors (PWRs). includina valves and systems with similar ournoses and desions. unless they also meet Qualifier 1, 2 or 3 of definition ll.H. This l i interor=*=*ian daae not naamme to be the intent of Defitution ll.H.

The exarpalaa aivan in DefIrvtion ll.H.. " main steam and feedwater oscino and other systems which l

oenetrate containment " enaa==8 the true intent of onnai8ter 4. For the cumosas of mntanment iaala'_ian and leakana ,,. we:Tient. a direct-evde hainina water r=ar*ar (BWR) olant differs from an l 1 indired-cycle ora =amtzed water r=ariar (PWR) olant in one fundamental resoed The BWR main l l

l ,

steam. feedwgter. and some other steam system Dice lines Denetratino the containment boundarv j cany reactor coolant. whereas the simdar curnose PWR lines camr uncontammated secondary-side

! ==*=' The PWR steam aenerators (SGs) omvide a maior amianal banter to the escane of hinhlv-f contamer)ated reador coolant which is not omeent in a BWR olant. Mod of the other SWR swatoms.

such as r 11-- water deanun ,=i' >=8 heat i ..sval. and hiah i =s Gei& -- K h=

) counterparts in PWR systems. such as curification. decaw had mmoval and hiah creasure inloction.

} which nerform sirnelar fuer*ians in addi*ian sudi sveams as servios air. drain. samolina and i vermian svstems nerform the same furwians in both BWR and PWR niants.

It is unr===aaahia to ===nme that BWR olant systems that are desinned and construded similartv to l l

their PWR olant countemarts and cerform ~---CM=" furdians should be subled to l sionificantiv more restrictive leakana testino reautrements than their counternarts at PWR olants. l l

unless there is a sientfIcant reason.

For the main steam and feedwater svatoms. the reason is the PWR steam oenerator, which orevents the turbine =*amm and foodwater fmm beino contaminated from contad with the reactor core. The

- PWR man steam and feedwater lines. as well as some auaaliary lines. sudt as steam drans and

) steam oenerator chemical addleian lines. cany water and steam wtudi are not radiaar*ively contamh i nated: therefore. there is not the conum with asemaa of contaminated fluid from mntainment that

{ there is with dired-cvde BWR main steam. feedwater and some other swatoms. Plant-to.ciant differences in the C--A-- and level of contamert=*ian in the main steam and feedwater amono PWR olants and amonn SWR olants are not sinnificant enmaared with the much armater notential for '

j r-i- of radiaar**vity to the environment throuah the BWR steam and feedwater lines. It is clear.

i then. the the intent of Quallfler 4 is to reouire !Maaa testino of valves and oenetrations at BWR Diants whidi aana a sioniaemntiv arester hasard to the environment than valves and oenetrations which perform amdar fundions at PWR plants.

I Table 1 a8'areted to and cart of the !"a'inem8aan for UFSAR Chance Raouest No PL 93 007 i (incornorated into this avain=*'an bv reference) comoares the cenetrabons at Grand Gulf Unit 1 with i similar-ourcose cenetrations at PWR olants and identifies those constrations which meet Qualsfler 4.

1 i From a study of Table 1. It is clear that the only oenetrations whidt meet Qualifier 4 are those that carry rea:, tor steam out of containment or that cany feedwater fmm non safety related comoonents

{' out=ida containment to the r=ne*ar. The only oenetrabons at Grand Gulf Unit 1 which cany reactor steam or feedwater into or out of containment are the followina:

1 1

4 i '

HAAMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995 i

C___ _ _______ _ _ _

4 ,

]

e 1

l GRAND GULF NUCLEAR STATION UNrr 1 l

CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM

} PAGE 6 of 14 i

2. May increase the consequences of an accident previously evaluated in the S AR (Continued):

l BASIS (CONTINUED):

l l Penetrations 5 8: Main Steam Unes A D Penetrations 9 & 10: Feedwater unes A & B j Penetration 17: Steam Sunoiv to RCIC Tuttiine and RHR Heat Exchanoers

! Penetration 19: Main Steam Une Drains l It has already been demonstrated that manual test connection. vent and drain valves do not meet

! Opplifiers 1. 2. or 3 of Aapendix J. Definition ll.H. Manual test connection. vent and drain vanes on l pgetratuns other than the eloht oenetratioris identified above also do not meet Quallfler 4. In acco6nce with Definition II.H.therdo not mouse.Twoe.C testinn_as_ containment isolahon valves.

Tnerefore. thov have been exduded fmm the Tvoo C leakaos testino roouimments of GGNS

> Imoroved Tedt Aw SR 3.6.1.1.1 (creviousiv Tech Scoc. 4.6.1.2.d thmuch I). as documented in l and tushfled by UFSAR Change Raouest No. PL-93 007. The channo reauest added notes to UFSAR Tables 6 2-44 and 6.2-49 that dartfv that these valves do not mouwe Tvos C testino. The valves were

subsecuentiv removed fmm the GGNS Twoe C leakane testino amoram.

! In addition to meetino Quallfler 4 of Definition ll.H.. the main isolation valves on the eioht steam and l feedwater constrations identified above also meet at least one of Qualifiers 1. 2. or 3. Seven of these j oenetrahons (the exception is Penetration 19) also have test connection valves witNn the containment oenetrabon boundary. whidt am listed in UFSAR Tables 6.2.44 and 6,2 49 as containment isolabon valves, in addluon. two feedwater penetrations (Penetrations 9 and 10) also have drain valves witNn the containment oenetration boundarv. (Penetration 19 does not have any test connedion. vent or

drain valves witNn the cenetration boundarv.) The test connection valves (a total of nine valves) are i currentiv in the GGNS Tvos C leakane teshno omornm. These nine valves. which are listed in Sechon I.B. *Execubve Summarv.* are the subsect of tNs Safety Evaluation.

We beiseve that the intent of Definition ll.H. Quakfler 4 is to roouire Twee C testino of the main isolation s valves in applicable otoino cenetratino crimary reactor contamment but not to reouire Tvos C testino l of the test contweion. vent and drain valves that may be connected to the cloina. This belief le based j on the follow 6no documents omduced and tasued by the NRC. and it may refled a oossible shift in i thinluno by the Nuclear Raoulatory Commission (NRC) fmm their oosition when Accendix J was intually immund.

l l L in 1992, the NRC staff sent a orocosed revision to Anoendix J to the NRC Commissioners l

for accroval and subseouent oublication ISee NUMARC (T. E. Tioton) letter. dated January i 17.1992. Subled: *NRC Pmoosed Revision to 10 CFR 50. Anoendix J. Containment Leak

! Rate Testinal. The 1992 orocosed revision to Anoendix J contains the followino definition f of Tvoe C test:

t j "Tvoe C Test means a oneumatic test to measure containment isolation valve

leakace rates."

l In addition. the NRC's 1992 orocosed revision to Anoend!x J contains the followino Sect' ion Ill.C.Sfa):

