ML061280116

ER-W3-2004-0331-000, Reactor Cavity Cooling System Temperature Clarification

ML061280116
Person / Time
Site:	Waterford
Issue date:	04/09/2005
From:	Angela Buford Entergy Nuclear South
To:	Office of Nuclear Reactor Regulation
References
ER-W3-2004-0331-000
Download: ML061280116 (10)

Text

1, OVERVIEW I SIGNATURES Facilitv : Waterford 3 Document Reviewed : EIR-13'3-2004-13331-000 Change/Rev.. 0 System Designator(s)/Description : Reactor Cavity Cooling System (ROCS), Containment Building (CB)

Description of Proposed Change:

CR-WF3-2004-(71335 identified licensing and design basis document discrepancies regarding the design temperature of the reactor cavity

. Calculation 6W12R134Q, Rev. l originally contained an ambient temperature of 120°F and was updated in 1977 to reflect a new ambient steady state design air temperature of 145°F for the upper reactor cavity and 135°F for the lower reactor cavity. However, FSAR Section 9.4.5.6, land FSAR Section 3.8.3.3.1 were not identified as affected documents nor updated to reflect the design change. W3-DBD-010, section 3.2.2. l, when issued later, also reflected the lower reactor cavity primary shield wall temperature of 120°F.

The proposed hngineering Evaluation documented in ER-W,33-2004-03 1-000 will revise FSAR Section 9.4.5.6.1 to state : "The Reactor Cavity Cooling System is designed to ventilate the mtmslar space between the reactor vessel and the concrete primary shield wail to maintain the ambient steady state air temperature from exceeding the maximum design basis air temperature of the reactor cavity during normal operations". Table 9.4-1 will be tevised to include these design temperatures.

FSAR Section 3.8.3,1 1 a) Normal Leads will also be revised by LR-W_',-2004-0331-001) to include clarification that the thermal load for the reactor cavity was analyzed with an ambienr steady state air temperature of 145°F for the upper cavity and 1 :35°F for the lower cavity.

Section 14-2. 1-2.3.32

.4 Acceptance Criteria will be revised to delete the.`B." acceptance criteria as it is now captured in the "A acceptance criteria %hat refers to l; able 9.4-1.

Cheek the applicable review(s): (Only the sections indicated must be ioduded in the Review,)

LDITORIAL CHANGE of a Licensing Basis Document i Section I 30.;9 EVALUATION EXEMPTION I Sections 1,.11, and III required L] 50.59 EVALUATI011 Preparer.-

Albert l3utord' Manse (print) ! ;

Reviewer :

/', Z Natnes(print) Siguatu ;

Cornp;,rl ;

~w

`t,ry tt~

.3 OSRC :

LI-'1117.01, Rev. 7 Effective Date: 2131(15 t C,tvaiiy ` Department Date Sections I and 11 required Chairman's Name (print) % Signarure ! Dake (I';L-tluireu only fc?: hrogrammane Exclusion Screen.ngs and 50.59 cva.ua6ons.]

Sections 1, 11, and IV required

Ii. SCREENINGS

1.

Does the proposed activity impact the facility or a procedure as described in any of the following Licensing Basis Documents?

if "YES", obtain NRC approval prior to implementing the change by initiating an LBD change in accordance with NMINi ENS-.L.I-133. (See Section 5.2[13( for exceptions.)

LUDs controlled under 50.59

! YES ( NO i

'PS Bases Technical Requirements Manual Care Operating h:mits Report NRC Safety Evaluation Report an supplements for the initial FSAR' NRC Safety Evaluations for arnerndmenta to the Opezating L icrnse' If "YES", perform an Exemption Review per Section III OR perform a 5/1.54 Evaluation per Section IV OR obtain NRC approval prior to implementing the change. ff obtaming NRC approval, document the LBD change in Section II.A.5 : no further 50.59 review is required. However, the change cannot he implemented until approved by the NRC. AND initiate an LBD change in accordance with mV1M ENS-1.1-113.

