GNRO-2010/00058, Report of 10CFR50.59 Evaluations and Commitment Changes - July 1, 2009 Through June 30, 2010

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Report of 10CFR50.59 Evaluations and Commitment Changes - July 1, 2009 Through June 30, 2010
ML102300239
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 08/17/2010
From: Perino C
Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2010/00058
Download: ML102300239 (30)
v • d • e


Text

Entergy Christina Perino GNRO-201 August 1 2010 u.s. Nuclear Regulatory Commission Attn: Document Control Washington, DC -_".1_-

SUBJECT:

Report of 10CFR50.59 and Commitment Changes -

July 1 2009 through June 2010 Grand Gulf Nuclear Station, Unit No.1 Docket No. 50-416 Il"'l"'\nlt:"1"'\ No. NPF-29

Dear Sir or Madam:

Pursuant to 10CFR50.59(d)(2) Entergy Operations, Inc. hereby submits a description of 50.59 and commitment change evaluations for the period of July 1, 2009 through June 30, 2010 in Attachment 1. Attachment 2 contains copies of the full evaluations.

If you have any questions or require additional information, please contact Dennis Coulter 601-437-6595.

This letter does not contain any commitments.

Sincerely,

~~.~~

CLP/DMC:dmc Attachments: 1. Table of Contents

2. 10CFR50.59 Evaluations and Commitment Change Evaluations cc: (See Next Page)

cc:

u.s. Nuclear Regulatory Commission ATTN: Mr. Jr. (w/2)

Administrator, Region IV 2 Blvd, 400 Arlington, TX 76011-41 U. Nuclear Regulatory Commission ATTN: Mr. lyon, NRR / DORl (w/2)

Mail Stop OWFN/8 B1 Washington, DC 20555-0001

Attachment 1 Table of Contents Grand Gulf Nuclear Station 10CFR50.. 59 Evaluation and Commitment Change Evaluation Report for the Period July 1,2009 through June 30,2010 10CFR50.59 Evaluations Evaluation Number Initiatin Document Descri tion 2009-0001-ROO Engineering Change 141 implements revised offsite and control 14143 room atmospheric dispersion coefficients developed using the latest five years (2002-2006) of meteorolo ical 13138 Evaluation 0 of Noble etals GGNS Licensing Basis This extends the Technical Requirements Document Change Manual surveillance frequency for the Horizontal Re uest 2010-013 Fuel Transfer S stem from 7 da s to 31 da s.

Commitment Change Evaluations Commitment Source Document Description Number N/A N/A There were no commitment change evaluations for the period that required NRC notification.

Attachment 2 10CFR50.59 Evaluations and Commitment Change Evaluations

GGNS 50.59 Evaluation Number SE 2009-0001-ROO

10 I.

IIICIIIly: GONS Iv. . . . . ' I Rev. t: 200H001 Rev. 00 PIopoaed CIwJge I DocwIwtt: Ie 14143 DMoe IptIon of ChMgrI: EC 14143 inplementl revised offalte and control room atmosptwlc dl8pers1an caefflcIents (xIQ'1) developed UIlng the latest ftve YN'I (2002-2008) of ... metearaloglcal data. The daM

_ at the Lola of Coolant Aacklent (LOCA), Cantral Rod Drop AcckIent (C tA), Fuel HIndIfng AccIdent Steam Une Break OUtsIde ContaInment (M8L8OC), and 0ffSJM Syetem Falure *

  • reviled to reItIct 1he impact d .,.w cIIIpenJIon ooeffIctentL Affected daM calcutallonl . .: Xe-a1111-98019 (FHA), Xe-a1PB3-05011, XC-01 M48-04OO4. and Xe-Q1111-98017 (lOCA). XC.Q1 N11-14OO4 (MSLSOC).

_Iv Xc-Q1N84-88004 (0ffgM System Fdul8), and XC..Q1111*98018 (CADA).

I." vMIdIIr of dIpendInt on anyottw chlnget 0 V.. 181 No If-V. .- ~ The chMgMCOftNd byth.. IoM C8llftGlbe

_ IIfIPIOWII'of. . ottw IIIed ( .......,.

e............ appropriate notIfIoIItIan to _ .. oompl d.

OSRC MeetIng I t s~ may be cbtahId via eIectronJo proc88M8 (e.g** PeRS. ER pIOC8I888), ~ methoda (e..g., ink 8ignaIure).

e-maI. or _ . If using an e-mail 01 telrlcarrtmuNcda. aa.:h I to tNs 'ann.

EN-l1-101-ATT-9.1. Rev. 5

10 C EVALUAnON FORM Sheet 2 of S II.

Does the Mel Change being evaluated represent a change to a method of evaluation

? If MV..." Queetlon.1 -7 are not applicable; anewer only Queatlon 8. If "No," anewer 0 Ves all queatlona below. ~ No Does the propoMel Change:

1. R ult In more than a minimal increase in the frequency of occurrence of an accident 0 Ves previously evaluated in the UFSAR? ~ No B

This activity implements revised atmospheric dispersion coefficients by revising the calculations and dose consequences associated with a Loss of Coolant Accident (LOCA). Control Rod Drop Accident (CRDA).

Fuel H ing Acci A), Steam e Break Outside Containment (MSLBOC). and Offgas System This activity does not affect or physically change any structure, system. or component (S and does not affect how any C Is controlled or operated. Since this activity only affects the evaluations of accidents, the activity will not impact the frequency of occurrence of these accidents or any accident previously evaluated in the FSAR.