I i

HAAMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995 l

4

. _ _ _ .__ _ _ _ _ _ _ . - _ . _ _ _ ._ _ -._ __. ___ __ ~ _ _ _

. ~ _ - - - - - . - - . - - - . - - - - - - - - _ ~ - _ - _ . -

)

i 4

j i GRAND GULF NUCLEAR STATION UNIT 1 CHANGES; TEsrs OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM l

i PAGE 7 of 14 j 2. May increase the consequences of an accident previously evaluated in the S AR (Continued):

BASIS (CONTINUED):

l '5. Valves That Need Not Be Tvos C Tested.

i j *(a) A containment isolation valve need not be Twee C tested if the valve

does not constitute a ootential containment atmoschere leakane cath durina or j followino an accMant. con *Merina the most limitina sinale active failure.' l

! The Be=== above in this eve!"=' ion have adeaustely e**=hilahed that. based on aualification i of their d**ian. matericis and construction. and on muiHala bardes to !eakaam maintained l

closed dunno power operation under an administrative control orocram. these test 1 connection valves do not constitute a potential containment atmosphere leakane cath durina or followino an accident.

l

! 2. The NRC's 1992 ornaaaad revision to Accendix J refers to a draft Reaulatorv Guide 1.)00(

l (number to be - "r.sd) for 'Saaelfic anidance concomina mecaa*=hia '"maa rate tam

• methods. oracad"res. and ansV= .

• The draft raanta*arv anwa did not meaamaanv the 1992 omaa==d A_aaandix J revision: however. the areamoanvino diaen==ian matettal l indicated that the raantatory anMe would endorse the 1987 revision of ANS 56.8 and would be revised to endorse any later revisions of ANS 56.8.

f l The 1987 maaroved revision to ANSI /ANS 56.8. American National Standard. ' Containment

! System Leakane Testina Raa"irements." addresses test connection valves. In Section 6.2.

' Test Boundaries and Connections for Testina." the standard states the followina:

l l *If it is neca==arv to inatall test connections between redundant containment j isolation valves. the connection should consist of a double barrier (e.o.. two i valves in se1es. one valve with a nionie and cao. or one valve with a nionie i

and blind flanae). These test connections are cart of the containment system i barrier. but due to their infrecuent use and multiale barriers. thav do not

! recuire leakaae rate testina as lona as the barrier conflourations are

! maintained usina an administrative control orocram."

i d

All of the nine test connection valves that are the subiect of this evaluation have a tricle barrier of two valves in series with a nicole and nice can. In addition. all of thegg itg connar*ians loin to their oroc=== maan in the Auviliary Buildino Steam Tunnel outboard from i the inhaand main containment isolation valves. Therefore. all of these nine test connection valves meet or exemd the reauirements of ANSI /ANS 56.81987 for exemotion from Tvoe

{1 Ctemina.

i

! 3. The NRC staff ormaared a Draft Reaulatory Guide. Task DG-1037 in Aununt.1994. entitled

' Performance-Ba*ad Containment Leak Test Procram.' In Section C. "Reaulatont Position."

the NRC staff oroaaamd to endorse Draft Nuclear Enemy Institute (NEI) Guideline NEl 94

} 01. Revision D. dated October 25.1994. ' Industry Guideline for Imdsmsauas Performance-

{ Based Ootion of 10 CFR 50 Anoendix J.* NEl Guideline 94-01 contains the followina words i under Section 6.0 General Recuirements on Pace 4:

I 'An LLRT is a test oerformed on Tvoe B and Tvos C comoonents. An LLRT is

} not reauired for the followino cases:

1

! HAAMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995 9

1

7.

1 O

i t

4 GRAND GULF NUCLEAR STATION UNIT 1 i CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM

! PAGE 8 of 14 i

f 2. May increase the consequences of an accident previously evaluated in the S AR (Continued):

i

BASIS (CONTINUEDh f *. Primary containment boundaries that do not constitute potential orimarv i containment atmosoheric cathways durtna and followina a Desian Basis l Accident (DBAt

• . Boundaries sealed with a cualified seal system: or.

f l *. Test connection vents and drains between orimary containment isolation

! valves which are one inch or less in size. administratively secured closed l and consist of a double barrier.'

} All of the nine test connection valves that are the subiect of this evaluation are 3/4 inch in nominal aias diameter and have a triole barrier of two valves in series with a plaaaa and l

pios can. therefore. they meet the third bulleted aualifier above. In addition. since they ara adtninistratively controlled closed at all times durina Dower oceration. thev also meet thG l first bulleted Qualifier above.

l l The reaulatory oosition in the Draft Reaulatorv Guide DG-1037 took excention to certain l auidelines m NEl 94-01. but it did not take exception to the guidance quoted above. therebv j imolvina NRC acceptance of the auidance.

The points above clearty establish that. since at least 1992 the NRC staff has been willina to l

accent the oosition that Accendix J does not reauire these nine test connection valves to be

Tvoe C tested. Based on the ooints above. Assendix J. as currentiv internreted by the NRC.

! does not recuire these nine test connedion valves to be Tvos C tested.

) The accident scenarios in the UFSAR assume no leakana throuah test connection vent and i drain valves: however. adminktrative i control

. of no leakaae throuah any test connscuen. vent or drain lines without the need to perform leakaae testina. These administrative controls ensure that at least a double banier to leakane is oresent at all times when containment inteartty is reauired. The followina are the edni;nimeuve

! controls which are considered in this evaluation:

( jL 1:vaaled restoration instrudions in the local leak rate test orocedures for containment i oenetrations specify the restored Dos #tions for the valves and Dice caos on each test. vent

$ or drain cGGnecuen used in local leak rate tests and reauire double verification that the l valves and emaa are correctiv restored. All of these nine test connedions have three barriers (two manual alobe valves and a oice cao) in senes.

l l tL These cenetrations and valves are lined up and verified to be in soecified oositions when j restorina each system to operability. in accordance with system ooeratina instrudions.

k L These test connection valves. since thev are located in the Auxiliary Buildina Steam Tunnel.

j are verified closed durina each cold shutdown. if not verified.within the_DfB_YlDila.92_dava, j The verification is normally performed immediately before. er in coniundion with. startuo of l the Dlant from the Cold shutdown Condiliga 1 Pine ciae on these oenetrations are controlled under the GGNS confiauration control j Droaram.

J HWiALONE\WINWORD\TVADDEL. DOC Revised: August 28,1995 1

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 9 of 14

2. May increase the consequences of an accident previously evaluated in the S AR (Continued):

BASIS (CONTINUED): Therefore. the consecuences of an accident oreviousiv evaluated in the UFSAR are not increased by these channes.

3. May increase the probabi'ity of occurrence of a malfunction of equipment Yes important to safety previously evaluated in the SAR. a No ggg: The Anaandix J !eakaae testina reautrements are intended only to minimize leakana of radioactivity from containment durina and folloMna an accident. j The occurrence and con =aa"ances of malfundions of aa" lament imcortant to safetv eagad be affected by l==haam thraea_h these valves: however. leakana rate testina of these valves could not be relied on to dated such leakana. since the valves are usually recositioned after the leakmae rate tests. Ims,=+r valve aaedianino is the most llkelv e="== of any such !eakaam.

