ML20196G244

Changes,Tests & Experiments Carried Out Without Prior Commission Approval. with USAR Changes Other than Those Resulting from 10CFR50.59

ML20196G244
Person / Time
Site:	Fort Calhoun
Issue date:	10/31/1998
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20196G242	List: LIC-98-0139, Forwards Rept of Changes,Tests & Experiments Performed Per 10CFR50.59,description of Revs to USAR Other than Those Resulting from 10CFR50.59 Changes,Regulatory Commitment Changes & USAR Revs Covering Period of 980301-1031 ML20196G244
References
NUDOCS 9812070292
Download: ML20196G244 (47)

Text

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Attachment A LIC-98-0139 ,

4 Changes, Tests, and Experiments Carried out  ;

Without Prior Commission Approval i

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9812070292 981204  ;

PDR ADOCK 05000285 K PW ;

LIC-98-0139 Attachment A Cl{ANGES, TESTS, AND EXPERIMENTS CARRIED OLTT WITHOUT PRIOR COMMISSION APPROVAL 7 March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised General Engmeenng

Description:

None Instruction (GEI)

Engineering Manual plug valves RW-144 and RW-145 were relocated immediately upstream ofIICV-2805A and IICV-2805B respectively. De plug valves Assistance Request were relocated upstream of the pinch valves to allow the pinch valves to be replaced and avoid the need to take the entire Raw Water supply header (EAR)-97-172 out of service.

Unreviewed Safety Question Safety Analysis:

Determination (USQD)98-010 De design and operation of the strainer backwash system remains functionally unchanged. There is no increase in the probability or consequences of an accident previously analyzed in the USAR. He modification rearranged the isolation valves without affecting system integnty or scismic qualification. Hus, there was no increase in the probability of occurrence or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. He modification introduced no rew failure mechanisms and thus did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than previously evaluated in the USAR. Based on the above, the modification did not reduce the margin of safety as defined in the Basis for any Technical Specification.

EA-FC-89-055, Rev 6

Description:

None USQD 98-004 Dis EA was revised to incorporate self-assessment iuvumnardations regarding clarification of compliance assessment methodologies, manual action updates, and incorporation of new information from support document EA-FC-97-044,"10 CFR 50 Appendix R Cable Identification." His activity was part of an overall Fire Protection Program upgrade.

Safety Analysis:

Revision 6 of EA-FC-89-055 documents FCS compliance with the requirements of 10 CFR 50, Appendix R. As such, this activity does not increase the probability or consequences of an accident previously analyzed in the USAR. He activity does not challenge physical barriers or separation enteria. He EA demonstrates the capability to shut down the plant safely without an increase in radiological consequences following a 10 CFR 50, Appendix R fire event. Hus, the activity does not increase the probability ofoccurrence or consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De EA was clarified and updated with station configuration information consistent with compliance with 10 CFR 50 Appendix R. No plant equipment or systems were changed. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. Specific fire protection related technical specifications were relocated to the USAR in accordance with Amendment No.160. His activity did not reduce the margin of safety as defined in the basis for any TS.

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LIC-98-0139 Attachment A Cl{ANGES, TESTS, AND EXPERIMENTS CARRIED OtJT WITHOlJr PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised gel-35 EAR 98-216

Description:

None DCP 0010047 DCN 0010037 Spent Fuel Pool (SFP) Cooling Circulating Pump Discharge Check Valves AC-l89 and AC-190 had their internals removed to eliminate excessise USQD 98-075 maintenance caused from operating pump discharge flow turbulence. De respective butterfly discharge valve (AC-191/192) is shut to prevent backflow through the idle pump. As a result, no flow is lost from the operating pump through the idle pump. Rese pumps are load shed on safeguards actuation.

Safety Analysis:

The only pertinent accident evaluated in the USAR is the fuel handling accident in the spent fuel pool and containment. His activity did not increase the probability of occurrence or consequences of this or any other accident previously evaluated in the USAR. He spent fuel pool cooling equipment is not adversely alTected by the removal of the check valve internals because, the idle pump's butterfly discharge valve is closed to prevent backflow through the pump. Hus, no flow is lost from the operating pump. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. In normal operation, closing the idle pump's butterfly discharge valve prevents backflow through the idle pump and a corresponding loss of flow to the SFP cooling system from the operating pump. %ese pumps are load shed on safeguards actuation to reduce the power required to service non-critical loads. Dus, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. A review of the TS determined that removing the internals of AC-189 and AC-190 does not reduce the margin of safety as defined in the basis for any Technical Specification.

TSI 98-03-01

Description:

None USQD 98-60 Statements in the Basis of TS 2.4 were revised and/or deleted to make the Basis more consistent with the current revision of the USAR sections referenced in the TS as well as with engineering analyses EA-FC-90-097 and EA-FC-97-038.

Safety Analysis:

The text changes to the Basis of TS 2.4 have no efTect on the operation of plant equipment. He current configuration of the containment spray system and fan coolers as documented and evaluated in the USAR was not afTected by the changes. The wording changes reflect current USAR descriptions and the EAs. Derefore, this activity did not increase the probabihty of occurrence or consequences of an accident previously evaluated in the USAR. De text changes to the Basis of TS 2.4 had no physical impact on plant equipment or function. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident initiating mechanism results from this change. De operation and function of plant equipment is unafTected by this change. Therefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. De Basis of TS 2.4 was revised to reflect the current EA (EA-FC-97-038) and USAR sections. None of the text changes to the Basis of TS 2.4 have any efTect on the containment response analysis and therefore do not afTect the safety margin for the contaimaent pressure design limit of 60 psig. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

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~ LIC-98-0139 Attachment A -

CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOlir PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source ' Desertption/5eennary of Safety Analysis USAR Page(s) -

Sectise(s), Table (s),

i er Figure (s) Revised 'i TSI 98-04

Description:

None USQD 98-024 %e Basis of TS 3.1 was revised to include the clarifying statements of the definition of CHANNEL CALIBRATION as contained in NUREG-1432,(CEOG Standard Technical Specifications).

Safety Analysis: -[

Improperly calibrated or failure of resistance temperature detectors (RTD) or tknrsvoiAes (TC) instrument channels is not the initiating event for acewients described in the USAR. De removal of RTDs or TCs from their wells for calibration increases the likelihood of damage to the sensor and failure of the required safety related instrument channel. His out-of-well testing does not determine if there is a temperature error in measuring the required process p .a. and only determines the sensor characteristics. Conformance with these clarifying statements will have a positive affect on the reliability and proper calibration ofinstrument channels. Assurance that parameters provided to operations / engineering are properly calibrated mitigates the radiological consequences of accidents as described in the USAR. De change helps er.sure that the safety related automatic operations and indications are within previously analyzed limits. Herefore, this activity did not increase the probability of occunence or consequences of an accident previously evaluated in the USAR. De failure of RTD or TC instrument channels is an inherent assumption of the design of safety systems described in the USAR. Conformance with these clarifying statements enhances the reliability and calibration of  ;

instrument channels, which reduces the probability of RTD or TC failure. Derefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. l Based on the above information, this activity did not reduce the margin of safety as defined in the basis for any Technical Sexcification. .

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LIC-98-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITilOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description /Sammary of Safety Analysis USAR Page(s),

Secties(s), Table (s),

er Figure (s) Revised TSI 98-05

Description:

Section 6.23.1 USQD 98-031 USAR 98-19 The Basis ofTS 23 was revised to be consistent with EA-FC-97-020, EA-FC-97-026, and USAR Section 3.4. The changes are consistent with the TS Section 23(I)a requirement of a safety injection and refueling water tank (SIRWT) boron temperature not less than 50*F. USAR Section 6.23.1 was revised to make the minimum refueling temperature consistent with other analyses and design basis documents relevant to the refueling boron temperature.

Safety Analysis:

Dese revisions changed the temperature of the water in the SIRWT at which refueling boron concentration is credited from 60*F to 68*F i consistent with calculation EA-FC-97-020 and USAR Section 3.4. USAR Section 1433.5 (Boron Dilution Incident during Refueling) credits the refueling boron concentration calculation and concludes that the minimum shutdown margin of 5% will provide adequate time, specifically 30 [

minutes or more for an operator to detect and terminate an inadvertent boron dilution event. Analysis EA-FC-97-020 uses a temperature of 68*F for the refueling boron concentrction. He boron concentration required for refueling was not changed. De change did not affect the refueling shutdown margin of 5% specified in USAR Section 6.23.1. USAR Table 3.4-1 already specifies the dissolved boron content for refueling at 68'F.

Derefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR.

USAR Section 1433.5 requires a minimum of 5% shutdown margin, calculated with a temperature of 68'F. The proposed changes do not alter the temperature used by the supporting analyses EA-FC-97-020 and EA-FC-97-026. De refueling boron concentration and temperature were not changed (i.e., calculation EA-FC-97-020 was not affected). There were no changes to any equipment or system operational characteristics.

Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety ,

previously evaluated in the USAR. The proposed changes to the Basis ofTechnical Specification 23 and USAR Section 6.23.1 will not result in any equipment failure within the Safety injection system or failure to provide adequate shutdown margin in the refueling mode. De changes do not introduce any new mode of operation or failure. Therefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. He changes to the Basis of TS 23 and USAR Section 6.23.1 do not affect any safety-related equipment or the refueling shutdown margin. He changes make the Basis of TS 23 and USAR Section 6.23.1 consistent with USAR Sections 3.4 and 143 as well as EA-FC-97-020 and EA-FC-97-026. Herefore, this activity did not reduce l the margin of safety as defined in the basis for any Technical Specification.