LBDi controlled under other rv

1 YES regulations Quality Assurance Program Manua12 50.59 REVIEW FORM Fire Protection Program" (includes the Fire Hazards Analysis)

DR-N-05 -1480 FSAR 9.4.5.6.1 DRNr-05-1 Sao FSAR 1&3,11 DRN-i)5-1543 FSAR 14.2.12.3.32.4 CHANGE # and/or SECTIONS IMPACTED CHANGE # (it applicable) and/or SECTIONS IMPACTED 1 Offsite Dose Calculations Manual'`'

I

-YFS", evaluate a¬uv changes in accordance with the appropriate regulation -

AND initiate an LBD change in ordance with NMli ENS-LI-113. No further 50.59 review is required.

If "YES.' sea Sec1ior 5, 2[5] No LBtD change is required.

z

¬3 `YES." notify the respons:8ie department and ensure a 50.54 Evatuation is performed. AftacP the 50.54 Review Changes.c the Einergenry P'an, °=re frotocfion Program, and Ofsite Dose Caicuwicn Mranuad must be approved by the OSRC :n accordance with,ti° ttN4 OM-1, '9.

if 'YES." evaluale the change in BWerdama +tit/ the rdquaremen',s of the facility's Operaiine~ License Conaition or under 5J.55, as appropriate L1-101-01, Rev. 7 Effective Date : 213!(15 CIIANGE # (if applicable) and/or SECTIONS IMPACTED!

Page 3 of 10

3.

Basis

2.

Does the proposed activity involve a test or experiment not described its the FSAR?

if "yes, perform a 50.59 Evaluation per Section IV OR obtain NRC approval prior to implementing the change An initiate an LBD change in accordance with NNINI LI-I13, If obtaining NRC approval. document the change in Section II.A.a ; no further 50.59 review is required. However. the change cannot be implemented until approved by the NRC.

Cj° Yes o

Explain why the proposed activity does or does not impact the Operating LicenseiTechnical Specificarions antttior the FSAR and why the proposed activity does or does not involve a new test or expcfn+=r not previcusly descnbed in the FSAR. Discuss ofet Ll3Ds if impacts-d. Adequate basis trust be provided within the Screening such that a third-party reviewer can reach the same conclusions, Simply stating that the change does not affect TS a: dye TSAR is not an acceptable basis.

Operating Lieense/7"echnical Specifications The change implemented by ER-W3-2004-4331-000 will not impact any Operating License,Technical Specifications (TS). TS 3.6.1.5 requires average Containment ambient temperature to be maintained below 120°F. This change will not affect the ability to maintain average containment ambient temperature less than !24°F, but clarifies that the reactor cavity contains elevated localized temperatures that can exceed 120°F. There is no other similar information in the Operating License Chechnical Specifications F4 AR The change, implemented by ER-W3-2004-0331-000 will impact FSA t Section 9.4.5.6.1 and 3.8.;.3.1

. The following sections were evaluated in this 50.59 Evaluation, U-t01-t)1, Rev. 7 Effective Date : 213105

1. FSAR Section 9.4.5.6.1 Currently states :

9.4-5_6 RP;Kbr tweet-jr Cou;i 9A-5,61 Desc=gn Bases The Reactor Cavity Cowmg System is damned to venfilaw t3 se anru.tlw ice between the reactor vessel and tfae cwrrete pisntory situeM watt to flora the eoncret~ &urface twri rafta-e to a maxlmztn of 129°F.

The oaten £s not safety refatai, but the Fens aruo por¬1ons of the ductwork are destgn

~o seMmie i regu=rerrertts.

Ss;ction 9,4.5,6.1 will be modified tsa state the following and is addressed in Section 1V of this Fvaluation_

The Reactor Cavity Cooling System is designed to ventilate the annular space between the reactor vessel and the concrete primary shield wail to maintain the ambient steady state air t.ert erttture from exceeding the maximtnn design basis air temperature of the reactor cavity during normal operations.