The orobahility of imornaar valve naadianina is minimized bv cerformance of valve aHanments and position veriflemans under the Aclm;a;.L uve Controls crocram described in Section 11.B.2 above. I aak rate testino of these valves is not da*'aned to identify miscositioned valves. ,

Therefore. administrative control of valve aa=dian. not leak rate testina. is the best way to limit I leekmae due to misoositionina of these valves.

Damaae to the valve's ?=M or disk is the only physical damaae that could be detected oniv bv leak rate testina these test connection valves. The leak rate tests of the main isolation valves in each Denetration throuah these test connection valves check for leakana throuah the oice on both sides of the test connection valve. as well as throuah the valve body and stem osckina.

Damaae to the e=M or disk of one of these test connection valves is unlikelv for the followina reasons.

a. All of the Denetrations. alaa and valves are dealaned and constructed in accoidance with ASME Boiler and Pra=="re Vasaal Code. Section Ill. S"haactinn NB or NC (Class 1 or 21.

b. These test connac'_lan valves are 9ead on!v for leak rate testino other valves in their re=aactive Genetr=" -A Thev are not enhiedad to flow durina clant oceration. and are normally ela=ad. aveaa* durina tests. Durina leak rate tests of the other valves. the only flow thra_na_h the valves is clean. filtered air or water.

c. These test conn =-A-7 valves may also be 9ead for ventina. drainina and fillina their ra=aae*ive orae === cicina durina cold shutdown conditions. Thev are not subledad to flow durina olant aaara'lan. and are normally closed. exceot durina ventina. drainino and fillino.

Under these condMinns. the systems are not at hiah oressure or temocrature: therefore. the dancer of sionificant damane to the *=='s and diake of these valves durina ventina. drainina and fillina is minimal. In addition. such ventina. drainina and fillina ocerations are infrecuently oerformed events ("=na!!v no more than once Der refuelino cycle). and.

therefore. do not subiect the valves to hiah numbers of cycles.

Therefore. this chance does not incran=a the probahi!My of occurrence of a malfunction of souloment important to safety oreviously evaluated in the UFSAR.

H:\AMALONE\WINWORD\TV&DDEL. DOC Redsed: August 28,1995

j l

1

}

).

1 l 4

GRAND GULF NUCLEAR STATION UNIT 1 l CHANGES, tests OR EXPERIMENTS SAFETY AND ENVIRONMEffTAL EVALUATION FORM t

i PAGE 10 of 14 t

1 i

! 4. May increase the consequences of a malfunction of equipment important to O Yes f safety previously evaluated in the SAR. c No

<l ,

j ggg: The Anoendix J leakaae testina reauirements are intended only to minimize leakaae of 1

radioactivity from the containment durina and followina an EUZent. The occurrence and conse-i cuences of malfunctions of eauinment imoortant to safety could be affected bv leakaae throuah l l these valves

however. leakana rate testina of these valves could not be relied on to detect sugh ltakage. since the valves are usually recositioned after the leakage rate tests.

l 1 imoroner valve oositionina is the most likely cause of any such leakane The sisbebnity of imoroner valve na=*lonina is minimized by performance of valve alianments and naadian verifications under the Adm;n;&^ u .uve Controis crocram described in Section ll.B.2 above. Leak I rate testino of these valves is not desianed to identify miscondioned valves. Therefore.

l ,

administrative control of valve naadian. not leak rate testina. la the best way to limit leakane due i to mispondionino of the test connection valves. l Dammaa to the valve's seat or disk is the oniv ohysical damane that could be detected only tar l leak rate testina these test connection valves. The leak rate tests of the main isolation valves in each nonetration throuah these test connection valves check for leakana throuch the nice ort both *1dae of the test cGnn6cuGn valve. as well as throuah the valve body ardi stem nackina.

{

j Dammaa to the seat or disk of one of these test connection valves is unlikelv for the followina

reasons,

a. All of the penetrations. oice and valves are desianed and constructed in accordance with ASME Boiler and Pra*=me Vessel Code. Section Ill. Subsection NB or NC (Class 1 or 21.

b. These test connection valves are used only for leak rate testino other valves in their resoective oenetrations. Thev are not subiected to flow durina clant oceration and are 1

normally danad. avema8 durina tests. Durina leak rate tests of the other valves. the only I flow throuah the valves is clean. filtered air or water.

c. These test connection valves may also be used for ventina. drainina and fillina their resoective ornea*= ninina durina cold shutdown coridhians. Thev are not subiected to flow durina olant oaaratian. and are normallv closed. excent durina ventina. drainino and fillina.

Under these corid#ians. the systems are not at hiah pressure or temocrature: therefore. the danner of sionificant dammaa to the seats and disks of these valves durina ventina. drainina and fillina is minimal. In adddion. such ventina. drainino and fillina ocerations are

- infr=anantiv nerformed events (uaum!!v no more than once per refuelino cycle). and.

therefore. do not subiect the valves to hiah numbers of cyclesi Therefore. this chanas does not increase the consecuences of a malfunction of eauioment

important to safety previousiv evaluated in the UFSAR.

1, j 5. May increase the possibility for an accident of a different type than any Yes j previously evaluated in the SAR. g No i

i ggg: The Anoendix J leakaae testina reouirements are intended only to minimize leakace of radioactivity from the containment durina and followina an accident. They do not create the oossibility of a different tvoe of accident.

H:\AMALONE\WINWORD\TVADDEL. DOC Revised: August 28,1995 1

1

(*

]

i i

i GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM  ;

j j PAGE 11 of 14 l 6 May create the possibility for a malfunction of equipment important to safety of O Yes g No a different type than any previously evaluated in the SAR.

gg[g: Any accident that could be oossible due to leakana throuch one or more of the valves j considered in this chance would be due to imoroner oositionina of the valve (s). Leakaae testina d

of these valves is not daabned to identify miscositioned valves. The orobability of imorocer

! valve aaMionina is minimized bv control of and performance of valve aHanments and naMion i verifications under the Administrative Controls orcoram described in Section ll.B.2 above.

i

{ The Anoendix J leakana testino reauirements are intended oniv to minimize leakaae of f radioactivity from the containment durina and followina an accident. Various malfunctions of the containment tsolation system components have already been analvZed for such accidents.

These test connection valves are not beina ohvsically modified or deleted. These chances delete leak rate testina of these test connection valves. Leakaos through these valves haa 4 already been considered in loss of containment intearity accidents that have already been

' analyzed in the UFSAR. Therefore. these channes do not create the possibility of a malfunction of a different tvpe than any evaluated oreviousiv in the UFSAR.