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LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITliOLTf PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Tabic (s),

or Figure (s) Revised TSI 98-03

Description:

None USQD 98-019 CR 199700982 The Basis ofTechnical Specification 2.4 was revised to delete three unnecessary statements. One of the statements says that three component cooling heat exchangers have sufficient capacity to remove 402 x 10 6 BTU /hr following a LOCA and references USAR Section 9.7.5. Ilowe*er, USAR Section 9.7.5 does not contain such a statement. He second statement that was deleted specified the number and type of containment air coolers in service during normal plant operation. This statement was deleted as the number and type of containment air coolers in service during normal plant operation varies with the time of year. Finally, the last statement that was deleted said that if component cooling water (CCW) was lost during the post-accident phase, containment cooling could be maintained until repairs are completed. However, not all of the CCW system is located in areas that would be accessible for repairs after a large LOCA.

Safety Analysis:

The changes to the Basis of TS 2.4 do not make initiators more likely for previously evaluated accidents. No physical changes were made to plant equipment or operational methods, ne changes to the Basis of TS 2.4 do not adversely affect the ability of the CCW heat exchangers or containment air coolers to perform their credited safety functions after an accident. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. No physical changes to plant equipment or operational methods were made. The changes do not adversely affect the ability of any safety-related equipment to respond to a previously evaluated accident.

Existing levels of redundancy were not reduced. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction ofequipment important to safety previously evaluated in the USAR. Since no physical changes to plant equipment or operational methods were made, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. The deleted statements are not driving factors for the minimum equipment operability requirements for the component cooling heat exchangers, the containment air coolers, or any other part of the CCW system. No changes to the minimum equipment operability requirements ofTS 2.4 were made. Containment pressure analyses of record are within the existing equipment operability constraints of TS 2.4 and those analyses are unaffected by this change. None of the statements deleted from the Basis of TS 2.4 define a margin of safety for TS 2.4. Derefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

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LIC-98-0139 Attachment A CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WmIOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised TSI 98-02

Description:

Section 6.2.1 USQD 98-021 USAR 98-14 The Basis of Technical Specification Section 2.3 was revised to be consistent with Specification 2.3(1)f, the USAR, EA-FC-92-072 and the LOCA analysis. He listing of methodologies of reactor criticality was also removed from the Basis ofTechnical Specification Section 2.3. De listing did not provide useful information and was not required to be in the Basis Section. USAR Section 6.2.1 was revised to correct an omission from a previous change and also to make the section consistent with other documents relevant to the one-pump requirement of the high pressure safety injection (IIPSI) system during a loss of coolant accident (LOCA).

Safety Analysts:

ne changes clarify the requirement for the number ofIIPSI pumps (I pump) in the Basis of TS Section 2.3 and USAR Section 6.2.1. He one-pump requirement was previously analped in several plant design basis documents including several sections of the USAR. The proposed changes do not cause the electrical distribution system or the llPSI system to operate outside the design limit. He changes do not alter the liPSI system operational characteristics such as required flow during a LOCA. The deletion of the listing of criticality methodologies climinates the possibility of misinterpretation of the Basis Section. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. De operational characteristics and function of the electrical distribution system and the IIPSI system are unchanged. There are no changes to the llPSI system, such as flow increase, that would require more than one IIPSI pump during a LOCA.

De degree of redundancy is unchanged, i.e.: a minimum of one llPSI pump is still available from each 480V bus. Herefore, this activity did not increase the probability of occunence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

Based on the information presented above, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. He changes affect the Basis of Technical Specification 2.3. Ilowever, the changes do not afTect any operating range or the setpoint of any safety-related equipment. The changes ensure consistency with Technical Specification 2.3(1)f and other documents but do not reduce the margin of safety as defined in the basis for any Technical Specification.

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LIC-98-0139 Attachment A

. CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL .

March 1,1998 through October 31,1998 Seeree Descripties/ Summary of Safety Analysis USAR Page(s),

Sectiee(s), Table (s),

or Figure (s) Revised TSI 98-01

Description:

None USQD 98-018  ;

The Basis ofTechnical Specification Section 2.0.l(2) was revised to be consistent with Specification 2.7 and the NRC's position on the term

" OPERABLE" as it applies to the single failure criterion for safety systems in nuclear power reactors.

Safety Analysis:

De change clarifies an explanatory example used in the Basis Section. It does not cause the electrical distribution system to operate outside the design limit. It does not change system operational characteristics. The change does not render the 4160V vital buses, the unit auxiliary

• transformers, or the house service transformers inoperable dunng an accident. He function and operational characteristics of the electrical I

distribution system was not changed. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. %e Basis change does not affect the 4160V vital buses or the house service transformers during normal plant i operation or during a design basis accident (DBA). No loads are being added to or deleted from the system. The degne of redundancy is unchanged i.e., the transformers and buses maintain their indwi-Juw. %crefore, this activity did not increase the probability of wm.us or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. He change dces not introduce any new type of equipment failure mode within the electrical distribution system. He change does not involve operation, maintenance, testing, or modification of the electrical distribution system. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. . He change does not affect any operating range or setpoint of any equipment in the electrical distribution system. Derefore, this activity did not reduce the margin of safety as defined in the basis for any  !

Technical Specification.

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LIC-98-0139 Attachment A CII ANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITIIOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or l'igure(s) Revised gel-35 EAR-98-I15

Description:

None USQD 98-038 his configuration change replaced components in the second stage steam extraction piping to the feedwater heaters with matenals which are more erosion resistant but equal in mechanical strength to the original materials. Installation was done during cold shutdown using physical and l administrative portion of the Main Steam bamers to ensure personnel safety and maintain equipment integrity. The system has no direct interactions with the s system.

Safety Analysis:

His configuration change involves secondary side components (extraction steam piping) for which there are no postulated accidents. Derefore, this activity did not increase the probabihty of occurrence or consequences of an accident previously evaluated in ihe USAR. He piping and nozzle material have equivalent mechanical strength, improved crosion/ corrosion resistance and equivalent or bett:r pressure retention integrity compared to the existing material. USAR Chapter 14 contains an analysis of an operational occurrence concerning the loss of feedwater heating and its effects on reactivity. He piping change has no efTect on the operating conditions of the piping system and changes no input into that analysis. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunct:on of equipment important to safety previously evaluated in the USAR. He configuration change involves activities, materiais and testing metnods which are identical to activities performed on several other systems that have been evaluated for accidents and/or postulated failures. Failure of the extraction steam piping in a manner similar to these other systems creates no new accidents, which have nucicar safety implications. He extraction steam piping to the high pressure heaters has no interactions with any safety related device other than that already identified, specifically, the loss of feedwater heating and the turbine building pipe break analysis (which is already bounded by failure of the main steam pipirg). Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. The extraction steam piping is not mentioned in the basis for any Technical Specification. Therefore, thus activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

gel-35 EAR-98-I16

Description:

None USQD 98-040 A new bronze ball valve (IA-4096) was added to the instrument air system downstream of FCV-959 and IA-4095 to provide a large volume clean, dry, air source for pressurizing the generator during generator testing.

Safety Analysis:

Instrument air is not required for safe shutdown of the plant and is assumed lost for analyzed accidents. Heref 3re, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Equipment important to safety does not rely on instrument air to perform its function. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Loss ofinstrument air is assumed for des gn basis accidents. As a result, there are no new accident types or malfunctions of equipment important to safety that is possible due to placin;g a new valve in the instrument air system. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difierent type than any previously evaluated in the USAR. He instrument air system is not part of the plant technical specifications, and no margins of safety discuss or take credit for instrument air. %erefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

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LIC-98-0139 Atto:hm-nt A CIIANGES, TESn, AND EXPERIMENTS CARRIED OUT %Tfl10UT PRIOR COMMISSION APPROVAL  !

March I,1998 through October 31,1998 i

Source Description / Summary of Safety Analysis USAR Page(s),

Sectiee(s), Table (s),

or Figure (s) Revised Engmeermg Change Descripties: None ,

!. Notice (ECN)-98-093 .

USQD 98-034 De "X" and "Y" velocity probes on each reactor coolant pump (RCP) were relocated from the motor housing to the seal flange.

Safety Analysis:

Relocating the velocity probes on each RCP does not make a previously evaluated accident more likely. Mounting the probes on the RCP seal -[

flange does not affect the ability of the flange to perform its function as part of the seal cartridge assembly. Testing is an electncal function check 2 only and has no impact on plant operation. The relocation of the velocity probes does not affect the abil;ty of safety related equipment to perform its credited post-design basis accident (DBA) function. Herefore, this activity did not increase the probability of occurrence or consequences of ,

an accident previously evaluated in the USAR. De relocation of the velocity probes did not make a previously evaluated equipment failure more likely to occur nor does it affect the ability of safety related equipment to respond to a previously evaluated equipment malfunction. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously ,

evaluated in the USAR. He relocation of the velocity probes did not create an initiating mechanism for a different type of accident. Testing is an l l electrical function check only and has no impact on plant operation. Relocating the velocity probes did not create a new failure mechanism for [

safety-related equipment. Derefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a i different type than any previously evaluated in the USAR. RCP vibration monitoring is not covered by the Technical Specifications. Herefore, '

this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.  !