Table 9.4-i will be revised to include the design temperatures of 145"1=' for the upper cavity arid 135`'1 for the lower cavity.

2.

Section 14.2,12.3.32.4 Acceptance Criteria will be revised to delete the "F3." acceptance criteria as it is

?tovv captured in the "A" acceptance criteria. that refers to Table 9.4-1 as noted above.

FSAR Section 3.8.3, CONCRETE AND STEEL INTERNAL STRUCTURES OF STEEL CONTAINMENT, 3_8.3.3.1 Loads states "'Fill the trajor loads to be encountereJ or to be pnsiulaced are listed bcloEV.

1"serrnal Load, T "-' These loads are caused by the expansion of time coatainrnent.intemal structure d;se to increased internal ambient temperaru. e during normal operation. Th-e tetnpetature of all components of the internal structure is assumed to un ;tbrntly stabilize at the same tempcratare as the internal ambient.

This is 120 °F ; the as cotwructed temperatccre is assumed to be 70 `°F. The thermal load due to neutron radiation within ilte primary shield wall is also considered."

Page 4 of 10

'lest or Experiments

4.

References

la-,s Section will the reviscd to include the follovvinr, note :

"The Reactor Cavity I'herrnal Load was analyzed using; an ambient steady state design air temperature of 145'F for the upper cavity and I 35"F for the lower cavity. This is acceptable based on the results of the design basis calculation for Primary Shield WaIL'°

`['his clarification is added since the design ambient ternperature of the reactor cavity is being allowed to reach a maximum arrrnbient air ternperamrc up to 145'F for the upper cavity and 1.35'F for the lower cavity-FSAR Section 3_8_3 defines the internal structures which incljdes compartments of the reactor cavity. Therefore, this clarification is added to prevent confusion and identify the correct analyzed the¬wtal loads for the subject internal structure. This change is acceptable since the design basis calculation 6W12R134Q, Rev. 1 supports these temperatures and defermines that the thermal loads ate acceptable and do not adversely affect the concrete_

The change is a documentation change only, and does not authorize any plant alterations or activities nor result ia any, changes to normal systcrzt operation. Therefore the change does not involve any test or experiments.

Discuss the methodology for perforcting LFTD searches. State The location oY relevant licensing document information and explain the scope of the review such as electronic search criteria used key wards) or the general extent of manual searches per Section ~.5. f (5)(dj of LI-I © i. NOTE. Ensure that manual searches are performed using controlled copies of the documents. if you have any questions, contact your site Licensing department.

LBDs Documents reviewed via LRS Autonomy keyword search using; 54.59 search criteria options :

LBDslDocuments reviewed manually-FSAR Section 3.8, Rev 138 FSAR Section 14.2. Rev 12C FSAR Section 3.9 Rev 13B FSAR Section 6.2 Rev 1313 TSAR Section 9.4 Rev 1313 fechrucal Specifications Section 34.6

'fecluaical Requirements Manual 3=4.6 CR-WF3-2004-1335 (identified Condition and Operability Evaluation)

CDCC :=44822 - Combustion Engineeriiz~g Lever FSF:-15-142, Reactor Vessel Cavity Reference Design Is the validity of this Review dependent on and other change`:

1-1-101-01, Rev. 7 Effective Date ; 213105 Keywords : [ 120 NEAR 10 concrete], [ 120 NEAR10 shield], ["Reactor Cavity NEAR 10 "concrete temperature"], ["Reactor Cavity" NEARI4 "concrete surface temperatures"], [

"concrete temperature"], ["120"], [ :`heat sink" NEARI0 initial], ["heat sink" nearl0 temperature], ["ambient air temperature"],

["thermal load"] ["Steel Internal Structures"]

CN-CSE-(31-37, W3-DBD-10, Rev 2-3, Containment Cooling 11VAC W'3-DBD-27, Rev ),Nuclear Island and Containment Building 6W.I2RB4Q, Rev. 1, Primary Shield Wail 6WI2DIIQ, Rev 4, Concrete Design Inputs FSAR Qtacsrion 1322.1, Shield Wall Temperature Gradient El Yes If "Y 1":S", list the required changeslsubmittais. The changes covered by this 50.59 Review cannot be implemented without approval of the other identifier) changes (e.g., license amendment request).