1 1 7. Will reduce the margm of safety as de6ned in the basis for any Technical O Yes Speci6 cation. g No l l

1 gglg: Deletina the Tvoe C leak rate test reouirements for these valves does not affect the ooeration or operatina carameters of any system or the olant. Leak rate testina is only a orecaution. These chances do not ohvsically add. modify or delete any comconent or setooint in )

i the olant. These changes delete leak rate testina of these test connection valves on main steam

! and feedwater oenetrations. They do not add. modify or remove fmm the olant any component reauired to be leak rate tested.

! The 1987 acoroved revision to ANSI /ANS-56.8. American National Standard " Containment l

System Leakane Testina Reauirements." addresses these test connection valves. In section 6.2.

l

" Test Boundaries and Connections for Testina." the standard states the followina

"If it is nara--v to irsataH test connections between redundant containment isolation valves. the connection should consist of a double barrier (e.a.. two valves in series. one valve with a nionie and can. or one valve with a nicole and blind flanae). These test connections are part of the containment system barrier. but due l to their infrecuent use and multiple barriers. they do not recuire leakane rate testina l as lona as the barrier conflourations are maintained usina an administrative control orocram?

All of the test connection valves identified in these chances have a tricle barrier of two valves in series and a oice can. In addition. these test connections loin to their process otoes between j inboard and outboard isolation valves. This main orocess valve orovides an additional barrier to j leakaae throyah the oenetration.

4 j Additionallv. ASME Boiler and Pressure Vessel Code.Section XI. Subarticle IWV-1200. exembts j these valves from inservice testina. includina the seat leakana testina which would otherwise be

reauired by Subarticle IWV-3420. due to their beina one inch nominal oice size or smaller.

l HAAMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995

y __ ._ _... _ ..___ _ _ . _ __._.._._____ _ _ _ __ _ _ _ _ _ _ _ ___ _ __ _

i j .

1 4

i l GRAND GULF NUCLEAR STATION UNIT 1 i CHANGES, tests OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 12 of 14 l 7. Will reduce the margin of safety as defined in the basis for any Technical Specification.

l (Coatinued):

! sASIS (CONTINUEDI: The marains of safety defined in Imoroved Tech Soec LCO 3.6.1.1 and j SR3.6.1.1.1 (oreviousiv Tech Soecs 3/4.6.1.1 and 3/4.6.1.2) Bases could only be reduced if i sionificant leakaae would not be detected except by leak rate testina. Such leakace could only j be due to misoositioned valves or ohvsical damaae to the pressure boundarv. The orobability of i

improcer valve oositionina is minimized by oerformance of valve alianments and Dosition verifications under the Administrative Controls orocram described in Section ll.B.2 above. l.gh rate testino of these valves is not deskinsd to identify misoositioned valves.

, in addition. these valves may be opened to fill and vent oloina durina system restoration after

{ thev are tested. Therefore. administrative control of valve coastion. not leak rate testina. is the j best way to limit leakaos due to miscosdionina test connection valves.

Damaae to the valve's seat or disk is the only physical damaae that could be detected only by j leak rate testino these test connection valves. The leak rate tests of the main isolation valves in j each Denetration throuah these test connection valves check for leakage throuch the nice on i both sides of the test connection valve. as well as throuah the valve body and stem Dacking.

l Damaae to the seat or disk of one of these test connection valves is unlikelv for the followina

reasons

$ a. All of the oenetrations. Dios and valves are desianed and constructed in accordance with j ASME Boiler and Pressure Vessel Code. Section Ill. Subsection NB or NC (Class 1 or 2).

j b. These test connection valves are used oniv for leak rate testina other valves in their

resoective penet. rations. Thev are not subiected to flow durina olant operation. and are

normally closed, exceot durina tests. Durina leak rate tests of the other valves. the ork flow throuch the valves is clean. filtered air or water.

c. These test connection valves may also be used for ventina. drainino and fillina their l

i resoective orocess otDino durina cold shutdown conditions They are not subiected to flow i durina Diant operation. and are normally closed. except durina ventina. drainina and fillina.

! Under these conditions. the systems are not at hiah pressure or temocrature: therefore. the dancer of sionificant damaae to the seats and disks of these valves durina ventina. drainina l and fillina is minimal. In addition. such ventina. drainina and fillina operations are l

infmouently_ oerformed events (usually no more than once per refuelino cycle). and.

? therefore. do not subiect the valves to hiah numbers of cycles.

Therefore. these chances do not reduce the marains of safety defined in the bases for improved l . Tech Soec LCO 3.6.1.1 and SR3.6.1.1.1 (oreviousiv Tech Soecs 3/4.6.1.1 and 3/4.6.1.21.

s}

1 I

)

I

H

\AMALONE\WINWORD\TV&DDEL. DOC Revised: August 28,1995 3

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 13 of 14 IIL Environmental Evaluation li O Not APP cable per Environmental Evaluation Applicability Review i

1 IMPLEMEN" RAT!oN oR PERFORMANCE oF THE ACT!oN DESCRIBED IN THE EVALUATED DOCUMENT:

A. Environmental Protection Plan

1. Will require a change in the Environmental Protection Plan. O Yes g No gg[g: The Tvoe C testina orocram is not addressed in the Environmental Protection Plan.

l B. Unreviewed Environmental Ouestion

1. Concems a matter which may result in a significant increase in any adverse O Yes environmental impact previously evaluated in the Final Environmental g No Statement (FES) as modified by the NRC staff's testimony to the Atomic Safety and Licensing Board (ASLB), supplements to the FES, envtranmental impact appraisal, or in any decisions of the ASLB.

g,gg: As lona as the crimary containment is maintained leak tiaht within the reaullements of 10 CFR 50. Aooendix J. octential radioactive releases to the environment as a result of accidents will not exceed the auidelines of 10 CFR 100. and there will be no neaative effect on the environment The orooosed channes to the Tvos C testina oroaram have been evaluated as discussed in Sections I (Chance to Technical Soecifications) and ll.A (Unreviewed Safety Question) above and have been determined to be accectable, as discussed in the Basis statements above.

These chanaes do not chance any ohvsical olant system or comoonent. nor do they chance an ooeratina or accident resconse orocedures described in the UFSAR or Technical Snecifications.

Since the overall containment leakana rates will be maintained within the limits allowed by 10 CFR 50. A aaandix J. the GGNS Technical Soecifications. and the GGNS UFSAR. even with these channes. there will be no effect on the environment.

2. Concems a significant change in effluents or power level.. O Yes g No sAsis: The test connection valves addressed in this evaluation are closed at all times durina Dower operation. and are opened only for testina. ventina. drainina. and fillina the associated oloina systems. All effluents that may be collected from these drains are handled in accordance with clant administrative procedures and the Health Physics oroaram.

These chances do not reoresent any ohvsical chanaes in the olant or in any olant ooeratina orocedure that could affect effluents or oower level. Discussion under Sections ll.B.1 throuch II.B.4 above clearly exclain why the orobability of occurrence and consecuences of accidents and eauioment malfunctions. which could increase effluents. will not be increased.