2.. Result In more than a minimal increase in the likelihood of occurrence of a malfunction of a 0 Ves structure, system, or component important to safety previously evaluated in the UFSAR? ~ No BASIS:

This activity implements revised atmospheric dispersion coefficients by revising the affected design basis dose calculations and dose consequences. This activity does not physically change any structure, system, or component ( C), does not affect how any SSC is controlled or operated, and does not change any plant procedures.. Therefore, this activity does not affect the cause or mode of a malfunction of any sse and does not increase the likelihood of occurrence of a malfunction of a SSC important to safety preViously evaluated in the FSAR.

3. Result in more than a minimal increase in the consequences of an accident previously 0 Ves evaluated in the UFSAR? ~ No BASIS:

The dose consequences of the accidents impacted by the application of the revised atmospheric dispersion factors are documented in the FSAR. This activity revises the affected accident analyses to include the revised dispersion coefficients. The revised dispersion coefficients are generally higher than those applied in the current dose ulations. Since dose is proportional to the magnitude of the dispersion coefficient, the dose consequences calculated in the revised analyses increase. The revised accident doses are compared to the doses from the current calculations reported in the FSAR in the table below.

EN-LI-101-ATT-9.1, Rev. 5

10 CFA 50.68 EVALUA110N FORII Sheet 3 of 6 Dos. (Flem TEDE)

Results

  • Current Applicable Location Revised Results Analysis Rqulatory Umlt Fuel Handlin. Accident (XC-Q1111- )

Exclusion Area Boundary 2.565 2.642 6.30 Low Population Zone N/A* N/A*

Control Room 2.648 2.804 5.0 Control - . - vrp Accident (XC-Q1111-98016)

Exclusion Area Boundary 0.147 0.151 6.3 Low Population Zone 0.064 0.072 6.3 Control Room 0.262 0.262 5.0 Main Steam Une Break Outside Containment (XC-QIN11-94004)

Equilibrium Iodine Exclusion Area Boundary 1.1905E-Q1 1.2262E-Ol 2.5 Low Population Zone N/A* N/A* N/A*

Control Room 1.5345E-Q1 1.S345E-Q1 5.0 Iodine Splicing Exclusion Area Boundary 2.32 2.39 .25.0 Low Population Zone N/A* N/A* N/A*

Control Room 3.01 3.01 5.0 L~(XC-Q1111-98017,XC~lP53..Q5001,XC-Q1M~)

Exclusion Area Boundary 8.45 8.70 25.0 Low Population Zone 4.56 5.15 25.0 Control Room 3.69 3.69 5.0 otrsas System Failure (XC-Q1N64-98004)

Exclusion Area Boundary 1.45 1.69 6.25 Low Population Zone 0.325 0.385 6.25 Control Room 0.124 0.124 5.0

  • The LPZ dose is not evaluated in these analyses because the LPZ dose is bounded by the EABdose.

A "minimar increase in consequences is defined as 10% of the difference between the current calculated dose value and the regulatory limit (Ref. NEI96-D7, Rev. 1). As shown below, the increases in dose consequences in all accidents at all dose locations are less than the "minimar increase. Therefore, this activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR.

EN-ll-101-ATT-9.1, Rev. 5

10 CFR 50.59 ALUATION FORM Sheet4ot5 Dose (Rem TEOE) n1M1111 main location Actual Increase Increase Fuel Handlln, Accident (XC- 111-98019)

Exclusion Area Boundary 0.374 0.077 Low Population Zone N/A N/A Control Room 0.235 0.156 Control Rod 0 ...... nt (XC-Q1111-98016)

Exclusion Area-8o'Jndary 0.615 0..004 Low Population Zone 0.624 0.008 Control Room 0.474 0.0 Main Steam Line Break Outside Containment (XC-QINll-94004)

Iodine Exclusion Area Boundary 2.381E-01 3.570E-03 LowP n Zone N/A N/A Control Room 4.847E-01 0.0 Iodine Spiking Exclusion Area Boundary 2.27 0.07 Low Population Zone NIA NIA Control Room 0.199 0.0 LOCA (XC-Q1111-98017, XC-Q1P53-GS XC-Q1M46-04004)

Exclusion Area Boundary 1.655 0.25 Low Population Zone 2.044 0.59 Control Room 0.131 0.0 Offps System Failure (XC-Q1N64-98004)

Exclusion Area Boundary 0.48 0.24 Low Population Zone 0.592 0.060 Control Room 0.488 0.0

4. Result In'more than a minimal increase in the consequences of a malfunction of a structure, 0 Yes system, or component important to safety previously evaluated in the UFSAR? 181 No BASIS:

This actMty implements revised atmospheric dispersion coefficients by revising the affected design basis dose calculations and dose consequences. This activity does not physically change any structure, system, or component (SSC) and does not affect how any SSC is controlled or operated. Therefore, this activity does not affect a malfunction of any sse and does not change the consequences of any EN-LI-101-ATT-9.1. Rev. 5

AnONFoFIII Sheet&of&

malfunction of a SSC important to safety previously evaluated in the FSAR.

5.. Create a possibility for an accident of a different type than any previously evaluated in the DYes U  ? 181 No BASIS:

T *vity Implements revised atmospheric dispersion coefficients by revising the affected design basis dose calculations and dose consequences. This activity does not phys* Iy change any structure, system, or component (SSC), does not affect how any SSC is controlled or operated, and does not change any p procedures. Therefore, this a ity does not create the possibility for an accident of a different type than any previously evaluated In the FSAR

6. Create a possibility for a malfunction of a structure, system, or component important to safety DYes with a different result than any previously evaluated in the UFSAR? 181 No BASIS:

This Ity Implements revised atmospheric dispersion coefficients by revising the affected design basis dose calculations and dose consequences. This activity does not physically change any structure, system, or component (SSe), not affect how any SSC is controlled or operated, and does not change any plant procedures. This activity does not affect a malfunction of any sse described in the FSAR. Therefore, this activity does not create a possibility for a malfunction of a SSC important to safety with a different result than any previously evaluated in the FSAR.