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

Page 5 of 10 S.

ENVIROtiNTEPITAI, SCREE.NltiG if any of the following questions is answered -yes," an Environmental Review must he perforated in accordance with v NIM Procedure ENS-EV-115, "Environmental Evaluations," and attached to this 50.59 Review. Consider both routine and non-routine (emergency) discharges when answering these questions.

Will the proposed Change being evaluated ;

50.59 REVIEW FORM See NMM Procedure ENS-V-117. "Air Emissions Management Program," for guidance in answering this question LI-101-01, Rev. 7 Effective Date : 2/3105 Yes No Involve a land disturbance of previously disturbed land areas in excess of one acre (i.e., grading activities, construction of buildings, excavations, reforestation, creation or removal of ponds)?

1. J7 E Involve a land disturbance of undisturbed land areas (i.e., grading activities, construcfon, excavations_

reforestation, creating, or rernovi.ng ponds)?

3.

Involve dredging activities in a lake, river, pond, or stream?

q.

[I

[K Increase the amount of thermal heat being discharged to the river or lake?

Increase the concentration or quantity of chemicals being discharged to the river, lake, or air?

6.

El Discharge; any chemicals new or different from that previously discharged?

7.

-E]

ED Change the design or operation of the intake or discharge structures?

8.

E, Modify the design or operation of the cooling tower that will change water or air flow characteristics?

9.

11 Modify the design or operation of the plant that will change the path of an existing water discharge or that will result in a new water discharge`?

10.

0 Modify existing stationary fuel burning equipment (i.e., diesel fuel oil, butane, gasoline, propane. and kerosene)?'

Involve the installation of stationary fuel burning equipment or use of portable fuel burning equipment (i.e., diesel fuel oil, butane, gasoline, propane, and kerosene)?`

i2_

El Involve the installation or use of equipment that will result in a new or additional air emission discharge?

13.

01 0 Involve the installation or modification of a stationary car mobile tank?

1-1.

Q Involve the use or storage of oils car chemicals that could be directly released into the environment`.'

15.

11 Z Involve burial or placement of any solid wastes in the site area that may affect runoff, surface water, or groundwater?

CURIT"Y PLAN SORE If any of the following questions is answered "yes," a Security Plan Review must be performed by the Security Department to determine actual impact to the Plan and the need for a change to the Plan.

Could the proposed activity being evaluated:

L1-101-01, Rev. 7 Effective Date : 213105 511.59 REVIEW FORM Documentation for accepting any "yes'* statement for these reviews will be attached to this 50.59 Review or referenced below.

Yes No Add, delete, modify, or otherwise affect Security department responsibilities (e.g.. including tire brigade, lire watch, and confined space rescue operations)'?

2. E Result in a breach to any security barricr(s) (e.g-tiVAC ductwork, fences, doers, walls, ceilings, floors, penetrations, and ballistic barriers)?

3, (l

23 Cause rnatcriais or equipment to be placed of installed within the Security Isolation Lone?

0 Affect (block, move, or alter) security lighting by adding or deleting lights, structures, buildings, or terrzporary facilities?

5.

Modify or otherwise affect the intrusion detection systems (e.g., E-fields, microwave, fiber optics)?

G.

El Modify or otherwise affect the operation or held of view of the security cameras:'

7, El Z Modify or otherwise affect (block, move., or alter) installed access control equipment, intrusion detection equipment, or other security equipment?