H:\AMALONE\WINWORD\TVADDEL. DOC Revised: August 28,1995

GRAND GULF NUCLEAR STATION UNrr 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM PAGE 14 of 14

3. Concerns a matter not previously reviewed and evaluated in the documents O Yes specified in ILB.1 above, which may have a significant environmental impact. g No ggg: As lona as the otimary containment isolation valves are maintained within the maximum allowable lenkMe rates wraW in the UFSAR. ootential radioactive releases to the environment as a ram of acMants will not exceed the auidelines of 10 CFR 100. and there will be no neaative effect on the environment.

The oroMW chanaes to the Tvoe C testina oroaram have beea evaluated as discussed in Sections I (Chance to Technical Soecifications) and ll.A (Unreviewed Safety Question) above and have been determined to be accootable. as discussed in the Basis statements above.

These chanaos do not chance any DayWcal olant system or component. nor do they chance any ooeratino or accident resoonse procedures desca 'in the UFSAR or Technical Soecifications.

Since the overall containment !enkMe rates will reniain within the limits reauired by 10 CFR 50.

Anoendix J. there will be no effect on the environment.

Signatures and Approvals Evaluated: 8-28-95 ORIGINATOR / DATE Reviewed / Approved: .s&/ /

6' REVIEWER / DATE

/

Plant Safety Review Committee Review lfbm$ '

CHAIPA 7 7 PJ

, PSRC / DATE

~

H:\AMALONE\WINWORD\TV&DDEL. DOC Revised: Acgust 28.1995

'. 95 - 09 8 - PS E l GRAND GUL." NUCLEAR STATION UNIT 1 l CHANCES, TESTS OR EXPERIMENTS [ AFETY AND ENVIRONMENTAL EVALUATION FORM l

L Safety Evaluation Overview A. Reference Data ORIGINATOR: Ken Walker, DEFr/ SECT: P&SE/RE IvAL. s: 95-0072-R00 DOCUMENT EVALUATED: EER 95-6156

REFERENCES:

EER 95-6156: GTC - 95/0299: GEK 73674A FSAR CHANGE REQUIRED? O Yes  % No CR # NA FSAR SECTIONS TO BE REVIsEu: NA TRM CHANGE REQUIRED? O Yes h No TECH. SPEC. CHANGE REQUIRED? O Yes  % No CR# NA

! Is TIIE VALIDITY OF THIs SW 7 EVALUATION DEPENDENT ON ANY CHANGES OTHER THAN N Yes THE CHANGE BEING EVALUei, (E.O. PROCEDURAL, OPERATIONAL CONDITIONS)? O No Validity of the SE is dependent upon Operations procedure changes being in compliance l - with the stipulations of EER 95-6156. Specifically, drive water pressure may not exceed

350 psid above reactor pressure if reactor power is greater than the high power setpoint.

Other stipulations are made in the EER, and this SE also relies on those even though they may not be specifically d!='n=d lierein.

[F YEs TO THE LAsT QUESTION, !! AVE TIIE ORGANIZATIONS REsPONslBLE FOR Tl!OsE g

CHANGES BEEN NOTIFIED?

THE REsPONs!BLE ORGANIZATIONS MUST BE NOTIFIED PRIOR TthfLEMENTING

! THIs CHANGE.

l B. Executive Summarv (ALSO SERVES AS INPUT TO NRC

SUMMARY

REPORT)

BRIEF DESCRIPTION OF CHANOE, TEST OR EXPERIMENT: EER 95-6156 allows control rod drive system drive water pressure to be temporanly increased up to 475 psi above reactor pressure for purposes of withdrawing control rods which will not move at normal pressure. This pressure increme is to be done under Operations Off normal Event Procedures, and is subject to certain limitatxxis as desenbed in the EER and in this eva* cation. No FSAR changes are l

required as the FSAR discusses only normal CRD system pressures and does not discuss limits on temporartly exceeding such presures.

l REASON FOR CHANGE, TEST OR EXPERIMENT: Control rods often prove difEcult to withdraw at normal control rod drive system drive water pressure, especially from position 00 and/or during restart from a reactor scram. It is necessary to allow plant Operators to temporanly increase this pressure in order to withdraw the rods. Current procedures (ONEP 05-1-02-IV-1) allow an increase only to 350 psi above reactor pressure. This often is insufEcient to iritiate movement, and a higher pressure is needed.

I

a e

SArrrY EvAWATIoN

SUMMARY

AND CoNCWslONs: No change to Technical SpeciScations or the  ;

TRM is necessary since the drive water pressure is riot specifically discussed in these l documents Discussions with the NSSS supplier, GE, (GTC-95/00299) indicate that drive i water pressures up to 500 psi above reactor pressure are allowable without presenting a risk of damage to the CRD mechanism such that control rod scram could be inhibited. One important consideration involves the chance for inadvertent over-notching of the CRD mechanism during withdrawals due to use of elevated pressures. Engineering evaluations by System Engineering and experience indicate that over-notching of up to 4 notches (08 positions) is feasible. If reactor power is above the high power setpoint (~70% power), such an event could result in movement exceeding that intended to be allowed by the Rod Withdrawal Limiter (e.g. 2 notches), thus violating the assumptions of the Rod Withdrawal Error (RWE) analysis. His evaluation therefore supports only the use of elevated pressures above 350 psid when below the high power setpoint. Below that power level, pressure increases up to 475 psid over reactor pressure are temporarily allowable since ovemotching of 3 notches, or even 4 notches, would not exceed the RWURWE travel limit of 4 notches.

Thus, with the above restrictions, there is no increase in the probability or consequences of any accident or malfunction previously analyzed. The RWL will continue to protect agamst the possibility of an unbounded Rod Withdrawal Error, ne possibility or consequences of a Control Rod Drop Accident are also unaffected by this change. No new types of events are created. Here are no additional changes to the system operstmg procedures nor are there any changes in system design. De control rod drive system will not be inhibited from performing its scram insertion function if called upon to do so. No margins of safety are being affected since there is no impact on compliance with the MCPR safety limit, plastic stram limit, or radiological dose limits. Therefore, no unreviewed safety question is created by allowing an increase in CRD system drive water pressure of up to 475 psid above reactor pressure orovided pressure is not increased over 350 psid above reactor pressure when above the high power setpoint.

H. Saf Evaluation O Not applicable per Safety Evaluation Applicability Revicw A. Technical Specifications

1. Implemasitation or gifvinimace of the action described in the evaluated O Yes document will require a change to the GGNS Unit 1 Technical g No Specifications a6Ei The drive water pressure differential is not limited or discussed in the Technical Specifications. The restrictions being placed upon elevated pressure differentials when reactor power is above the hig,h power setpoint will enstre that the basis for TS 3.3.2.1 are maintained. Control rod scram times as required by TS 3.1.4 will not be affected by the proposed change. Any problems with rods which are immovible will continue to be addressed adequately by TS 3.1.3. Any minor rod malfunctions resulting in rod pattern violations will be adequately addressed by TS 3.14.