CR 199601168

Description:

Sections 4.3 & 9.2 [

USQD 98-037 Figure 4.3-11 [

USAR 98-27 USAR Figure 4.3-11, " Pressurizer Level Contml Program" was changed to make it easier to read. Tables 4.3-14 and 9.2-2 were changed to correct i conductivity units, change values to be consistent with limits in the applicable chemistry procedure, and delete chemical parameters not having [

Iimits in the chemistry procedure. Two statements in Section 9.2 regarding volume control tank capacity wer changed to make them consistent i with an earlier statement. Cli-23 strainer design flow value was changed to make it consistent with the pump downstream ofit. Cil-3 metering i pump nominal fluid temperature was changed to make it consistent with the tank that feeds it. Finally, a correction was made to a statement  ;

regarding local control of boric acid pumps.

Safety Analysis:

No physical changes to plant equipment or operational methods were made by this US AR change. Chemistry values for pnmary and makeup water are consistent with the_ existing OPPD procedure for chemical limits, which are based on EPRI guidelines for pressurized water reactors. [

Dese chemistry value changes do not make an accident more likely. He USAR changes do not adversely affect the ability of any system to j 4 perform its credited safety function after an accident Derefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. De USAR changes do not make a safety-related equipment failure more likely or adversely affect the  ;

ability of any safety related equipment to respond to a previously evaluated equipment malfunction. Herefore, this activity did not increase the l probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. %e USAR [

changes do not create an initiating mechanism for a different type of accident or equipment malfunction. Since no physical changes to plant  ;

equipment or operational methods were made, and chemistry values are consistent with the existing OPPD procedure, this activity did not create i the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. De USAR changes have no effect on the equipm:nt operability requirements of Technical Specification 2.2, which covers the chemical and volume control system (CVCS). Derefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

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LIC-98-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OtJf WITIIOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Seuree Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised ECN-96-295

Description:

Table 4.3-5 ,

USAR 97-14 r Resistance temperature detector (RTD) temperature sensors with outputs to the emergency response facility (ERF) computer were installed in the General Electric (GE) RCP motor anti-reverse rotation devices (ARD) on Reactor Coolant Pumps RC-3 A, RC-3C and RC-3D to monitor high temperature conditions. The new RTDs were installed on top of the ARDs in the existing wells. The RTD connection heads are outside the motor top hats well above the normal level of the upper oil reservoir to preclude any new oil leakage path from the RCP motor.

Safety Analysis:

ne RTDs are passive temperature devices, which only provide additional indication of ARD condition and have no affect on ARD operation. The i RTDs are not credited in any accident previously evaluated in the USAR. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. He RTDs are not safety-related equipment and serve only to provide indication of GE RCP motor ARD condition. Therefore, this activity did not increase the probability of cccurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The RTDs only provide temperature indication and high temperature alarms to t the ERF computer. As stated above, the RTDs are not safety-related equipment. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. There are no Technical Specifications concerning the GE RCP motor ARDs or the use of RTDs to monitor their condition. Herefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

USAR 98-09

Description:

Sections 6.2,9.12 USQD-98-005 Memo EOS-DEN In accordance with the USAR Verification Project, USAR Sections 6.2 and 9.12 were revised. He clarifications concerned the instrument air (IA) 0010 system, the ability to change boron concentration in the safety injection (SI) tanks using the high pressure safety injection (IIPSI) system, and the ,

position of the main steam isolation valves (MSIV).

Safety Analysis: >

He changes are in agreement with the current desigre basis and serve to correct and/or clarify the affected USAR sections. As such, this activity i did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. This activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. His activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. His activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

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LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised PC 5%79 through

Description:

Sections 9.3.2 &

50688 14.24 USAR 98-46 Power was removed from shutdown cooling valves IICV-347 and IICV-348 by opening the breakers for the valve's motor operator at the 480V USQD 98-007 motor control center (MCC). His was done to eliminate the possibility ofIICV-347/348 opening during Modes 1,2, and 3 (when meeting the requirements of OI-SC-1 " Initiation of Shutdown Cooling" and OI-SC-2 " Termination of Shutdown Cooling") as a result of an Appendix R fire scenario. De consequence of these valves opening is a breach of the high/ low pressure boundary, resulting in a LOCA. His also eliminated the possibility of HCV-347/348 being damaged and unable to open for shutdown cooling. His was postulated on an Appendix R scenario, which initiates a close signal coinciding with the motor operator torque switch being bypassed by an electrical short.

Safety Analysis:

By removing power from IICV-347/348, this activity elimincted the possibility of an intersystem LOCA as a result of an Appendix R fire scenario.

Interlock testing, valve exercise _ and position indication is performed on IICV-347/348 during the refueling outage. Leak testing of IICV-347 is completed prior to terrnination of shutdown cooling. IICV-348 is monitored for leakage by OP-ST-RC-3001. Visual confirmation of valve closure is position indication. After the breakers are opered, verification that the valves remain closed is performed. Removal of power from IICV-347/348 eliminates the possibility of a LOCA outode of containment. Since these valves are not required during Operating Modes 1,2, and 3 (when meeting OI-SC-1 & OI-SC-2), they are assured of staying in their safe position. At least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> are available before the breakers would be required to be closed if an attemate hot leg injection or shutdown cooling is required. This gives the operators sufficient time to close the breakers, which are accessible in a post-LOCA envirrament. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Opening the breakers to llCV-347/348 eliminates the possibihty ofdamage to these valves so that they are available for shutdown cooling. His eliminates the potential for an intersystem LOCA or damage to equipment.

While removal of power after the valves are closed w 11 defect the Auto Closure Interlock, the valves will already be closed and unable to be opened with the power removed. Hus, the need for the automatic signal to perform this function is unnecessary. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

As stated above, this activity is intended to preclude the possibility of a LOCA outside containment. Dis activity is not addressed in the safety analysis since it is considered to be a beyond design basis accident and is a result of the Fire probabilistic risk analysis. Opening breakers to these valves ensures that they are maintained in the safe position during Modes I,2, and 3 (when meeting OI-SC-I & OI-SC-2). After Valves IICV-347/348 are closed and testing has confirmed they are closed, the breakers are opened at their respective MCC and verified closed. De MOV retains the final closed torque and is not overcome by seismic or DBA forces. Although this will cause the position indication lights to go out, position indication is available from the alarm in the control room, which alens the operator if either valve should come ofTits closed seat. Relief Valve SI-I 88 located between IICV-347/348 would lift at 2000 psig and provide quench tank alarms (pressure & temperature ifliCV-348 opened) durmg Modes I,2, or 3 (when meeting OI-SC-1 & OI-SC-2). Therefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. Valves IICV-3D/348 are not identified in the Basis for any Technical Specification. De testing requirement of TS Table 3.3, item 12 for the autoclosure interlock was not afTected. De interlock is functional whenever power is available to the valve. Testing is still performed on a refueling frequency. He opening of the breakers is procedurally controlled so they are available for shutdown cooling, hot leg injections and fire actions. Derefore opening the breakers in Modes 1, 2, and 3 (when meeting OI-SC-1 & OI-SC-2) maintains operability of the valves. Herefore, this activity did not reduce the margin of safet" as defined in the basis for any Technical Specification.

II

LIC-98-0139 Attachment A Cil ANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITiiOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Resised ECN-98-105

Description:

Figure 1.2-15 UFQD-98-035 USAR 98-24 A change was made to allow a different welding process to be used to attach the 3/8" studs to the embedded angles which restrain the channels retaining the rubber seal fabric between the intake structure wall and the circulating water pressure tunnel. He spacing of the studs was changed in some locations. A %" plate protechng the seal fabric which existed for some time but was not shown on the construction drawing was reinstalled. The concrete floor of the tunnel was also patched to protect the reinforcing steel.

Safety Analysis:

No previously evaluated accidents invohe the intake structure or circulating water pressure tunnel. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. No equipment important to safety is affected by this activity. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment importmt to safety previously evaluated in the USAR. Failure of the pressure retaining ability of the seals or any portion of the tunnel has no impact on any safety-related equipment in the intake structure because the seals and concrete repair do not afTect the flood integrity of the intake structure. He tunnel seals serve no safety-related equipment and failure of the seals will not cause flooding of any safety-related equipment. Derefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. No Technical Specification margin or operating limit is affected by the changes in the seal details for the intake tunnel or by the concrete repair. Herefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

USAR 98-06 Description Section 11.1 USQD 98-002 His activity revised the USAR to remove the requirement for sampling ar,d analyzing a spent regenerant tank prior to transfer. De spent regenerant tank samples are not required per the offsite dose calculation manual (ODCM). De USAR chage also clarified that rather than sampling liquid waste aller collection, the waste is sampled prior to release. Waste holdup tanks are sampled prior to processing to ensure they are processed efficiently. Monitor tanks are sampled after processing, prior to release as required by the ODCM. Samples ofindividual collection vessels are normally taken only for investigation. He ODCM does not reauire sampling prior to transfer to a waste holdup tank and thus, these samples are not normally taken.

Safety Analysis:

%e change removed requirements for sampling liquid waste that are not required by the ODCM or needed for processing or release. ODCM and 10 CFR 20.1302 sampling requirements continue to be met. He change is administrative in nature and does not affect plant systems, components or operation. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluat-d in the USAR. De activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. He activity did not reduce the margin of safety as defined in the basis for aay Technical Specification.