S.

El Z tVodify or otherwise affect primary or secondary power supplies to access control equipment, intrusion detection equipment, other security equipment, or to the Central Alarm Station or the Secondary Alarm Station?

Modify or otherwise affect the facility's security-related signage or land vehicle barriers, including access roadways?

10.

© F2 Modify or otherwise affect the facility's telephone or security radio systems?

tV.

50_59 EVALUATION Does the proposed Change :

BASIS:

LI-401-01, Rev. 7 Effective pate: 213105 Does the proposed Change being evaluated represent a change to a method of evaluation ONLY? If 0 Yes "Yes," Questions 1 - ` are not applicable ; answer only QuLstion S, If "No," answer all questions below.

1.

Result in more than a minimal increase in the frequency of occurrence of an accident previously J Yes evaluated in the FSAR?

M No Neither the primary shield wall nor the Reactor Cavity Cooking system is an accident initiator for any accidents evaluated in The FSA.R Additionally, allowing the reactor cavity ambient air temperature to increase above the 12{3°F ambient air design temperature, does not change the likelihood of an accident occi¬rring. The proposed changes would result in increased thermal stresses, but these stresses were previously evaluated in calculation 6'%,V 12RB4Q, Rev_ I and were determined to be acceptable as long as the ambient steady state air temperature is maintained under the design limit of 145"F for the upper cavity and 135° for the lover reactor cavity, The Prirua y Shield Wall and the reactor support ring girder are safety related structures that could be potentially impacted by a temperature increase. However, these structures are designed for temperatures higher than the proposed changes_ Calculation 6W 12RB4Q, Rev. 1 evaluates the design temperatures for these structures which conclude that the proposed change would have no effect on these safety telated structures No Furtherrrro¬e, the proposed changes have no impact on operator actions, operation complexity, or other human factors that could xesult in an accident initiation.

"therefore, this change will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the FSAR.

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a F~ Yes structure, system, or component important to safety previously evaluated in the FSAR?

The FEAR states that the containment subcompartments are Subject to pressure transients and jet impingement forces caused by the mass and energy releases from postulated high energy pipe ruptures within their boundaries. Subcompartments wifo which high energy raptures we postulated include the reactor cavity, the pressurizer, and the steam generator. 'I`he reactor cavity is a heavily reinforced concrete structure drat performs the dual :unction of providing reactor vessel support and radiation shielding. The two major pressure relief paths are the annular space around the upper vessel flange and the annular space around the six piping penetrations through the primary shield wall.

The evaluated change would not increase the probability of malfunction since the revised 145'3,F for the upper cavity and 135F lbr the lower cavity was determined to be within the thermal limits of the Reactor Cavity concrete without adversely afTecting the concrete integrity. 'I he reactor cavity will be capable (if providing reactor vessel support and radiation shielding without increased probability of occurrence of a malfunction after the evaluated chance is implemented No

Therefore, this change will not Result in more than a minima! increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the TSAR.

BASS :

5(7.59 REVIEW FORM Therefore, this change will not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR LI-1f31-01, Rev. 7 Effective Date : 213105

`l'lie Reactor Cavity Cooling Systern is designed to maintain the temperature in the reactor vessel cavity shell dial the primary shield wall concrete temperature is maintained at a temperature below that which will cause dehydration of the primary shield wail concrete, die reactor sup port ring gullet to be overstressed, or its the=1 prowih exceerlc(i (Reference W 3-DBD-10 Rev. 2-~). The Reactor Cavity Cooling System is not safety related and ts not the subject of any "technical Specifications or Technical Requirements Manual.

The Primary Shield Wall and the reactor support ring girder arc safety related structures. Waterford 3 Calculation 6Wl2RB4Q, Rev. 1 shows that the primary shield wall is designed for a maximum ambient steady state air temperature of 145°F for the tipper cavity and 135'F for the lower cavity, 'This Calculation also includes an -analysis of the Reactor Vessel Support Ring Girder and structural steel at the higher temperatures.