_l.

1 1 . l

. l I

! l

B. Unreviewed Safety OuestiQD IMPLEMENTADON OR PERFORMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT f

i 1. May increase the probability of occurrence of an accident previously evaluated O Yes j in the SAR. g( No  ;

BASIS: The events in the SAR which are potentially impacted by this change include all l

events relying upon a reactor scram for mitigation as well as the Rod Withdrawal Error j and the Control Rod Drop Accident. Discussions with the NSSS supplier (GTC 95/00299) indicate that drive water pressures of up to 500 psid above reactor pressure

] are acceptable without damage potential even though these are not specifically allowed 1 by the vendor manual. Relief valves are also provided on the drive water piping to j prevent exceedmg potentially damagmg pressures. Thus, the ability of the control rod i drive to accomplish the scram function will not be affected and the potential for a failure i to scram when required is not increased f The Rod Withdrawal Error analysis (UFSAR 15.4.2) considers the inadvertent contmuous withdrawal of a control rod to a point where fuel damage could occur. The nossibility of unintentional rod withdrawal under limiting conditions is not affected by having a higher drive pressure. (The consequences are discussed under B.2 below.) Further, the Control Rod Drop Accident addresses the results of.an uncoupled control rod falhng out of the core under worst case conditions. 'Ibe likelihood of a control rod t+:-- g uncoupled and dropping is not impacted by the proposed change l

1 l

. ~ . . ~ . . . . . . . . . -- .- - . . . .

j O

j' i

i 2. May increase the consequences of an accident previously evaluated in the SAR_ C) Yes S

i Iwgt As described above, the ability of the CRD system to perform a reactor scram is

not impacted by having a drive water pressure up to 475 psid above reactor pressure since no physical damage can result from such pressures. Should a scram occur while i drivmg a rod, accumulator pressure will still be directed to the underpiston area as j designed. Thus, ti. consequences of any event requiring a reactor scram are not l increased.

l Conceming the RWE event, two conditions must be considered: ahon and below the high i power setpoint (HPSP). Above the HPSP, the RWE analysis (UFSAR 15.4.2) assumes

that rod totion is stopped prior to a rod exceedmg I foot (2 notches) of travel. It is

! conceivable based on experience and engmeering estunates, that inadvertent withdrawals of up to as much as 4 notches could occur at the pressures under consideration. His would vio. late the assumptions of the RWE analysis even with the withdrawal limiter functinaing properly. T1ua, this evaluation does not suonort elevated drive water pressures exceedirg the 350 psid value already allowed by ONEP 05-102-IV-1 v&gg reactor nower is above the HPSP. When power is below the HPSP, a 2 foot (4 notch) withdrawal is allowed prior to stopping movement. As discussed in EER 95-6156, i experience, jnA , aam. and engmeermg estimates indicate that over-aatchia: in excess of this amount is not evW at the proposed differential pressure of 475 psid. Further, while rod withdrawal speed is expected to be faster for higher drive water pressure, this is not a factor ameting the RWE analysis results Thus, should an iiirds.t-i rod withdrawal occur under limiting or near-limiting core conditions, the -anariaae of the RWE analyses would be met and the resultmg consequences would be no more severe The MCPR safety limit would not be exceeded, nor would the 1% plastic stram limit be violated. Note that this evaluation assumes that normal procedural controls applicable to operation of the CRD system and RWL which are credited in the RWE analysis contmue to apply (UFSAR 15.4.2.2.2).

He consequences of a CRDA are also not increased by higher drive water pressure. The radiological consequences (fuel failures) resultmg from the CRDA are influenced prunarily by the control rod worth which is a function the core enadi+inan and rod pattern at the time the event occurs A lugher drive pressure does not affect these items except that it is possible for an overnarching event to result in a rod pattern which temporarily violates the constramts of the banked position withdrawal sequence Such a condition is already anticipated by the Technical Specifications (TS 3.1.6, Action A), however, and is considered withm the bounds established by the RWE analysis as di=enM in tha TS Basis. PosMlated system failures in conjunction with a higher drive pressure could result in an unplanned withdrawal at higher than normal (3 in/sec) speeds, however such speeds still remain less than those assumed in the CRDA (UFSAR 4.6.2.3.2.2) and are therefore

%unded by the existmg analysis.

l e

3. May increase the probability of occurrence of a malfunction of equipment O Yes important to safety previously evaluated in the S AR. p No m: No other modifications to equipment or procedures important to safety are being made by this change. Discussions with GE personnel indicate that the system is capable of handling the increased pressure without adverse affects. A higher drive pressure will result in a slightly increased speed of withdrawal or insertion, however this is not considered a " malfunction" since it is expected and can occur even with normal variations in system pressure. Exanunation of the system design also shows that increased drive water pressure is not more likely to result in a unplanned withdrawal or insertion without the presence of additional failures.

4. May increase the consequences of a malfunction of equipment important to O Yes safety previously evaluated in the S AR. J3 No m The FSAR considers several malfunctions of the CRD system (UFSAR 4.6.2.3).

The proposed change potentially impacts only the consequences of an unplanned withdrawal assummg additional failures occur. The consequences of other events are either unrelated or negligibly affected by drive pressure prior to an assumed failure. In the case of an unplanned withdrawal, a slightly faster withdrawal speed could result, however this remams bounded by existing CRDA analyses. l S. May increase the possibility for an accident of a different type than any 0 Yes previously evaluated in the SAR. g No m Increasing drive water pressure to a maximum of 475 psid above reactor pressure does not create the possibility of an accident of a different type. No physical changes are being made to system design, and no additional changes are being made to system operation. The UFS AR already considers numerous malfunctions, ruptures, and multiple j failures of the CRD system and its components This relatively slight increase in allowed i dnvc water pressure under controlled conditions is well within system piping design

) capacity. Any impact on rod mispositioning is already considered as described in B.1 j and B.2 above.

l 6. May create the possibility for a malfunction of equipment important to safety of O Yes

a different type than any previously evaluated in the S AR. p No m As described in B.5 above, the UFSAR already considers a large spectrum of events related to CRD system malfunctions. No additional malfunctions which could be created by an increased drive flow could be identified.

-- --. . - . . . - . . - - - . - . .- - . . .-- - - . . . ~ - . ~ ~

i l

• l i'

4 7. Will reduce the margin of safety as defined in the basis for any Technical O Yes Specificanon. y No l

- BASIS No margm of safety is irnpacted. With the restriction placed on excessive l

pressures above the HPSP, there is no danger of exceedmg the MCPR safety limit or the  ;

i 1% plastic strain limit discussed in the basis for TS 3.3.2.1. The 280 cal /gm limit

related to the CRDA event di=~ sed in TS 3.1.6 basis is likewise not threatened by the

! proposed change since do change is being made to existing rod pattem controls or l l associated Technical Specifications. Also, control rod scram ability and times will not )

a be affected by the proposed cha1ge so assumed reactivity insertion margms remam the same.