7. Result In a design basis limit for a fission prodUct barrier as described In the UFSAR being DYes exceeded or altered? 181 No B S:

This activity revises the design basis dose calculations and dose consequences. This activity makes no physical changes to any sse or assumptions regarding the operation of any sse Important to safety.

Therefore, this actl does not affect a design basis limit for a fission product barrier (fuel cladding, reactor coolant system boundary, and containment) and will not result in a design basis limit for a fission product barrier described in the FSAR being exceeded or altered.

8. Result in a departure from a method of evaluation described in the UFSAA used in establishing 0 Yes the design bases or in the safety analyses? 181 No BASIS:

The GGNS design basis dose analyses use analytical and numerical (AAPTOR) methods to calculate the accident dose consequences. However, the dose analysis methodology is not explicitly described in the FSAR in any detail. Atmospheric dispersion coefficients are input parameters to the dose calculations but are not part of the methodology. The revised atmospheric dispersion coefficients were developed in accordance with NRC uirements using methodology preViously approved by the NRC. Since dose is directly proportional to the dispersion coefficient (X/a), the doses were revised by multiplying the doses calculated using the current methodology by a simple ratio of the revised X/O to the X/O applied in the current analyses. Therefore. the revised calculations do not apply a new analysis methodology or change an element of the methodology such that this activity does not represent a departure from a method of evaluation described in the FSAR.

It any of the above qU88tlons Is checked "Yes," obtain NRC approval prior to Implementing the change by Initiating a change to the Operating License In accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATr-9.1, Rev. 5

GGNS 50.59 Evaluation Number SE 201 0-0001-ROO

I.

Grand Gulf Nuclear Station tG(;N~51 pr(~DO'S8,a VIIGIIY_ / Document: Evaluation # I Rev. #: 00 Revision 0 NMCA Evaluation: Evaluation of the Use of Nobel Metals at GGNS De~scrilDtlon of \.;nclnge:

t:nCJlntter'lng VII4C1II~I. EC-13138 the Technical and Basis for the GE-Hitachi On-Line NobleChem 1M (OLNC) process for at GGNS. The OLNC process will be used in COlnJunc~tlo,n with the Water (HWC) process for controlling intergranular stress corrosion The OLNC process provides a environment that promotes the rec:orrlDlnlnlg of and oxygen in the reactor and recirculation lines. The catalytic nature of the OLNC process promotes the usage of lower hydrogen rates when compared to HWC alone.

The OLNC process 1) a more effective for IGSCC than the traditional 16 HWC process alone and reduces the main steam line radiation levels from volatile N components.

The OLNC process utilizes the noble metal compound sodium hexahydroxyplatinate [Na2Pt(OH)6]. The compound Is Into the reactor the Feedwater during plant operations. The ionizing characteristics of the cause it to breakdown into the ionic forms of Platinum (Pt), Hydroxide (OH), and Sodium The platinum ions out onto the internal reactor components and the reactor recirculation line a thin film (a few atoms thick) on the metallic surfaces. The platinum causes a reaction between the free hydrogen and oxygen molecules in the reactor coolant.

The catalytic reaction encourages the chemical reaction of hydrogen and oxygen to form water molecules. (The reaction reduces the need for an excessive amount of hydrogen to react with the oxygen. The ratio of hydrogen to oxygen is sUg htly higher than the stoichiometric molar ratio of 2.)

The reduction of oxygen in the reactor coolant lowers the electrochemical corrosion potential of the coolant to levels that mitigate IGSCC..

16 The OLNC process will also reduce the amount of radionuclide N in the main steam lines. The 16 nuclide N 16 is produced in the reactor core when oxygen is radiated: 0 + n -J- N + H1* The N16 16 combines with the free hydrogen to form the volati Ie compounds NH3 and NH4 and then carried over to the steam lines where it decomposes. The magnitude of HiS in the steam lines is dependent on the Sinn$:ltHrf=~S may be obtained via electronic processes PCRS, ER proces.se!;), manual methods e-mail, or telecommunication. If using an e-mail or telecommunication, form.

EN-LI-1 01-ATT-9. 1 Rev, 5

amount of free hydrogen injected into the reactor vessel from the Hydrogen Chemistry System. The 18 reduction of free hydrogen levels In the reactor coolant will produce lower N levels in the main steam lines.

The equipment and SSC interfaces (i.e., the civil, electrical, and mechanical interfaces) for the OLNC process have been installed In the plant by other ECs. The design and licensing for the equipment and i .... ~!'!Il. . . *'!a.... Jll!Ik. are addressed in the individual ECs. The ECs are identified below:

EC-13132, Mitigation Monitoring System Installation EC-13133, On-Line Noble chemistry Injection Skid Installation EC-13134, Feedwater Tap installation EC-13135, RWCU Tap Installation EC-13136, Hydrogen Water Chemistry System program Changes EC-13137, Physical Modifications to the Hydrogen Water Chemistry System Is the validity of this Evaluation dependent on any other change? Ves No If "Ves," list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other Identified changes (e.g., license amendment request).

Establish an appropriate notification mechanism to ensure this action is completed.

Based on the results of this 50.59 Evaluation, does the proposed change 0 Ves ~ No require prior NRC approval?