12

?

LIC-98-0139 Attchment A  :

CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL ,

. March 1,1998 through October 31,1998 t

i Seeree Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

er Figure (s) Revised PC 53028

Description:

None ,

USQD 98-030 _

ne Core Operating Limits Report (COLR) was updated to revise the departure from nucleate (DNB) limitir:g conditions for operation (LCO) setpoint requirements. His was done as a result of a non-conservatism discovered in calculations for required overpower margin (ROPM) from two of the transient analyses performed for the Cycle 17 Reload Analysis (CEA Drop and CEA Withdrawal). De allowable peaking factors were reduced to ensure that the current curves for DNB LCO monitoring when BASSS is inoperable are conservative. De non-conservatism also I affects DNB LCO monitoring when the better axial shape selection system (BASSS) is operable. Reducing the B-array powers by 3% ensures appropriate conservatism is applied to the BASSS generated DNB allowable power levels.

Safety Analysis:

f The allowable peaking factors were reduced to ensure conservative DNB and kW/ft LCO monitoring. De accident consequences from the reduction in peaking factor limits at the corresponding power levels were evaluated in appropriate engineering analyses. All accident consequences are bounded by appropriate limits and were evaluated in accordance with NRC approved methods. De change improves the consequences of an anticipated operational occurrence / postulated accident with respect to not violating the departure from nucleate boiling (DNB) specified acceptable fuel design limits (SAFDL). Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR.

No physical changes to safety equipment were made. Although BASSS monitoring setpoints were revised, BASSS is not a safety system that provides automatic actuation or protection. %crefore, this activity did not increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the USAR. De analyses examined necessary and bounding equipment tnalfunctions as part of this procedure change. No increase it. the consequences of a malfunction of equipment important to safety previously evaluated in the USAR was identified. Reducing the allowable peaking factors or BASSS power levels created no new credible failure types. All other parameters are bounded by the previously developed analyses for Cycle 17. Changes to the COLR allowable operating space ensure that the SAFDLs for plant transients / accidents are retained within the regulatory allowable limits. Plant trip systems and operating limit restrictions protect the margin of safety. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. De evaluation performed for the procedure change ensured that the margin of safety is naintained.

• SAFDLs are protected by plant trip systems and operating limitations. Therefore, this activity did not reduce the margin ofsafety as defined in the basis for any Technical Specification.

13

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

(

1 LIC-98-0139 Attachment A l CllANGES, TESTS, AND EXPERIMENTS CARRIED OUT %TnIOlJT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s), i or Figure (s) Revised I PC 50579

Description:

None USQD 98-020 The refueling boron concemration of Technical Data Book Section VI," Core Operating Limits Report" was changed to the new requirements for l the Cycle 18 core. Refueling boron concentration was revised from 1900 ppm for Cycle 17 to Cycle 18 values of 2100 ppm at beginning of cycle, 2000 ppm when the core average burnup is a 500 MWD /MTU, and 1900 ppm when the core average bumup is a 4000 M%B'MTU. Increasing the required boron concentration is conservative and ensures that the Technical Specification required 5% delta-rho shutdown margin requirement is maintained.

Safety Analysis:

The change ensures that the proper boron concentration is present for Cycle 18. Incorrect boron concentration in the safety injection refueling I water tank (51RWD and safety injection tanks (SIT) could affect the ability to shutdown the plant and maintain adequate shutdown margin f

following an uncontrolled heat extraction or LOCA. His change ensures the proper analyzed boron concentration is present for those functions.

Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. The refueling boron concentration (~1.2 w/o) in the RCS, SIRWT, and SITS does not affect the equipment involvei as long as the solubility limit of 23.53 w/o is not exceeded. His value is much less than the solubility limit and therefore does not affect the probability ofoccurrence of a malfunction of equipment important to safety previously evaluated in the USAR. Since the changes ensure toe proper analyzed boron concentration is present m the SIRWT ind SITS, this activity did not increase the consequences of a malfurmion of equipment important to safety previously evaluated in the USAR. This change did not affect ariy equipment or setpoints. The change cannot create a malfunction ofequipment if the solubility limits of boron in water is not exceeded. He value was changed consistent with the Cycle 18 core design requirements and is not near the solubility linde. Derefore. % activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. He change in the refueling boron concentration maintains the refueling boron concentration shutdown margin required by the Technical Specifications for criticality for the Cycle 18 cre. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

l 14 ,

I- '

LIC-98-Ol39 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITilOUT PRIOR COMMISSION APPROVAL March I,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised Condition Reports

Description:

Section 14.24 (CR) 199601428 &

199800189 His change remoses USAR statements concerning dropping a heavy load (reactor vessel head) in containment and damaging equipment needed to USAR 98-05 achieve or maintain safe shutdown and/or maintain decay heat removal. He statements were from OPPD's Phase II responses to NUREG-0611 USQD 98-13 " Control ofIIcavy Loads at Nuclear Power Plants," and were determined to be incorrect based on plant conditions during specific scenarios.

OPPD's Phase I responses to NUREG-0612 contain the licensing basis requirements. While the Phase il responses are considered enhancements to the Phase I responses, they are not licensing requirements.

Safety Analysis:

De accident is a heavy load (reactor vessel head) drop in Containment. No change was made conceming the weight of the lift, frequency of the lift, mode of operation, or followed load path. De change did not affect operating procedures, required inspections, training or load limitations and handhng restrictions associated with the polar crane. He safety evaluation confirmed that the FCS licensing basis for heavy loads is the Phase I response and that the Phase II activities are an enhancement to the Phase I response. He USAR change takes appropriate credit for the Phase I administrative controls applied to the head movement, which preclude its movement over safe shutdown piping such that in the event of a load drop, decay heat removal would not be lost. Therefore, this activity did not increase the probability of occurrence or consequences of at accident previously evaluated in the USAR. De proposed changes make no changes in hardware or procedures concerning heavy load handling in containment. He radiological release consequences of a reactor vessel head drop have already been considered in the load drop on the reactor vessel and are bounding. In addition, Phase i riivities ensure that the probability of a load drop are sufTiciently remote so as to prevent damage to equipment important to nuclear safety. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. An accident of a different type than previr 2 sly evaluated is not created because the change does not afTect OPPD's Phase I responses concerning operating procedures, required inspections, training, or load limitations and handling restrictions associated with the polar crane. Dese activities are credited in the NRC SER associated with OPPD's Phase I response to NUREG-0612. De consideration of a heavy load drop was not changed and thus the possibility of a malfunction of equipment important to safety of a different type than previously evaluated was not created. Here are no applicable Technical Specifications for heavy loads and thus, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

CR 199700407

Description:

Sections 1.2, 7.3, 7.6, USAR 97-20. 9.4,9.10, and Revisions were made to USAR Sections I,7,9, and Appendix M to clarify or correct information. He changes were made as a result of a Appendix M Licensing Basis Self-Assessment of the AFW system and are supported by existing design basis calculations, analysis, and'or drawings.

Safety Analysis:

ne USAR text clarifications and administrative changes are all supported by the existing design basis. Herefore, this activity did not increase the probability of occurrence consequences of an accident previously evaluated in the USAR. Since the changes are supported by the existing design basis, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in th: USAR. Since the changes are supported by the existing design basis, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. Here are no applicable safety margins or design va:ues defined in the Technical Specifications. Derefore, this activity did not reduce the margin of safety as defined in the basis for any Twhnical Specification.

15

F LIC-98 0139 Attachment A CllANGES, TESTS, AND EXPERIMENTS CARRIED OlIT WIT 110lJr PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised TM 98-017

Description:

None USQD 98-046 The liquid (variable) legs of LT-106 and LT-10iX were cross-tied because ,he normal variable (wet leg) tap on the bottom of the pressurizer for LT-106 (pressurizer cold level indication) was plugged and nonfunctional His caused a slow or incorrect level indication. The crosstie provides a secondary wet leg pressure, which provides an accurate signal to the transmitter. The cross tie does not afTect the operability or availability of LT-10lX since the line does not require any flow and the addition of LT-106 does not afTect the wet leg pressure at LT-10lX.

Safety Analysis:

A malfunction of pressurizer level instrumentation, which responds in the event of a design basis accident is not an accident evaluated in the USAR. Cross connecting the liquid side of the non-safety related LT-l% loop to the safety related LT-10lX loop liquid side will have no effect on the response of the level control devices in the LT-10lX loop. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. The changes to the level sensing tap does not affect instrument uncertainty, performance or response time of the alTected instruments. The consequences of a malfunction (incorrect signal from the 101X loop) were not changed and this TM did not afTect the redundant loop (101Y). Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. Failure of the 10lX loop is accounted for by the redundancy in the loop with the Y channel. Tubing leakage is accounted for in the existing LOCA analyses. The connection source for the transmitter LT-106 is not changed from the original. Only the path that the sensing signal uses was changed. He quality and pressure boundary integrity is equivaient to the existing materials. Therefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. Calibration of the associated instrumentation was not affected and no devices in the loop were changed. Therefore, the response and performance of the loop was not afTected and no margin of safety as defined in the basis for any Technical Specification was reduced.