1 1

j There is no evidence that the NRC made any assumptions regardmg dnve water 1

differential pressure in the conclusions described in the GONS SER. One mention of an j j NRC concern regardmg failure effects on drive water pressure and resulting speed (SER, j Sect. 4.0) is resolved in the UFSAR analysis. '

l O Not applicable pu Envuonmemal Evaluanon l IIL EnvironmentalEvaluation Applicability Review IMPt.EMMrTATION OR PERFORMANCE OF THE ACTION DESCRIBED IN T1!E EVALUATED DOCUMENT A. Environmental Protection Plan

1. Will require a change in the Environmental Protection Plan. O Yes g No gasIs The CRD system is not discussed in the EPP.

B. Unreviewed Environmental Ouestion

1. Concerns a matter which may result in a significant ir.crear.e in any adverse O Yes environmental impact previously evaluated in the Final Environmental Statement (FES) as modified by the NRC staffs testimony to the Atomic y No I Safety and Licensing Board (ASLB), supplements to the FES, environmental impact appraisal, or in any decisions of the ASLB.

3Asm: The proposed change has no potential impact on the environment in excess of that already &miin the FES.

2. Concems a significant change in efBuents or power level.. O Yes

% No BAsg: No change to effluents or power level. .

3 Concems a matter not previously reviewed and evaluated in the documents O Yes speci6ed in II.B.1 above, which may have a significant environmental No impact.

1 36313: The proposed change does not present a possible significant enytronmental impact.

J

' Signatures and Approvals Evaluated- , ., p [f [

ORIGNATOR/ DAT/ /

Reviewed / Approved: A f (M REVIEWER / DATE 8-/-9f l

Plant Safety Review Committee Review l

t I in h -l~ d THAI , PSRC / DATE l

I l

I.

i

+

i i

C'5.[o(- %

GRAND GULF NUCLEAR STATION UNIT 1 CHANGES, TESTS OR EXPERIMENTS SAFETY AND ENVIRONMENTAL EVALUATION FORM L Safety Evaluation Overview A. Reference Data ORIGINATOR: J M.LASSETTER DEFr SECT: CHEM EVAL. #: 95-0078-R01 DOCUMENT EVALUATED: TRM c/r 95-071

REFERENCES:

TRM LCO 6.3.10. 6.3.11. NUREG 0578. 0737. REG GulDE 1.97. UFSAR CHAPTER 11.15.18. LICENSING RESEARCH SYSTEM (LRS). FINAL ENVIRONMENTAL STATEMENT.

FSAR CHANGE REQUIRED 7 / Yes O No CR# 95-072 FSAR SECTIONS TO BE REVISED: 11.5.2.2.4.1(editorial chance)

TRM CHANGE REQUIRED? / Yes O No TECH. $*EC. CHANGE REQUIRED? O Yes / No CR # . CR or (n/a)

IS THE V 'LIDirY OF THIS SAFETY EVALUATION DEPENDENT ON ANY CHANGES OTHER THAN

/ Yes THE CHA<GE BEINO EVALUATED (E.O. PROCEDURAL, OPERATIONAL CONDir!ONS)? O No IF YES TO THE LAST QUESTION, HAVE THE ORGANIZATIONS RESPONSIBLE FOR THOSE yg CHANGES 3EEN NOTIFIED 7 THE RESPONSIBLE ORGANIZATIONS MUST BE NOTIFIED PRIOR TO IMPLEMENTING THIS CHANGE.

B. Executive Summarv (ALSO SERVES AS INPIR TO NRC

SUMMARY

REPORT)

P AIEF DESCRIPTION OF CHANGE, TEST OR EXPERIMENT: TRM LCO 6.3.11 Action 9.1 is beina deleted. This reauires re-numberina action. "B.2". to "B.1" and chancing the referenced Action in Condition "C" from "B.2" to "B.1" Existina ActioQ1 reauires. within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. imolementation of ore-olanned attemate metnod of monitorina the accrooriate parameters when the sccident ranae (AXM) noble aas radiation monitor function is inocerable.

This action may be deleted because it is bounded by a more limitina TRM LCO . TRM LCO 6.3.10 reauires continuous monitorina of caseous radioactive effluents.

contains comoensatory measures for inocerable eauioment and apolies to the same oathways and carameters (functions) desenbed in the action to be deleted (TRM LCO Acy.,n 6.3.11. Action 8.1). TRM LCO 6 3.10 reauires compensatory actinns (arab samplina and analysis) in a shorter time frame ( four or elaht hour sarm e collection. analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) than TRM LCO 6.3.11( 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).

In summary. the orcoosed activity is deletion of a redundant reauirement. TRM LCO 6.3.10 Actions B.and G. reauire responses which are adeauste for oreolanned.

i alternate monitorina methods. An editorial UFSAR chanae will be submitted to specify that provisions exist for collection of orab samples in the event of a loss of Accident Ranae Monitorina instrumentation (UFSAR c/r 95-072).

SE 95-0078 R01 Page 1

, i 1

REASON FORCHANGE TEST OR EXPERIMENT: A redundant TRM LCO action exists l (6.3.11.B.1). Other TRM reauirements (LCO 6.3.10 Action B.1.B.2 and G.1. G.2. l I

orovide more restrictive comoensatont measures which meet the recuirements of the deleted action.

The reouirement to continually monitor caseous radioactive effluents is indeoendent I of accident or non-accident conditions .

SAFETY EVAwATION

SUMMARY

AND CONCI.USIONS: The crocosed chance will not result in a chanae to the Radioloaical Effluent Controls Procram as recuired by Technical Soecification 5.5.4.

The associated D17 Noble Gas Effluent monitors (Ebertine SPING for S9GT-A & B.

CMeral Electric for OffGas/RadWaste Containment. Turbine and Fuel H andlina) are always the ore-olanned attemate method of monitorina provided they gg i 92grrble and onscale. If these monitors are inocerable. the ore-olanned, altemmaves are procedurally in olace as recuired by TRM LCO 6.3.10. The ore- i olanned altemative monitorina methods are in fact always in effect and cassive in (

nature. No additional actions are recuired.

Eggylations reauirina accident ranae noble cas monitorina instrumentation are ,

I based on inadecuacies identified followina the TMI-2 accident. The reculations '

reviewed for this safety evaluation contain soecifications for installed noble cas effluent monitors but do not delineate actions reauired when those instruments are inocerable. Obviousiv. Deriods of inoperability are exoected to be minimized, The TRM recuirement to retum noble aas accident rance monitorina channel (s) to OPERABLE status within seven days and associated reportina orovisions are unaffected by the orooosed chanae.

II. Safety Evaluation O Not applicable per Safety Evaluation Applicability Review A. Technical Soecifications

1. Implementation or performance of the action desenbed h the evaluated O Yes document will require a change to the GGNS Unit i Technical / No Specifications.