Preparer: Jeff Farnsworth / ENERCON Services / Mechanical Engineering I 2 - /~ - 20 10 Name (print) I Signature I Company / Department / Date kC.f;:-e Reviewer: See lAS I ENERCON Services I Mechanical Engineering I 2- 17- 2 0 I 0 Name (print) I Signature I Company / Department / Date OSRC:

Chairman's Name (print) I Signature / Date OSRC Meeting # 00 ~ .. .:zo/O EN-LI-101-ATT-9.1, Rev. 5

II. 50.59 EVALUATION Does the proposed Change being evaluated represent a change to method of evaluation ONLY? If Questions 1 - 7 are not applicable; answer only Question 8.. If "No," answer all questions below. Yes No Does the proposed Change:

1. Result In more than a minimal increase in the frequency of occurrence of an accident Yes previously evaluated in the UFSAR? No BASIS:

The addition of noble metal compound and the OLNC process do not adversely affect any equipment treated by the OLNC process. The frequency of occurrence of accidents evaluated in the UFSAR is not related to the ability to mitigate IGSCC in the vessel or associated systems. The installation, testing, and operation of the OLNC system does not impact the ability of any structure, system, or component to perform its safety function. Core cooling mechanisms, reactivity control features, or pressure control systems are not impacted. There are no changes to operating procedures which would increase the frequency of occurrence of a previously evaluated accident. None of the precursors to a loss of coolant accident or other accidents in the UFSAR are impacted by this change.

The noble metal compound has no impact on the carbon steel main steam lines, and thus, the frequency of occurrence of a main steam line break (MSLB) in not changed from that previously evaluated in the UFSAR. The control rod drive equipment will continue to be operated and maintained within design specifications, and thus, the frequency of occurrence of a control rod drive accident is not changed from that previously evaluated in the UFSAR..

Neither the installation nor injection processes Involves movement of fuel In the core or fuel pools nor do they modify any fuel handling equipment, thus the frequency of a fuel handing accident will not change.

Therefore, noble metal application with OLNC does not result in more than a minimal increase In the frequency of occurrence of an accident previously evaluated in the UFSAR.

EN-LI-101-ATT-9.1, Rev. 5

2. Result in more than a minimal in the likelihood of occurrence of a malfunction Yes of a structure, system, or component important to safety previously evaluated in the No UFSAR?

BASIS:

In combination with Hydrogen Water Chemistry, the OlNC process mitigates IGSCC and has no known detrimental effects. The likelihood of occurrence of a malfunction of the treated components in the reactor vessel and associated piping will not increase. The effects on equipment important to safety covered in the OLNC process will be negligible because the total thickness of multiple applications of noble metal compound is on the order of a millionth of an inch. The loading and operating condition (e.g.. , stress, pressure, environment, flow, and temperature) of affected SSCs will not change. Extensive fuel surveillances and examinations conducted by the OLNC vendor (General Electric-Hitachi's (GEH) "On-line NobleChem' Application Technical Safety Evaluation for Grand Gulf Nuclear Station", Rev.. 0, Nov.. 2009) confirm that the likelihood of fuel cladding failures is not Increased, nor are the chances of further degradation of already failed fuel.

The ionizing characteristics of the noble metal compound will cause an increase in the reactor coolant conductivity. The increase in conductivity will be very small and has no adverse impact on water chemistry. Coolant conductivity is monitored per Plant procedures and the limits established by the procedures are not affected by the OLNC process.

The very small amount of noble metal compound causes no reduction in the plant nuclear safety functions evaluated in the UFSAR. Due to the very small amount of material deposited on equipment surfaces, there is no increased likelihood of interference causing an SSC malfunction.

Thus control rod blade function and channel flow impacts are considered negligible. During the OLNC injection processes, plant operations will not be significantly affected. No initiators for anticipated operational occurrences (transients) in the UFSAR are affected. No changes to the transient analysis inputs are made.

The OlNC process will not degrade any SSC including those important to safety. Therefore, the addition of noble metal compound by the OlNC process does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the UFSAR.

EN-LI-101-ATI-9.1, Rev. 5

3. Result in more than a minimal increase in the consequences of an accident previously Ves evafuated In the UFSAR? No BASIS:

The noble metal compound does not affect the amounts or types (source terms) of radioactive materials assumed to be released in any UFSAR accident analyses. The OLNC process has no adverse impact on the boron moderator or reactivity control. The OLNC process has no adverse impact on the current type of fuel assemblies or future fuel assembly designs. The OLNC process does not adversely affect any engineered safety feature assumed to function in the UFSAR accident analyses, and thus, the plant's accident mitigation functions are not affected. The addition of the noble metal compound does not affect the peak enthalpy used in the design criteria of the control rod drop accident. The radioactive source terms for the accidents (MSLBA, LOeA, FHA, and CRDA) analyzed in the UFSAR are not affected by the noble metal compound.

Vendor evaluations (as provided in General Electric-Hitachi's (GEH) "On-line NobleChem' Application Technical Safety Evaluation for Grand Gulf Nuclear Station", Rev. 0, Nov. 2009) indicate that platinum particle deposition on fuel surfaces will have a negligible impact on the fuel crud layer and Its associated heat transfer characteristics, thus the peak cladding temperature reached during a LOCA will not increase. Further, the assessments in the above vendor evaluation conclude that the inherent zirconium oxide layer on the fuel surfaces from the manufacturing process will inhibit any possible catalytic affect which could increase the Zr-Oxygen reaction and associated hydrogen generation.

Per the vendor evaluation above, which includes results of a multi-cycle ONlC surveillance program, the safety related functions of fuel cladding to maintain fuel geometry and fission product retention capability are not affected. The amount of noble metal deposited is extremely small relative to the existing crud layer and does not significantly impact the heat transfer characteristics either during normal operation or post-accident. Therefore, OLNC does not result in more than a minimal increase in the consequence of an accident previously evaluated in the UFSAR.

EN-LI-101-ATT-9.1, Rev. 5

4. Result in more than a minimal increase in the consequences of a malfunction of a 0 Yes structure, system, or component Important to safety previously evaluated in the No UFSAR?