TM 98-012 Description None USQD 98-032 A blank flange to assist in trouble shooting leakage at containment mechanical penetration M-22 replaced flow element FE-2981. He flang#s material properties (design pressure and temperature) conform to the original system design. The leakage cooler retum line was not required to be operable during the trouble shooting and testing of penetration M-22.

Safety Analysis The material properties of the blank flange (design pressure and temperature) conformed to the original system design. He flange did not perform an active function or interfere with any control circuitry. Therefore, this activity did not increase the probability ofoccurrence or consequences of an accident previously evaluated in the USAR. A leak check was performed after the flange was bolted together. As stated previously, the flange did not perform an active function or interfere with any control circuitry. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR.

The possibility of an accident of a difTerent type than any previously evaluated in the USAR was not created since the material properties of the blank flange conform to the original system design. He leakage cooler return header was not required to be operable during installation of the temporary modification and therefore, the TM did not create the possibility of a malfunction of equipment important to safety of a different type than previously evaluated in the USAR. Since the material properties of the blank flange ccMorm to the original system design, no margin of safety as defined in the basis for any Technical Specification was reduced.

16

LIC-98-0139 Attachment A CllANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised TM 98-015

Description:

None USQD 98-0.t4 While in Mode 5 conditions (refuchng shutdown with the reactor cavity floodoi), the cc:nponent coohng water (CCW) system was secund and partially drained to allow IICV-402B (containment isolation valve for containment air coils) to be removed for repair. IICV-402B was replaced with a flanged spacer to allow the CCW system to be retumed to service while repairs were accomplished.

Safety Analysis:

ne temporary modification (B1) was only installed for several hours while the plant was in Mode S. Containment Cooling Coil VA-8B for post accident containment cooling was not required during this time (loss of shutdown cooling is not a previously evaluated accident in USAR Chapter 14). Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR.

Based on the above information, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The TM had the potential to impact operation of the CCW system with the resultant loss of shutdown cooling, which is an accident not previously discussed in the USAR. However, the installation of the flange spacer meant that the CCW system could perform its design function of removing decay heat as part of shutdown cooling. Derefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated m the USAR. Although the TM isolated Containment Cooling Coil VA-8A, VA-8A is ' tot required unless the reactor is critical per Technical Specification 2.4. Herefore, this ,

activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

TM 98-ON

Description:

None USQD 98-011 A temporary modification (TM) was installed using two I" capped lines in the turbine plant cooling water (TPCW) system. Side stream flow was routed through a corrosion monitor prior to returning it to the closed loop. He purpose of the TM was to evaluate the existence of corrosion and'or microbiological activity in the system.

Safety Analysis:

ne TM does not afTect any safety-related systems. The TM used existing piping and valves, which are part of the system as well as newly installed fittings of compatible materials. The TPCW system does not perform any accident mitigation functions. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Neither the TPCW system nor the equipment cooled by it are systems that are important to safety. He activity did not change the way the TPCW system operates. Derefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. He TM used a TPCW side stream for the corrosion monitor and retumed it to the closed loop. No changes to the operating mode of the plant were made that created the possibility of an accident of a difTerent type than any previously evaluated in the USAR. His TM does not affect safety-related equipment. Therefore, this activity did not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. He TM does not afTect any equipment covered by Technical Specifications.

Herefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

17

LIC-98-0139 Attachment A Cil ANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITIIOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Descdption/ Summary of Safety Analysis USAR Page(s),

Section(s), Tabic (s),

or Figure (s) Resised TM 98-003

Description:

None USQD 98-006 Temporary Modification 98-003 changed Flow Indicator FI-l112 fmm indicating Auxiliary Feedwater Pump FW-54 suction now to indicating FW-54 discharge now.

Safety Analysis:

Dere are no previously evaluated accidents, which take credit for FW-54 or discuss FW-54 suction or discharge flow. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. He only changes made under the TM were wirmg changes at Panel Al-l 15 (non-safety related) to cause F1-1112, which resides on Control Room Panel CB-10/11 to indicate FW-54 discharge flow. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De TM enhances the ability to accurately determine the flow rate to the steam generators with FW-54 runr:ing. FW-54 minimum recirculation flow and first stage take-off flow rates are still available on local flow indicators FI-II13 and F1-1 I 14. Rus, all original indications are still available. Derefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTcrent type than any previously evaluated in the USAR. FW-54 is a non-safety-related pump. Here are no Technical Specification margins of safety pertaining to FW-54 or its associated flow instruments. Herefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

Mcification Request

Description:

Section 7.5.2.7 (MR)-FC-93-01 I USAR 97-10 Power Ratio Calculator Devices RR-006, RM-006X, and RM-006Y and associated wiring were removed and abandoned in place. The holes left by removal of the devices were patched and painted. His modification was done as the intent of the power ratio calculator to calculate an axial shape index (ASI) was superseded by the installation of the online mini-CECOR/BASSS.

Safety Analysis:

The power ratio calculator did not perform a control function but rather provided ASI indication. As stated above, the intent of the power ratio calculator to calculate ASI was superseded by the installation of the online mini-CECOR/BASSS. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since the power ratio calculator did not perform a control function and ASI indication was superseded by mini-CECOR/BASSS, this activity did not increase the p obability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the US AR. Based on the above discussion, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any pirviously evaluated in the USAR. Since the power ratio calculator provided indication only (no control function), this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

18

u LIC-98-0139 Attachment A CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL -

. March 1,1998 through October 31,1998 Seuree Desertption/ Summary of Safety Analysis USAR Page(s), y Section(s), Table (s),

er Figure (s) Revised DCP 10039 Description- None DCN 10029 USQD 98-068 His temporary modification (TM) welded a pipe sleeve with a closure cap and vent valve to the body of Main Feedwater Check Valve FW 161 to contain feedwater leakage. He TM was not intended to repair the main pressure boundary but rather to prevent possible damage to safety related components in the vicinity. No change to the design features of the associated systems or the performance characteristics of the valve was made as a result of the TM.

Safety Analysis:

De closure function of FW-161 is to minimize blowdown of Steam Generator RC-2B if a rupture occurs between the main feedwater isolation valve (MFIV)in Room 81 and FW-161. This is an accident mitigation function, which this TM does not affect. De pipe sleeve is attached to the outside surface of the valve body, which is approximately 1-1/2" thick at that location and already significantly warmed, by the fluid. The welding to the outside of the valve body had negligible effect on the valve internals or the disc closure. %crefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. De welding of the pipe sleeve to the outside of the valve body had negligible effect on the pressure boundary (valve body) and the alignment of the stop and hinge pins or disc seating. He valve closure function was not changed. Herefore, this activity did not increase the probability of ecc-.mce or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. %c analysis of the TM shows that the integrity of the valve is not afTected by the addition of the leak containment attachment (pipe sleeve). However, rupture of the piping system is an analyzed cendition and this repair is bounded by the existing analysis. As stated previously, the safety function of the valve to close on demand is not afres ,1 by this TM. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a ditTerent type than any previously ,

evaluated in the USAR. His TM did not change the operating characteristics or function of the system. %crefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification. ,

I

+

}

4 I

i 19 l

-ua.-- w - ,e---in--,- ,-,--wme e- - - - - - w - up*w w-er =- ~== w-+---*---*---w "-u- = - ~ * - - - * . m~-===--'-'-w w e ^* w++ = - + - - - = - = = - .-- =w- *- - - - -a---- _ - - -

LIC-98-0139 Attachment A CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s), .

Section(s), Table (s),

or Figure (s) Revised MR-FC-97-028 - Desertpties: Figure 5.9-13, Sht.16 USQD 98-022 -

USQD 98-033 Dis modification installed high point vent valves immediately upstream of low pressure safety injection (LPSI) loop injection valves (HCV-327, USAR 98-33 HCV-329, HCV-331, HCV-333) in containment. Test valves were also installed bcsween the LPSI loop injection valves and check valves SI-194, SI-197, SI-200 and SI-203). He test valves assist in testing of back-leakage through the check valves. He purpose of the modification is to allow the LPSI header to be vented to prevent the formation of nitrogen voids large enough to cause significant water hammer loads to occur.

Safety Analysis:

Venting the LPSI header does not place the system outside its design basis nor does it increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Venting ensures that no critical void exists in the line thereby preventing water hammer affects on the system. De test valves are normally closed and administratively controlled by plam procedures. Penetrat ion M-17 is classified a reactor coolant exposed system (USAR Figure 5.9-13, sht.16). USAR Section 5.9 states that valves on lines, which branch from between the main process isolation lines, such as drain valves, are normally msntained closed by administrative controls. De high point vent valves are normally maintained closed under procedures 01-51-1 and OI-SC-l. Venting the LPSI header ensures that piping or support damage from water hammer events does not occur. The test valves assist in testing the LPSI injection check valves. Adequate administrative controls are in place to prevent a reduction in availability of equipment important to safety. Derefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. He activity is not postulated to create a new accident. It ensures the LPSI system is capable of performing its safety related function by armosing any potential void in the system. Therefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. De availability of the LPSI system as required by TS 2.1.1 and 2.3 is not adversely impacted by this activity. Herefore, the margin of safety as defined in the basis for any Technical Specification was not reduced.

l.

l 20 i

I

LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMEbifS CARRIED OUT WIT 110UT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary af Safety Analysis USAR Page(s),

Section(s). Table (s),

or Figure (s) Revised CR 199700976

Description:

Section 14.14 CR 199800387 USQD 98-025 Revisions were made to USAR Section 14.14 to establish consistency between emergency operating procedure (EOP) guidance and the USAR USAR 98-39 concerning isolation of the affected steam generator (SG) during a steam generator tube rupture (SGTR) event. The evaluation examined the SGTR radiological consequences of the expected operator responses for isolating the affected SG 30 minutes (or more following a reactor trip per EOP guidance) versus the conservative USAR analysis methodology ofisolation within 30 minutes following a SGTR. He evaluation encompassed the order in which cooldown and isolation are performed as well.