BASIS: The affected recuirements are not in the Technical Soecifications. No chance to the Radiolooical Effluent Controls Proaram as described in the TS will result from the crocosed chance. The affected reauirement alternate. ore-olanned methods of monitorina of radioactive noble aas effluents when accident ranae Instrumentation is inocerable. will utilize existin2 orocedures which fulfill the commitment for routine monitorina of radioactive noble cas effluents.

2M_9BM&R03__Pem 2

l

, l

. l

. I i

l B. Unreviewed Safety Ouestion IMPLEMEbrTATION OR PERFORMANCE oF THE ACTION DESCRIBED IN 1EE EVALUATED DOCUMENT.

1. May increase the probability of occurrence of an accident previously evaluated O Yes in the S AR.

/ N'o  :

aggi: The orooosed activity has no effect on the orobability of accidents as described in Chaoter 15 of the UFSAR. Accident orobability is determined in oart by the operatina characteristics of installed

. eauioment and the procedures and oroarams used to maintain the eauioment. Installed eauioment is unaffected by the orocosed activity as the proposed activity is limited to comoensatorv measures for l inocerable caseous effluent monitorinu wuuis.nont.

2. May increase the consequences of an accident previously evaluated in the SAR. O Yes 4

/ No  :

' 1 aggi Radioloaical consecuences from oreviousiv evaluated accidents are not affected by the proposed activity. The comoensatory actions l which result from inoperable eauioment continue to ensure adeauate l' monitorina cacability of noble aas effluents for lona term assessment of 4

Offsite releases. Radioloaical source terms are not affected nor are j existino marains of safety for any olant structures. systems or l comoonents.

GGNS UFSAR Section 11.5.2.2.4.1 states that in the case of inocerable j

asseous effluent monitorina eauioment. orovisions have been made to i obtain arab sarroles for laboratory analysis. The orocosed chance does not affect the actions which fulfill this commitment.

Section surveillance orocedures for inocerable affluent asseous radiation monitors will include soecific actions and orecautions reaardina arab samole collection of noble cas samoles under accident conditions.

3. May increase the probability of occurrence of a malfunction of equipment O Yes important to safety previously evaluated in the SAR. / No SE 95 0078 R01 Page 3

a

{4 j .

i I m The proposed chanae is restricted to comoensatory measures for l

ingperable sowom9nt . These measures are independent of souloment ,

i maintenance , calibration. and surveillance activities . therefore the l

! orobability of that eauioment malfunctionina cannot increase as a result i of the orooosed chanae. Accident Ranae monitorina is based on

! NUREG 0737 reauirements. USFAR Section 18.1.27.1 contains GGNS' 1 I resoonse to the Noble Gas Monitorina reauirements of NUREG 0737.

i 4

The orooosed chanae does not affect GGNS' resoonse.

l 4. May increase the consequences of a malfunction of equipment important to O Yes j safety previously evaluated in the S AR. / No 1

) m The radioloaical conseauences of eauioment malfunction will not be affected by alternate monitorina methods orovided for inocerable l

j

. eauioment. The orocosed chance is indeoendent of eauioment

ooeratina. maintenance and calibration orocedures. The reauirement to I i

restore inocerable channel (s) to OPERABLE status with seven days (TRM LCO 6.3.11. Action B.2) and subsecuent reoortina reauirements are unaffected by the orooosed chance.

5. May increase the possibility for an accident of a different type than any C Yes  ;

previously evaluated in the SAR. / No s The orooosed activity is indeoendent of accident tvoes. The grggsed chance will not result in a chance to olant confiauration or oceistina orocedures . The orocosed chance is restricted to methods for 1 monitorina noble cas releases when accident rance monitors are j inocerable.

6. May create the possibility for a malfunction of equipment important to safety of C Yes j a different type than any previously evaluated in the SAR. / No m The orocosed chanae will not affect olant Structures Systems or Comoonents. No additional functions are reauired as a result of the J I

orocosed chanoe. Eauioment will not be operated in a manner other than oreviousiv described in the UFSAR.  ;

7. Will reduce the margin of srfety as defmed in the basis for any Technical O Yes l Speciscaten. / No i m The orooosed activity wijl not result in chances to operatina Darameters as=^ciated with olant structures. systems or components. I The orocosed chanae is restricted to actions reauired to comoensate for l inocerable accident ranae eauioment and will utilize existina  !

methodoloav. Reauirements to return channel (s) to OPERABLE status ,

are unaffected by the orooosed chance.

J l

IIL Environmental Evaluation O Not applicable Per Environmental Evaluation Applicability Review l IMPLEMENTATION OR PERf0RMANCE OF THE ACTION DESCRIBED IN THE EVALUATED DOCUMENT:

SE 95 0078.R01 Page 4

. s F

A. Environmental Protection Plan

1. Will require a change in the Environmental Protection Plan. O Yes

/ No m: The orocosed activity is restricted to existina eauioment and orocedures and will not affect the EPP. The EPP deals with non-radioloaical issues. the orocosed activity is limited to monitorina of radioloaical effluents. .

1 B. Unreviewed Environmental Ouestion

1. Concems a matter which may result in a significant increase in any adverse O Yes environmental impact previously evaluated in the Final Environmental / No Statement (FES) as modified by the NRC staffs testimony to the Atomic i Safety and Licensing Board (ASLB), supplements to the FES, environmental impact appraisal, or in any decisions of the ASLB.

m The orooosed activity is a clarification of methods used to comoensate for inoperable accident rance effluent noble cas monitors, Existina TRM LCO actions for routine effluent monitorina renukg oeriodic collection of arab samoles uoon loss of continuous monitorina cecability. These same actions. with consideration for aqcident rance release rates are adeauate for inocerable accident rance asseous effluent monitors. The FES does not contain soecific reouirements for comoensatory monitorina methods. FES section 4.2.5 does state that routine measurement of all orincioal release points is reauired. The orocosed chance does not contradict this FES statement.

2. Concems a significant change in effluents or power level.. O Yes

/ No e The orocosed activity will not increase emissions from GGNS. The proposed chance is limited to monitorina of radioactive effluents and does not affect carameters associated with ceneration of those effluents.

No reduction in measurement accuracy or increase in samolina ceriodicity will result from the orocosed activity. The nature of the orocosed activity excludes any effect on oower level.

3. Concems a matter not previously reviewed and evaluated in the documents O Yes specified in II.B.1 above, which may have a significant environmental / No impact.

m: The crocosed activity is restricted to oreviously d@jurbed areas The orooosed activity is limited to exitino orocedurer and eauioment. No environmental imoact will result from the orooosed activity.

Signatures and Approvals cemussuutsvrts

6 ,.'

Evaluated: b" d MM ORIGINATOR / DATE T M.T f Reviewed / Approved: h fM REVIEWER / DATE 9-40-95 l

i Plant Safety Review Committee Review l

1 l

M fY$f CHAIRMAN, PSl(C / D/rE l

l l

\

l l

l i

I i

SE 95 0078 R01 Page 6 i

i e

i 1

_ __ _ . , _ . .