BASIS:

In combination with HWC, coated with the noble metal compound help to mitigate IGSCC.

The effect on equipment important to safety is negligible, because the total thickness of multiple applications of noble metal compound is on the order of a millionth of an inch. The loading and operating conditions (e.g., stress, pressure, environment, flow, and temperature) of affected SSCs will not change. The use of noble metal compound has no effect on plant nuclear safety functions or radioactive material sources evaluated in the UFSAR" During and after OLNC application, plant operations and SSCs are not adversely affected by the noble metal deposit. Fuel cladding is not detrimentally affected, thus the consequences of a fuel failure are no more severe than before OLNC Implementation.

Because noble metal compound does not degrade any SSC involved in the process, including those important to safety. The platinum ions plating out on the reactor surfaces can be activated by thermal neutrons but the amount of platinum in the reactor is negligible when compared to the source terms postulated for the UFSAR accident analyses. The radioactive material sources assumed to be released in the safety analyses are not affected. Therefore, the noble metal compound and the OLNC process do not result in an increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR..

EN-LI-101-ATT-9.1, Rev. 5

5.. Create a possibility for an accident of a different type than any previously evaluated in Ves the UFSAR? No BASIS:

The noble metal compound and the OLNC process do not adversely affect the reactor or plant ancllV!SeS previously evaluated in the UFSAR. Minor changes made to the plant's operating procedures will be necessary for OLNC application. The procedure changes will not affect the initiating events described in UFSAR such as opening, closing, starting, or stopping an evaluated component or system. No new equipment Interaction, accident scenario, or accident precursors will be created by noble metal compound application or the OLNC process described in the UFSAR.. The presence of platinum on reactor fuel surfaces will not initiate a component failure or a system malfunction. The platinum will not Interfere with any pressure boundary component; interfere with core heat transfer mechanisms; alter reactor coolant flow or inventory; or change core reactivity or power levels.. Thus the noble metal compound and the OLNC process do not create a new fission product release path, result In a new fission product barrier failure mode, or create a new sequence of events that results in unacceptable fuel failures or other types of accidents which could result in radiological consequences not evaluated in the UFSAR.

Therefore, the injected noble metal compounds cannot create the possibility of a new type of accident not previously evaluated in the UFSAR.

6. Create a possibility for a malfunction of a structure, system, or component important to 0 Ves safety with a different result than any previously evaluated in the UFSAR? ~ No EN-ll-101-ATI-9.1, Rev. 5

BASIS:

A thin film of platinum molecules will plate onto the surfaces wetted by the noble metal compound..

The OLNC process will be controlled to ensure the platinum film does not exceed the equivalent of 10 IJg/cm2 loading on the fuel assuming that all platinum injected deposits only on fuel surfaces.

Vendor provided procedures will be used to manage the injection of NobleChem into the Feedwater System. The plating of platinum will be monitored by the OLNC mitigating monitoring skid. The platinum film will be a only few atoms thick. (One atom layer is about 10-8 inches thick.)

The total thickness of the platinum film is less than 10-6 inches, which is less than the manufacturing tolerance for the components wetted by the OLNC process. Thus, running clearance between moving parts is not affected by the OLNC process. In addition, the pathways for flow through the reactor core and recirculation piping are not affected by the OLNC process.

The weight added by the platinum film is insignificant compared to the design weight established for each of the reactor components. The application of the OLNC process will have no impact on the structural integrity of reactor components; therefore, such components are no more likely to fail than that previous to ONLC addition.

The deposit of platinum on the fuel assemblies will have no adverse impact on fuel heat transfer rate or the ability of the fuel cladding to resist failure. Platinum from the OLNC process and crud from the reactor coolant will deposit on the fuel assemblies. The normal deposits of crud will overshadow the small amount of platinum that will plate out on the fuel rods.. The thermal conductivity of the crud will dominate the heat transfer characteristics of the deposits.. Minor restructuring of corrosion products on exposed piping and other reactor internal surfaces may result from application of noble metals, however the amount that redeposits onto fuel surfaces is considered negligible considering the inventory of crud on fuel.

Noble metal deposits by the OLNC process do not adversely affect any plant operating condition, nor plant operations with respect to nuclear safety. In combination with HWe, noble metal compound addition has been shown to mitigate IGSee of the treated components in the reactor vessel, and thus the likelihood of occurrence of a malfunction of the treated components in the reactor vessel and associated piping Is not increased. Also, SSC environment, operating rates and loading (e.g., stress or pressure) will not be affected. Thus, there would be no change to the operational status or operational conditions of any sse during or after OLNe process. In addition to the responses to Questions 2 and 4 above, a noble metal application with OLNC process does not change the behavior of any sse import to safety.

Therefore, the OLNe application does not create a possibility for a malfunction of a sse important to safety with a different result than any previously evaluated In the UFSAR.

EN-ll-101-ATT-9.1 Rev. 5

7. Result in design basis limit for a fission product barrier as described in the UFSAR D Ves being exceeded or altered? No BASIS:

The noble metal compound does not adversely affect any fission product barrier such as the reactor pressure piping, and related pressure retaining components.. Vendor evaluations provided in General Electric-Hitachi's (GEH) "On-Line NobleChem ' Application Technical Safety Evaluation for Grand Gulf Nuclear Station", Rev.. 0, Nov. 2009) show that fuel cladding heat transfer characteristics are negligibly impacted by OLNC process. No increase in hydrogen generation will result from platinum catalyst effects. Based on the assessments, operating experience, and performance characterizations obtained to date, the fuel design bases remain applicable and adequate for normal operation, as well as anticipated operational occurrences, and accidents.

Noble metal application does not change or reduce the design basis limit, capability, or the function of the containment and the emergency ventilation system.. No safety analyses inputs are changed and the analyses results in the UFSAR (which demonstrate that design basis limits are not exceeded) are not affected by the noble metal deposit or by the OLNC application process.