Safety Analysis:

He probability of a SGTR event is not affected by this evaluation of timed and sequenced operator responses and consequences for a SGTR. He USAR provides a bounding analysis for operator response to a SGTR accident. Assuming a different timed and sequenced operator response, it was determined that offsite radiological doses remain within a small fraction (i.e.,10%) of 10 CFR 100.1 I limits. Therefore, this activity did not increase the probability ofoccurrence or consequences of an accident previously evaluated in the USAR. No new malfunctions of equipment important to safety result from assuming the operator does not isolate the affected SG within 30 minutes atter a SGTR. The analysis method is conservative or equivalent to the guidance of the EOPs. The consequences of delaying operator response to isolate the affected SG following a SGTR by 120 minutes still results in exclusion area boundary (EAB) and low population zone (LPZ) doses that are a small fraction of 10 CFR 100.1 I limits. Thettfore, this activity did not increase the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident types are created by evaluating the timing and sequencing of operator response to isolation of the affected SG and the method by which this is accomplishert. soccifically cooldown and then isolation per the EOPs or isolation and then cooldown per the USAR.

Providing an wnt thatjustifies wito acceptable results the method by which an operator responds to a SGTR accident per the EOPs creates no new malfunctions of equipment importmt to safety. The margin of safety remains unchanged for this evaluation because the EAB and LPZ doses for a SGTR remain nearly constant for the worst case assumptions for a SGTR using the guidance of the EOPs versus the USAR for afTected SG isolation and cooldown. Even with th: assumption of cooldown and then isolation up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the limiting SGTR, violation of any of the dose consequences criteria do not occue The EAD and LPZ doses continue to remain within a small fraction of 10 CFR 100.1 I limits.

Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specificauon.

21

LIC48-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT %TTilOUT PRIOR COMMISSION APPROVAL March 1,1998 throyh October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised EA-FC-97-020

Description:

Section 3.4 EA-FC-98-011 EA-FC-98-012 The USAR was revised to update the nuclear design and measured data contained in select US AR tables to reflect reactivity parameters which are USAR 98-36 specific to Cycle 18 operation. He data that was incorporated reflect analysis and testing which is based upon Cycle 18 data, assumptions, and USQD 98-052 physics models. Data was obtained during Cycle 18 low power physics testing and from Engineering Analyses EA-FC-97-020, EA-FC-98-011, and EA-FC-98-012 using computer codes and methodology approved by the NRC.

Safety Analysis:

The data incorporated into the US AR reflect analysis and testing which h based upon Cycle 18 data, assumptions, and physics models. He data were produced in the source Engineering Analyses using computer codes aporoved by the NRC. De addition of this information to the USAR does not affect any plant hardware or process parameters, which could increase J.c i robability of an accident previously, evaluated in the USAR.

He USAR changes do not affect any assumptions, data, or models used to determine the radiological consequences of previously evaluated accidents. He accident consequences evaluated in the USAR remain bounding and therefore, this activity did not increase the consequences of an accident previously evaluated. The changes do not impact any equipment important to safety. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of ecuipment important to safety previously evaluated in the USAR. De USAR changes do not create any credible equipment failures that could initiate a sequence of events or system interaction leading to a new accident. De purpose of the changes is to update the USAR to reflect desigr, predicted and measured reactivity parameters, which are specific to Cycle 18 design and operation. No new failure modes of safety-related equipment were identified. Herefore, this act' did not create the possibility of an accident or malfunction of equipment mportant to safety of a different type than any previously evaluated m the USAR. He USAR changes do not involve changes to structures, systems, or components. He Basis of TS 2.10 was reviewed and no impact on the operating limits or margin of safety was identified. Herefore, this activity did not teduce the margin of safety as defined in the basis for any Technical Specification.

22

LIC-98-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT WTfHOlJr PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised CID 980373/01

Description:

Section 9.4 USAR 98-52 Two statements in USAR Section 9.4 were revised. De statements describe valves in the flowpath to auxiliary feedwater (AFW) pump FW-10's steam turbine driver, specifically YCV-1045, located directly above FW-10, and YCV-1045 A and B located on each Main Steam line in Room 88 The statements implied that using the handwheel can manually open the three valves. He statements were revised because they are not entirely correct with respect to the design basis. All three valves fail open upon loss ofinstrument air. The actuator spring on each valve forces the valve open. The handwheel can only be used to close a valve ifit is open. He only practical method to open these valves manually is to isolate the air supply and bleed air from the actuators.

Safety Analysis:

De operation of the subject valves was not changed by this activity, which is a text clarification supported by the existing design basis. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. The activity has no effect on the availability or performance of the steam supply flow path for FW-10 and therefore, no effect on FW-10's Wormance. De safe position for these valves is to fail open. This has not changed. De necessity of performing a local manual action to open these valves is not credible because of their fail open design. Manual closure using the handwheel is available and is not affected by this activity. Therefore, this activity did not increase the probability of occurrence or the consequences of s malfunction of equipment important to safety previously evaluated in the USAR. The availability and reliability of the AFW system to provide decay heat removal was not changed by this activity. The features inherent in the AFW system to mitigate the effects of single failures are unchanged. De design of the steam supply valves to FW-10 provides ade luate redundancy and provides for credible manual operation if needed in the event of failures. Derefore, this activity did not create the possibility o'an accident or malfunction of equipment important to safety of a different type than any previcasly evaluated in the USAR. No Technical Specification safety limits or margins have a basis in the local manual operation of FW-10's steam supply valves.

CR 199700977

Description:

Sections I.2,3.7,4.3, CR 199700991 6.2, 6.4, 7.3, 9.8, CR 199700992 Sections of the USAR were changed to resolve discrepancies found during a USAR verification review. The scope of the change concerned 9.10,11.1, and USQD 98-017 sections relating to the containment air coolers, and the component cooling water (CCW) and raw water (RW) systems. The chanFes included Appendix N USAR 98-12 corrections, clarifications, enhancements, and consolidation of data to improve the quality of the USAR. No physical change to the plant or operational methods were made.

Safety Analysis:

No physical changes to the plant or operational methods were made. The USAR changes do not adversely affect the ability of the containment air coolers, or the CCW and RW systems to perform their. credited safety functions after an accident. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since no physical changes to plant equipment or operational methods were made, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De USAR changes do not create an intiating mechanism for a difTerent type of accident or malfunction of equipment important to .afety previously evaluated in the USAR. The USAR changes do not adversely affect the ability of the containment air coolers sr the CCW or RW systems to perform their credited functions after an accident. He USAR changes have no afTect on the equipment operability requirements of the Technical Specifications. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

23

- LIC48-0133 Attachment A -

CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March 1,1998 throuBh October 31,1998 Searce Descripties/Semmary of Safety Analysis _

USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised USAR 98-02

Description:

Sections 6.2 & 6.4 USQD 98-008 CR 199700987 USAR Section 6.2.3.4 was revised to clari!y the shell and tube side flows and inlet temperatures upon which the shutdown cooling heat exchanger CR 199700988 heat transfer rate value of 37,100,000 BTU'hr is based. USAR Table 6.2-3 was revised to make the values for tube and shcIl side design temperature and pressure loss consistent wiih the heat exchanger manufacturer's specification sheet. A statement in USAR Section 6.4.3.5 about containment cooling coils internal pressure rating was changed to reflect what is supported by the available documentation. A statement in Section 6.4.5 was removed to preclude possible misinterpretation about crediting the containment heat removal system after an accident.

Safety Analysis:

No physical changes were made to plant equipment or operational methods. De USAR changes do not adversely affcet the ability of the shutdown cooling heat exchangers or containment air coolers to perform their credited safety function after an accident. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Based on the above information, i.e., no physical changes to plant equipment / operational methods, this activity did not increase the probability of occ-im cc or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The USAR changes do not create an initiating mechanism for a different type of accident. He new design tmwem e values for the shutdown cooling heat exchangers still bound the maximum expected operating temperatures for shutdown cooling or post-accident modes. The pressure retention capability of the containment air cooling coils is still adequate for the intended service. Herefore, this activity did not create the possibility of an accident or malfunct6: of equipment important to safety of a different type than any previously evaluated in the USAR. De USAR changes do not affect equir ut operability requirements for the respective Technical Specifications and therefore, the activity did not reduce the margin of safety as defined .n the basis for any Technical Specification.