The proposed change does not involve anv physical alteration to any supports or anchoring with regards to seismic specifications. However, these structures are designed for the temperatures in the proposed change as evaluated in Waterford 3 Calculation 6W 12RB4Q, Rev. 1.

3.

Result in more than a minimal increase in the consequences of an accident previously evaluated in Yes the FSAR?

© do "The change in the maximum concrete temperature of the primary shield wall and other changes caused by higher ambient air temperatures in the reactor cavity were analyzed in Waterford 3 calculation 6W12RB4Q Rev_ ( during original construction, demonstrating acceptable thermal stresses. The initial temperature of the passive heat sinks, that include the reactor cavity primary shield wall, assumed in the Post Accident Containment pressure and temperature Analysis, ECM9f3-¬115, is 1?0°F. This value is based on the containment temperature limit of 120°F per Tech Spec 3.6.1.5. This analysis allows for elevated localized temperatures to exceed the 120 °1' iirnit as long as the overall containment temperature is maintained below the 120 °F limit. TS 3.6.1.5 requires an average Containment ambient temperature to be below 120°F. This change will not affect the ability to maintain Containment ambient temperature below 120 F because this Tech Spec has been and is currently being met. The proposed change will not invalidate the results of methodology of the containment pressure and temperature response analysis. The proposed activity does not alter the passive heat analysis because the overall containment average temperature will not exceed the 120'F limit.

The evaluated change does not compromise the integrity of the Reactor Cavity concrete, and the consequences of containment pressure and temperature response analysis, The evaluated change was also determined to have no affect on tae related components.

4 Result in more. than a minimal increase in the consequences of a malfunction of a structure, U Yes system, or component important to safety previously evaluated in the FSAR?

\\o

50.59 REVIEW FORM Page 9 of 10

BASIS, The Reactor Cavity Cooling System is not a safety related system and is therefore not credited with mitigating any accident consequences. The change in the maximum concrete temperature of the prirnary shield wail and other changes caused by higher ambient steady state air temperatures in the reactor cavity were analyzed in Waterford 3 calculation. 6W 12RB4Q Kev. i during original construction, demonstrating acceptable thermal stresses.

The Reactor Cavity Cooling System (RCCS) is not a safety-related sisters. However, the fans and portions of the ductwork are designed and installed to satisfy seismic Category I requirements. Where the collapse of ductwork can cause damage of safety-related equipment located close to the duct, that portion of the ductwork is seismically qualified to remain. intact in the event of a safe shutdown earthquake. The evaluated change will not impact the 12CCS; however, the naluated change will allow for a higher value for steady state design air temperature resulting in a higher concrete temperature in the reactor cavity and higher thernul stresses that are imposed on the concrete.

These effects were considered when deriving the steady state design air temperature as determined in Waterford 3 calculation 6Wl2RB4Q, Rev, l which justifies that the thermal stresses would be insignificant and not challenge the structural integrity of the reactor cavity.

Review of the FEAR concludes that the containment integrity must be kept within a certain pressure and temperature to ensure proper function of the containment System, Structure, or Component (SSC)_ The maximum containment pressure will not be breached by the evaluated change_ The containment temperature bounded by the FSAR is not a localized temperature, but a containment ambient temperature of 120 "F which allows for elevated localized temperatures as long as the average maxi urn average containment temperature is maintained. TS 3.6.1.5 requires average Containment ambient temperature to be below 12WF. This change will not affect the ability to mai.otain Containment ambient temperature.