Therefore, noble metal application does not result in any design basis limit for a fission product barrier as described in the UFSAR to be exceeded or altered.

EN-LI-101-ATT-9.1, Rev. 5

8. Result in a departure from a method of evaluation described in the UFSAR used in Yes establishing the design bases or in the safety analyses? No BASIS:

The noble metal compound does not adversely affect any treated equipment. such as the reactor pressure piping. and related pressure retaining components. Injection of noble metal does not require changes to any safety related design limit or safety or accident analysis methods or inputs that may affect design basis limits for a fission product barrier as described in the UFSAR.

There are no changes to radioactive release models or assumptions. Assumptions regarding fuel crudding and heat transfer characteristics remain valid.. The design and safety analyses in the UFSAR are not affected by the noble metal deposit or by the OLNC process.

Therefore, noble metal application does not result in a departure from a method of evaluation described in the UFSAR in establishing the design bases or in the safety analyses.

If any of the above questions is checked "Yes," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN-LI-103.

EN-LI-101-ATT-9.1, Rev. 5

GGNS 50.59 Evaluation Number SE 2010-0002-ROO

I.

Facility: Grand Gulf Nuclear Station Evaluation # I Rev. #: . . . ; ; ; -. . . . . . .---.....~ . . . . . . . _

Proposed Change I Document: LeDeR 2010-013-TRM 6.9.7 Description of Change:

ext~enCling the Technical <eaulretmetnts Manual (TRM) 6.9.7 HORIZONTAL FUEL SIIIr\JAlllla"t'A Requirements from 6.9.7.1 through c. from a 1'1"01110,1\/\/ of 7 the following:

a. Room 1A525, ....,.... tll ....... ,.... elevation 182' the room through which the transfer sealed.

b.. All interlocks with the and fuel handling platforms are OPERABLE.

c.. All HFTS carlriacre position indicators are OPERABLE.

An extensive review of HFTS data was performed to determine the feasibility of lengthening the rrec]uelncv of the surveillance This review included HFTS surveillance tests performed dUring and

' .... JCli,nrlll'.3n/ 2004), operations and refueling floor logs dating back to June 2005. A review of CRs the last three refueling outages. Specifically, these records included:

  • Surveillance Procedure 06-QP-1C71-V-0002, Attachment III, Refueling Interlock Check Horizontal Fuel Transfer
  • Shift Management System (Nuclear eSOMS) narrative log entries and LCO tracking log pa(:t<a~:Jes for the completed surveillance procedure were reviewed to check for any unsatisfactory to HFTS. The comment sections of the surveillances were also reviewed fqr applicable comments. Both surveillance tests as well as partial tests (generally for retest following maintenance) were reviewed. This effort revealed no unsatisfactory performances of surveillances associated with TRM SR 6.9.7.1.

Operations and refueling floor were reviewed for HFTS failures or breakdowns due to performance deficiencies with the Room sealing, refueling and fuel handling platform interlocks, or primary carriage position indicators. Any noted fuel handling platform interlock deficiencies were self-revealing and properly caused the system to lock up.

In addition, a review of HFTS operating, design, and licensing information was performed which included the following:

  • Operating Instruction SOI-04-1-01-F11-2, Horizontal Fuel Transfer Mechanism, Revision 111
  • Surveillance Procedure, 06-QP-1C71-V-0002, Refueling Interlock Check, Revision 110
  • Radiation Protection Procedure, 08-S-02-75, Coverage and Control of Refueling Operation, Revision 11
  • Licensing Basis Document Change (LBDC) No. 2005-066, Clarify TRM 6.9.7.1b and UFSAR Chapter/Section 16B.1-200, 9.1.4.2.3.11
  • TRM Bases Section 6.9.7, Horizontal Fuel Transfer System
  • UFSAR 3.2, Table 3.2 System Qualification, and Note (p)
  • UFSAR 7.6.1.1, Refueling Interlocks System - Instrumentation and Controls
  • UFSAR 7.6.2.1, Refueling Interlocks System - Instrumentation and Controls
  • UFSAR 9.1.4.2.3.11, Fuel Transfer System
  • UFSAR 15.4, Reactivity And Power Distribution Anomalies
  • UFSAR 15.7.4, Fuel Handling Accident
itOlnatlJres may be obtained via electronic processes (e.g., PCRS, ER processes), manual methods (e.g., ink signature), e-or telecommunication. If using an e-mail or telecommunication. attach it to this fonn.

EN-LI-101-ATT-9.1, Rev. 6

The n,.n.nn4::"ort to TRM SR 6.9.7.1 affects all three surveillance reaUlretmetnts a through c, as described below:

Room Auxiliary Building, elevation 182', the room through which the transfer is sealed. The is once every 7 and this extends that frequency to once every 31 Per the TRM 6.9.7 Bases:

The DurDo~~e of the horizontal fuel transfer SIJE~ClrjrCa"lon is to control n~,".<:'()nn4~1 access to those Do~:enCla/i'V high radiation areas Imrne<1latelv adJiac~~nt.

fn accordance with the HFTS Instruction SOI..Q4-1-01-F1 4.1.2 (instructions for preparing for HFTS operation), prior to operation of the HFTS, the fuel transfer tube room on Elevation 182' is checked to verify the room is locked, and These actions provide redundant protection against the room during In accordance with Surveillance Procedure 06-0P-1C71-V-0002, Attachment IIf, Refueling Interlock Horizontal Fuel Transfer Step 5.4.1, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to of HFTS at feast once per 7 Room Auxiliary Building Elevation 182' is verified sealed. Extension of this surveillance from 7 days to 31 days avoids unnecessarily interrupting fuel movement to perform the required testing. Reviews of operating and surveillance history and refueling logs show no door/seal performance issues or inappropriate room access. Therefore, extension of the surveillance frequency is justified. The Auxiliary Building Room 1A525 sealing check is performed to protect against radiation overexposure accidents only to ensure compliance with 10CFR20. fn addition to a verification of sealing of this room, the access area to the room is also required to be posted and locked as a Exclusion/Locked High Radiation Area in accordance with Radiation Protection Procedure, 08-S..Q2-75, Coverage and Control of Refueling Operation, Revision 11 prior to fuel movement.