24

LIC-98-0139 Attachmmt A CHANGES, TESTS, AND EXPERIMENTS CARRIED OlJr Wili:OUT s'RIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Descriptien/ Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised l USAR 98-07

Description:

Sections 6.2 &

USQD 98-015 14.15.6 CR 199700057 USAR Sections 6.2.2 and 14.15.6 were revised to:

1. delete reference to separate small and large break LOCA procedures with respect to initiation of simultaneous hot leg and cold !cg injection,

2. provide clarification as to how this is performed, correct the boric acid solubility concentration limit for LOCA, and -

3. provide for minor editorial enhancements. He changes establish consistency between the LOCA analysis, USAR, and emergency operating procedures (EOPs). -

Safety Analysis:

De changes establish consistency with the EOPs by deleting USAR references to separate small and large break LOCA procedures. Clarification is also provided for establishment of simultaneous hot leg and cold leg injection and the boric acid solubility limit. He LOCA analysis input assumptions remain valid. Herefore, this activity did not increase the probability of occurrence or consequences of an accident presiously evaluated in the USAR. As stated previously, the changes establish consistency between the EOPs, the LOCA analysis, and the USAR with no change to the EOPs or LOCA long term cooling analysis. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De LOCA long term cooling analysis and existing EOPs remain valid. Only the USAR description of the process implementation was changed and improved along with correction of the LOCA analysis beric acid solubility concentration limit. Derefore, this activity did not create the possibility of an accident of a different type than any previously evaluated in the USAR. No malfunction of equipment important to safety of a different type than any previously evaluated in the USAR is created by revising the USAR to eliminate reference to separate small break and large break LOCA procedures which no longer exist. As stated previously, the LOCA long term cooling analysis and existing EOPs remain valid. He Technical Specification margin of safety is not reduced by the USAR changes because the actions to be tak. n by the operators after a small or large break LOCA continue to be within the bounds of the LOCA long term coolmg safety analysis. EOP-03 (LOCA Optimal Recovery Procedure) and EOP-20 (Functional Recovery Procedure) guidance is based upon the NRC approved CE Emergency Procedure Guidelines (CEN-152). ,

t 25

LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT %TrilOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Fignre(s) ResIsed USAR 98-37

Description:

Sect on 9.4 CID 980293/01 USQD 98-055 A statement in USAR section 9.4 was revised to delete reference to the raw water (RW) system as a make-up source for FW-19 (emergency feedwater storage tank (EFWST)) when AC power is not available. Without AC power, the RW pumps are not available. The analysis performed for Station Blackout concluded that no makeup is needed for FW-19 during a loss of AC power event. FW-19 has adequate capacity to accommodate the Station Blackout coping duration of four hours without makeup.

Safety Analysis:

The makeup to FW-19 is part of an accident mitigating system and does not affect the failure of a component that mitiates an event desenbed in USAR Section 14. The availability of AFW to a steam generator for decay heat removal is unaffected by this change. The inherent capacity of the EFWST bounds the Station Blackout ecping duration for loss of all AC power. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since the RW system is not required to provide makeup to the EFWST during a loss of AC power event, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The RW system has no design basis requirement to provide makeup to FW-19 during a loss of all AC power event. Herefore, the unavailability of the RW system does not affect the ability of the auxiliary feedwater system to deliver water to a steam generator for decay heat removal. Consequently, this activity did not create the possibihty of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. No Technical Specification safety limits or margins have any basis in the availability of RW makeup to FW-19.

CR 199801372

Description:

Section 6.3.4.1 USQD 98-057 USAR 98-35 Section 6.3.4.1 of the USAR was clarified to refer the reader to USAR Section 9.3.6, which desenbes the limited conditions under which the containment spray pumps can be considered as available shutdown cooling pumps. This eliminates the potential for these sections to contain conflicting information.

Safety Analysis:

The change made to USAR Section 6.3.4.1 refers the reader to a USAR Section that already descnbes the conditions under which the spray pumps can be considered as available shutdown cooling pumps. He USAR change makes no physical or operational change to the containment spray system. He ability of the system to respond to an accident was not affected. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. No physical or operational changes were made to the containment spray system. He ability of the system to accommodate an equipment failure was not affected. Acrefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. For the reasons stated above, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. Technical Specification 2.1.l(4)(c) already contains provisions for allowing containment spray pumps to be considered as available shutdown cooling pumps. He USAR change is consistent with that provision. Herefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

26

LIC-98-013D Attachment A CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT %TillOUT PRIC3 COMMISSION APPROVAL March I,1998 through October 31,1998 Source Description /Sammary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Rnised ECN 97115

Description:

None USQD 98-058 He following administrative changes were made to several Piping & Instrumentation Drawings (P& ids) referenced in the USAR:

P&lD M-254 sheet cover was changed to revise tag # PI-I165 to PI-I135 P&ID M-254 sheet 2 was revised to correct the line size for FW-10's lube oil water return connection, and to correct the class boundary for the piping between FW-19 and PI-1305 (Nitrogen pressure for FW-19), and to show the gauge glass isolation valves on LG-1165 (FW-19 loop seal level).

P&lD M-254 sheet 4 was revised to depict PI-l135 and to show the line size for the root valve.

Safety Analysis:

FW-10 and its support systems are accident mitigation components and have no influence on the failure probability of the nuclear steam supply system and related systems analyzed in USAR Section 14. The performance of FW-10 is unaffected by the administrative change updating the piping size of the return line from the oil cooler. Design AFW flow, and therefore decay heat removal capability, are unafTected. The relabeled non-safety related instruments and valves have no accident function. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since no physical changes to equipment or operational methods were made, this activity did not increase the probability of occurrence or the consequences of a rnalfunction of equipment important to safety previously evaluated in the USAR. He AFW system is an accident mitigation system. Hus, its failure does not create the possibility of a new type of accident.

Revision of the P&lD has no affect on the flow of cooling water to the lube oil heat exchanger or on FW-10's performance. The line sire and design flow rate of water through the heat exchanger is unchanged from the manufacturers original configuration. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR-Since no physical or operational change was made to FW-10, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification. He other instruments and devices are not involved in any Technical Specification safety margins.

27

LIC-98-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITilOUT PRIOR COMMISSION APPROVAL Man:h 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Sectie a(s), Table (s),

or Figure (s) ResIsed ECN 98-178

Description:

Ne 4 USQD 98-059 De engineering change notice (ECN) consists of the installation and testing of a relief valve on the waste disposal system and the installation and testing of a flow control valve on the post accident sampling system (PASS). The installation of the relief valve ensures that Containment Isolation Valves IICV-500A and llCV-500B will perform their safety functions.

Safety Analysis:

ne installation of the relief valve on the waste disposal system and the flow control valve on the PASS ensures that the design pressure and tu.e,4m of the containment isolation valves is not exceeded when taking a PASS sample. The ECN ensures that taking a PASS sample does not impact the ability of the containment isolation valves to perform their safety function. He direct dose contribution from the relief valve dischargeAVD-25 does not impact the existing habitability conditions. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. De ECN ensures that the design pressure and temperature of the containment isolation valves is not exceeded. The ECN ensures that taking a PASS sample does not impact the ability of the containment isolation valves to perform their safety function. Therefore, this activity did not increase the probability of-nuse or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De installation of the relief valve on the waste disposal system and the flow control valve on the PASS ensures that the design pressure and tu,r.4uie of the containment isolation valves is not exceeded due to taking a PASS sample. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. The availability of the containment isolation valves as required by TS 2.6 is not adversely impacted by this activity. Therefore, this activity does not reduce the margin of safety as defined in the basis for any Technical Specification.

28

o_. , _

LIC-98-0139 Attachment A

' CllANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOLTf PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998

. Seurte Dese 1pties/Semetery of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figerv(s) Revised USQD 98-067

Description:

None Work Request 0001457, Tube and coupler scaffolding was erected during power operations to provide safe personnel access to Valve FW-161 in the overhead ceiling area SCF S-98-209 for in the vicinity of Safety injection Tank SI-6D. The scaffold was suspended from existing attachment brackets (hangers) that are bolted to existing WDN 14424-01 embedded Unistrut channels in the underside of the 1045 level floor beams and on the bioshield wall. The scaffold was erected immediately above and immediately north of SI-6D.

Safety Analysis:

None of the equipment potentially impacted by the work activities has the capability to result in a previously evaluated accident. There is no failure mode of the equipment that could either cause an accident or actuate the reactor protective system. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Work activities were performed in such a way that the probability of equipment damage was so low as to bejudged insignificant. The work was tightly controlled in accordance with plant procedures to eliminate errors or failures that could have resulted in dropping material from working heights. Double rigging was required for any material of significant weight; ladders were tied off when in the vertical position. All work was to be terminated and the control room notified in the event that any material was dropped on plant equipment. The postulated equipment damage does not reduce the availability of equipment assumed necessary to mitigate the consequences of an accident as evaluated in the USAR. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The evaluation did not identify any mechanism by which a different type of accident than any previously evaluated in the USAR could occur. As stated previously, the postulated equipment damage does not reduce the availability of equipment assumed necessary to mitigate the consequences of an accident.

Therefore, this actisity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. The scaffolding was erected in wn6w with all TS requirements including responding to damage resulting from unlikely but postulated wnw Therefore, the margin of safety as defined in the TS was maintained.

29

- - _ - -____-- __- .- _ -- =_ - - _ _ - . - - - _ . - - - - . _ _ _ _ _ - - -

LIC-98-0139 Attachment A CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT %THIOUT PRIOR COMMISSION APPROVAL '

March 1,1998 through October 31,1998 Seuree Description /Suminary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised j PC 53040

Description:

None [

USQD 98-016 OI-SFP-2 Operating Instruction OI-SFP-2 was revised to require that the addition of demineralized water be evaluated and verified to ensure that the boron concentration of the spent fuel pool does not drop below refueling boron concentration. His ensures that the TS required value (500 ppm) for the storage of unirradiated fuel is always maintained.