Since; the evaluated change does not compromise the integrity of the Reactor Cavity concrete, the consequences of a malfunction are not changed. The evaluated change. was determined to have an insignificant affect related to increasing the consequences of a malftutetion on all the related components ; therefore, it can he concluded that all postulated design functions from the FSAR are still valid post implementation of the evaluated change. The reactor cavity will be capable of providing reactor vessel support and radiation shield{rig without increased probability of occurrence of a malfunction after the evaluated change is implemented.

Therefore, the change will not result in any increase in the consequences of a malfunction of'a structure which are analyzed for more bounding severe design basis accident conditions (i.e. Loss ©f Cooling Accident), system, or corrtponent 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 FSAR? i~ Yes 140

BASIS, The evaluated orange dogs not create a possibility of an accident that has not been previously evaluated in the FSAR.

An increase in the allowable steady state design air temperature to 145 "E for the upper cavity and 1;35°F' for the lower cavity does not pose a threat to the integrity of the cavity or any components located within the cavity; therefore, there is no possibility of any additional, non-analyzed accident occurring. No new, failures or failure modes of any structure, systems, or components are induced by this change.

Therefore, this change does not create a possibility for an accident of a different type than previously evaluated in the FSAR.

b.

Create a possibility for a malfunction of a structure, system, or component important to safety with Yes a different result titan any previously evaluated in the FSAR?

Y~ No LIw101-(11, Rev. 7 Effective Hate -. 2005

BASIS:

The Reactor Cavity Cooling System is not a safety related system and is therefore not credited with nutigating any accident consequences. The change in the maximum arrtbient steady state ai:r temperature near the primary shield wall and other changes caused by higher arttbient temperatures in the reactor cavity were analyzed in Waterford 3 calculation 6W 12RB4Q Rev I during original construction, derrtonstrating acceptable thermal stresses.

Since the evaluated change does not cotrpromise the integrity of the Reactor Cavity concrete, the possibility of a malfunction as a result of these changes would not occur. The evaluated change was determined to have an insignificant affect related to creating the possibility of a trzalhrnetioas on all the related components ; therefore, it can be concluded that all postulated design tmtetions from the FSAR are still valid post implementation of the evaluated change. The reactor cavity will be capable of providing reactor vessel support and radiation shielding without increased probability of occurrence of a malfunction after the evaluated change is implemented.

Therefore, this change does not create a possibility for a malfunction of a structure, systems, or components 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 FSAR being etcceded [~ Yes or altered?

No BASIS :

The evaluated change does not affect the design basis for the fuel cladding, RCS boundary, or the overall containment pressure. The evaluated change wil.) only affect the design basis for the Reactor Cavity Cooling System and maximum design ambient air temperatures in the upper and lower cavity, The overall average maximum containment temperature will not change The Reactor Cavity is not part of the barriers that are subject to 10 CFRS0.2 criterions, nor does it impact a fission product barrier.

Therefore this change will not result in a design basis limit for a fission product barrier as described in the FSAR being exceeded or altered_

Result in a departure from a method of evaluation described in the FSAR. used in establishing the Yes design bases or in the safety analyses'?

o BASIS; The Evaluated change is not a raethodological change, and would only be considered a verified specification change for the Reactor Cavity concrete structural integrity justified by documented calculation results. The calculation S W 12RF34Q updated in 1977 was performed using the same manner as the original calculation, but with higher ambient air temperatures

. `fhe methodology currently described in the FSAR for eatablish:ing the &sign basis or safety analysis will not be altered. The current FSAR methodology allows for elevated localized temperatures that exceed the overall triaximum containment average temperature of 120 °F, The cu-rent change will only allow art increased localized temperature for the Reactor Cavity portion of containment. This is acceptable since the maximum average containment temperature of 120 "1"' will not be exceeded because the increase in the reactor cavity area temperature will have negligible effect on the overall containment area temperature.

Therefore, this change does not result in a departure from a method of evaluation described in the FSAR used in establishing the design basis or in the safety analysis.

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 NhvlM Procedure ENS-LI-113.

LI-101-01, Rev. 7 Effective Date : 213105