TRM SR 6.. 9.7.1.b requires operators to "verify all interlocks with the refueling and fuel handling platforms are OPERABLE." This requirement is by the note: "Not required to be met for HFTS operations when equivalent administrative controls are in effect." The required frequency is once every 7 days, and the proposed change extends that frequency to once every 31 days.

Per UFSAR 9.1.4.2.3.11:

The transfer control system functions on a semiautomatic basis with provision made for manual override. Automatic sequencing is accomplished by use of an electronic control system located in the auxiliary building. The control system monitors the step-by-step procedure of installation or removing cargo from the carrier and assures proper sequencing of the transfer operation. Control panels are prOVided in both the containment and auxiliary buildings. The transfer mechanism is equipped with sensors and instrumentation appropriate for confirming the successful completion of each step and signaling the control system which automatically initiates the next step. The completion of a transfer operation is signaled at the control panel. When the HFTS system is in operation, interlocks or equivalent administrative controls are provided to prevent incorrect operation. Interlocks and/or administrative controls assure the con-ect operational sequence of the transfer system components and fuel handling equipment.

Per UFSAR 7.6.1.1.3.3:

The rod block interlocks, which prevent control rod withdrawal whenever the tue/loaded refueling platform is over the core, and the refueling platform interlocks, which prevent operation of the fuel loaded refueling equipment over the core whenever any control rod is withdrawn, provide two independent levels of interlock action. It is pertinent to note that the strict procedural control exercised during refueling operations may be considered another level of backup.

EN-LI-101-ATT-9.1, Rev. 6

Per UFSAR 7.6.1.1 rAn'Jlrt=tn to meet the IEEE 279-1971 criteria for

'rt'L"!,rl"I"u failure will not cause an accident. are ret,uellna interlocks have been caneTUI!!V a~eS/(Jneta rAnr, Jnlr1:::J/'~V of sensors and circuitry to level and assurance that the bases will be met.

interlock failure will not cause an accident or result in Do~rentJal physical damage to fuel or radiation exposure to fuel nar.,allfJO 1"\j",,~,.'ofi,',.,e<

Per UFSAR 7.6.11.3.7:

  • "".,lrtli"'t'L"!l functional of all interlocks before any refueling outage will positively indicate that the interlocks rln.,;;arglrg the situations for which they were ae~)lar.,ea.

Per UFSAR 7.6.2.1.1:

The Int~~rJC)CK,S. in combination with core and refueling procedures, limit the probability of an inadvertent The nuclear characteristics of the core assure that the reactor is suber/tical even when the worth control rod is withdrawn. Refueling procedures are written to avoid situations in inadvertent is The combination of refueling interlocks for control rods and the redundant methods of preventing inadvertent criticality even The interlocks on hoists provide yet another method of avoiding inadvertent The function of the HFTS interlocks with the bridge is to protect the suspended fuel bundle, control blade, etc.,

from with or in contact with the fuel transfer carriage. This function ensures that the bridge is not moved into the HFTS while the HFTS carrier is not ~n the vertical position. AU subcomponents of the HFTS are non-seismic category I, with the exception of the fuel transfer tube and hatch which are seismic category I. The function of the interlocks is not a design bases function under 50.59 since the function is (1) not required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical SPE!Cltilcal:lon,s, or (2) credited in safety to meet NRC requirements. The fuel handling accident UI.-::J"'U";).-::J~;U in UFSAR 15.7.4 is a bounding analysis and is based on a random failure of the fuel handling grapple. The refueling interlocks as discussed in section 7.6.1.1 and 7.6.2.1 primarily addresses inadvertent criticality avoidance. The interlock provides collision protection for a single bundle with the HFTS upender only.

This event would be bounded by UFSAR chapter 15.7.4. Therefore, the extension of surveillance frequencies does not adversely affect the function or method of performing the functions of the HFTS.

TRM SR 6.. 9.7.. 1.c requires operators to "verify All HFTS primary carriage position indicators are OPERABLE."

The required frequency is once every 7 days, and this change extends that frequency to once every 31 days.

For the HFTS primary carriage position indicator, procedural controls require that for every movement of the carriage, position indicator light checks are required. Additionally cable loading is required to be verified during the carriage movement. Should a malfunction occur of the carriage, it would easily be identified and self-revealing to the HFTS console operator. Should a need to shut the transfer tube hatches be required, the hatch area can be easily observed to ensure the HFTS carriage is stored in the auxiliary building prior to closure of the transfer tube hatches. No new failure mechanisms are being introduced as result extending the frequency for this surveillance.

EN-LI-101-ATT-9.1, Rev. 6

Is the validity of this Evaluation dependent on any other change? Yes No If list the required changes/submittals. The changes covered by this 50.59 Evaluation cannot be implemented without approval of the other identified changes (e.g., license amendment request)..

Establish an appropriate notification mechanism to ensure this action is completed .

Based on the results of this 50.. 59 Evaluation, does the proposed change Yes No require prior NRC approval?

Reviewer:

OSRC:

EN-LI-101-ATT-9.1, Rev. 6

II.

Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If "Yes, Questions 1 - 7 are not applicable; answer only Question 8. If UNo," answer Yes all questions below.. No Does the proposed Change:

1. Result in more than a minimal increase in the 1I~.I,.n:;ill"""Y of occurrence of an accident Yes nrA~"IOIJSlV evaluated in the UFSAR? No t

BASIS: A review of the ¥..." ..... ""... '" '-II Il~LlL'[J1 15 was conducted to determine what aC(~dE!nt~s. ext,encllna the surveillance frequency for testing the horizontal fuel transfer UI~U~.~~ a fuel bundle drop accident onto stored irradiated InValVlno HFTS are not discussed in this section nor any other UFSAR Since accidents involVing HFTS 004!ra'ttorlS are not discussed in Chapter 15, this change will not result in any increase in the of occurrence of an accident previously evaluated in the UFSAR.

2. Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a Yes structure. or component important to previously evaluated in the UFSAR? No BASIS: HFTS is not considered in the prevention or mitigation of any malfunctions or accidents and its are not defined to The extension of surveillance frequencies does not introduce a new failure mechanism into the system, nor does it cause more than a minimal decrease in the reliabifity of the The method by which testing is performed is not being changed.

This change, since it only involves a Technical Requirements Manual (TRM) surveillance frequency will not affect any sse important to therefore there is no affect on the occurrence of a malfunction of any sse.

3. Result in more than a minimal increase in the consequences of an accident previously DYes evaluated in the UFSAR? I2SJ No BASIS: The HFTS system is not involved in either the initiation or mitigation of any accident identified in the UFSAR. Therefore, this change will not affect the consequences of an event preViously evaluated in the UFSAR.
4. Result in more than a minimal increase in the consequences of a malfunction of a structure, 0 Yes system, or component important to safety previously evaluated in the UFSAR? [8J No BASIS: The HFTS is not considered in the initiation or mitigation of any analyzed malfunction or accident, and its components are not defined as "important to safety." Extension of the surveillance frequencies does not affect the offsite dose consequences of a malfunction previously evaluated.

EN-LI-1 01-ATT-9. 1, Rev. 6

5. Create for an accident of a different than any previously evaluated in the Yes UFSAR? No BASIS: The interlocks and HFTS indicators affected by this change provide collision protection for fuel bundle or blade the HFTS only. Should a collision occur, a fuel bundle or blade could in HFTS canal. A collision in fuel bundle or blade drop in the HFTS canal would be bounded by worst case UFSAR fuel handling accident. The UFSAR Chapter 15.7.4 fuel accident considers the drop of a fuel onto stored spent fuel bundles. A a fuel bundle in the transfer canal would be a less severe accident than previously evaluated since the would occur onto the floor the HFTS canal and not onto other fuel bundles.

h~,.,~t'nl"'~ extension surveillance does not create new accidents or a different type of

'!!!JiI"iI"."'4,rU DlreVIOUiSIV evaluated in the

6. Create a for a malfunction of a structure, system, or component important to safety Yes with a result than any previously evaluated in the UFSAR? No BASIS: The HFTS is not defined important to safety involving the routine control of transfer tube the frequency from 7 to 31 days has no effect on the reliability of important to The interface of the HFTS with other systems is not being changed. This does not involve any modification to any structure, system, or component important to safety tne'reT~Dre it cannot create any new malfunctions not previously evaluated.

Based on the operating of the HFTS, any noted deficiencies have been self-revealing and properly caused the to lock up. Should the HFTS interlocks fail, administrative controls may be instituted to ensure no collision can occur between a fuel bundle and the HFTS system as allowed by TRM 6.9.7.

For the HFTS primary position indicator, procedural controls require that for every movement of the carriage, position indicator light checks are required. Additionally, cable loading is required to be verified during the movement. Should a malfunction occur of the carriage, it would easily be identified and to the HFTS console operator. Should a need to shut the transfer tube hatches be required, the area can be easily observed to ensure the HFTS carriage is stored in the auxiliary building prior to closure of the transfer tube hatches. No new failure mechanisms are being introduced as a result of extending the frequency for this surveillance.

The Auxiliary Building Room 1A525 sealing check is performed to protect against a personnel radiation overexposure accidents only to ensure compliance with 10CFR20. In addition to a verification of sealing of this room, the access areas to the room is also required to be posted and locked as a Exclusion/Locked High Radiation Area in accordance with Radiation Protection Procedure, 08-5-02-75, cO'/en~ae and Control of Refueling Operation, Revision 11 prior to fuel movement. Entry in this area would prohibited and controlled for all fuel movement, therefore no new failure mechanisms would be introduced since the room will remain sealed during any fuel movement activities. These controls provide an adequate barrier for radiological safety.

7. Result in a design basis limit for a fission product barrier as described in the UFSAR being DYes exceeded or altered? ~ No BASIS: Fission product barriers include fuel rod cladding, the reactor coolant system, and the primary/secondary containment bUildings. Design basis limits would be those limits such as temperature, pressure, enthalpy, and strain. Extension of the surveillance frequencies does not affect any design basis limits for any of the fission product barriers. Therefore, this change will not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.

EN-LI-101-ATT-9.1, Rev. 6

8. aec)arture from a method of evaluation described in the UFSAR used in Yes bases or in the No BASIS: The of the surveillance 11~\.luvl'I\JY does not constitute a to any calculational or methodical framework used in the ~n!:u\l~'~ as described in the UFSAR. Therefore this criteria is not aPI)UcaDI,e.

If any of the above questions is checked "Yes," obtain NRC approval prior to implementing the change by initiating a change to the Operating License in accordance with NMM Procedure EN*LI*103.

EN-LI-101-ATI-9.1, Rev. 6

GGNS Commitment Change Evaluation There were no commitment change evaluations for the period that required NRC notification.

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