Safety Analysis:

ne procedure does not allow dilution less than refueling bcron concentration. Verification of the calculation for the maximum allowed amount of demineralized water ensures that adequate shutdown margin is maintained (KEFF s 0.95). Herefore, this activity did no' mcrease the probability  !

of occurrence or consequences of an accident previously evaluated in the USAR. Maintaining the spent fuel pool inven'ory helps ensure that equipment important to safety is operable. Using demineralized water ensures that the concentration of boric acid in the pent fuel pool doesn't get too high. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of e quipment important to I safety previously evaluated in the USAR. Restoration of spent fuel pool inventory using demineralized water ensures that excessive amounts of boric acid are not collected in the spent fuel pool by replacing water that has evaporated. No equipment was afTected by this change. Herefore,  !

this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. He procedure requires verification of the calculation for the amount ofdemineralized water necessary to restore spent fuel pool inventory and its impact on the boron concentration. He value is limited to refueling boron concentration and the TS required minimum is 500 PPM when storing unirradiated fuel. Ecrefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification r

PC 51295

Description:

None USQD 98-026 ,

TDB-III-26.A Figure 1 TDB-III-26.A, Figures I and 2 diesel generator (DG) output versus temperature and DBA load versus time plots were updated. A line was added and 2 to the figures to delineate what DB A load is compared to the Engine 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating to determine the upper operating temperature limit. Two more figures were added to show the proper operating limits for the case when the coolant used is water rather than ethylene glycol.

Safety Analysis:

The determination of the upper operating temperature limit for the diesel generators has no efTect on the probability of an accident. De use of the figures defines where the diesel generators are operable for outside ambient air temperatures. Compliance with DG operability requirements ensures that the DGs are available to mitigate the consequences of an analyzed accident. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Compliance with DG operability requirements ensures that the probability of occurrence or consequences of a malfunction of equipment important to safety are not increased. He DGs are used for accident mitigation purposes. As long as the DGs are maintained operable within the requirements of TS 2.7, the emergency power supply response to a DBA will be as analyzed. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. As long as the DGs are maintained operable within the requirements ofTS 2.7(Derefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

30

LIC-98-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT WrrHOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Tabic (s),

or Figure (s) Revised PC 50580

Description:

Sections 1.3, 3.1, 3.3, USQD 98-029 3.4, 3.6, 3.7, 3.9, 7.5, USAR 98-17 ne core loading was changed to the pattem analyzed for Cycle 18. The COLR was updated to account for changes to the operating limits 7.7,14.1,14.2,14.3, EA-FC-97-033 required as a result of the loading change. He USAR was updated for Cycle 18 operation to reflect analyses performed to support the change. 14.4,14.6,14.11 De linear heat rate measurement uncertainties were updated and combined to reflect new analyses performed for their application.

Figures 3.4-1,3.4-2, Safety Analysis: 3.4-4, 3.4-5, 3.4-6, 3.4-7,14.11-1,14.1l-Shutdown margins and times required for operator actions are maintained the same as previous values. The accident consequences from the core 2,14.11-7,14.2-1 design change are evaluated in appropriate engineering analyses. All accident consequences are bounded by the appropriate limits and are through 14.2-8,14.6-evaluated in accordance with NRC approved methods for the evaluation of these transients / accidents. Therefore, this activity did not increase the I through 14.6-5 probability of occurrence or consequences of an accident previously evaluated in the USAR. No new materials are introduced into the Reactor Coolant System. The change to the core loading does not create the possibility of a difTerent malfunction than that already analyzed for USAR Section 14. Necessary and bounding equipment malfunctions were analyzed as part of the Cycle 18 Reload Analysis. All are bounded by previous analyses. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety predously evaluated in the USAR. De core loadmg change creates no new credible accidents of a different type than previously analyzed in the USAR. Applicable analyzed enrichment limits, fuel design hmits and radiological consequences are bounded by previous analyses.

Changes to the COLR allowable opnating space ensure that the specified acceptable fuel design limits (SAFDL) for plant transients / accidents are retained within the regulatory allowable limits. Applicable analyzed enrichment limits, fuel design limits, and radiological consequences are bounded by previous analyses. Core enrichment limits are maintained to the analyred limits. Plant tnp systems and operating limit restrictions protect fuel design margins. Therefore, the changes do not create a difTerent type of malfunction of equipment because system responses are within the range of predously analyzed values. He evaluations performed for this procedure change ensure that no margins of safety as defined in the Technical Specifications were reduced.

USAR 98-16

Description:

Section 8.2, USQD 98-036 Figures 8.2-1 & 8.2-2 USAR Figures 8.2-1 and 8.2-2 were updated to reflect the actual arrangement of the transmission line connections in the 161 kV switchyard.

USAR Section 8.2.1 was updated to reflect these figure changes Safety Analysis:

The existing 161 kV and 345 kV lines to and from the Fort Calhoun Station and its adjacent switchyard were not changed. All off-site power design basis requirements are met by the existing arrangement. Dere are no equipment or operational changes imolved with this activity. De existing off-site power design basis requi ements are unchanged. Therefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since there are no equipment or operational changes involved with this activity, the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR was not increased. Since no equipment or operational changes are involved, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previousiy evaluated in the USAR. Based on the above information, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

31

LIC-98-013D Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITIIOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised USAR 98-20

Description:

Section 3.4 USQD 98-043 EA-FC-98-009 The changes updated the data, figures and tables of the USAR, technical data book (TDB) and procedure to reflect reactivity parameters specific to PC 53180 Cycle 18 operation. The incorporated data reflects analysis based upon Cycle 18 data, assumptions, and physics models. The data was produced PC 53181 in EA-FC-98-009 using NRC approved computer codes and methodology.

PC 53182 PC 53183 Safety Analysis:

ne data that was incorporated is based upon Cycle 18 data, assumptions, and physics models. He data was produced in EA-FC-98-009 using NRC approved computer codes and methodology. The addition of this information into the USAR, TDB, and procedure does not affect any plant hardware or process parameters that could increase the probability of an accident previously evaluated in the USAR. A review of USAR Section 14 determined that the changes do not affect any of the assumptions, data, or models that were used to determine the radiological consequences of previously evaluated accidents. The radiological consequences remain bounding. Therefore, the activity did not increase the consequences of an accident previously evaluated in the USAR. A :eview of the USAR accident analyses did not find that the change had any direct or indirect impact on equipment important to safety. Herefc e, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. This activity does not create any credible equipment failures that could initiate a sequence of events or system interactions Icading to an accident of a difTerent type than any previously evaluated. No new failure modes of safety-related equipment desenbed in Section 14 of the USAR were identified. Derefore, this activity did not create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. A review of TS 2.10 " Reactor Core" determined that the margin of safety was not reduced by this activity.

32

LIC-98-0139 Attachment A CllANGES,1ESTS, AND EXPERIMENTS CARRIED OLTr WITIIOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised USQD 98-041

Description:

None MR-FC-95-024 A steam trap was installed downstream of the steam chest drain valve (MS-364) on auxiliary feedwater pump FW-10. An instrument root valve was added to the steam sensing line to DPT-1039 in the speed control loop allowing the acceptance range for FW-10's delta P to be changed based on using DPI-1038. He upper and lower limits for pump delta P were changed based on using DPI-1038 in performing the surveillance test (SE-ST-AFW-3006).

Safety Analysis:

ne auxiliary feedwater (AFW) system is a safeguards system that functions in response to a design basis accident where decay heat removal is needed. The changes to FW-10 enhance its ability to deliver water to a steam generator when called upon. His modification has a positive efTect on the reliability of FW-10's turbine and a positive effect on the speed control loop by reducing instrument uncertainty. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. The changes to FW-10 enhance its ability to deliver water to a steam generator when called upon because the likelihood of turbine rotating element damage is reduced with the trap installation. Reducing the uncertainty inherent in using instrumentation when performing surveillance testing enhances reliability. He consequences of a malfunction of FW-10 is inadequate or no AFW flow. Dese consequences have not changed. The modification enhances the ability of the pump and its control loop to perform its design basis function. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. He AFW system (including FW-10)is an accident mitigation system and therefore, its failure will not create a Cl apter 14 type accident. He AFW system contains redundant components and flowpaths. Herefore, a single failure will not prevent the system from performing its design basis function. Failure of an AFW pump is considered as part ofgeneral system design. Dese failures such as inadequate AFW delivery, or rupture of the steam line to the turbine have been previously evaluated. The modification changes do not negatively affect the AFW system's mode of operation. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. Amendment i 87 deleted the statement in TS 3.9(2) which said that FW-10's discharge pressure is verified to be at least 40 psig above steam generator pressure at rated steam flow. Herefore, the changes made by this modification do not reduce the margin of safety as defined in the basis for any Technical Specification.

33

LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March I,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised USAR 98-54

Description:

Section 9.7 PC 51299 USQD 98-042 A provision was added to the component cooling water (CCW) system operating instruction which requires the LCO on raw water (RW)/CCW heat exchanger inoperability to be entered if CCW is not in service to that heat exchanger and river temperature is a 83*F. CCW is therefore required to be in service to all four RW/CCW heat exchangers if river t%miuie is a 83*F, or else the heat exchanger LCO applies. For river temperatures a 83*F, this excludes RW/CCW heat exchanger CCW valve failure to open as a credible single failure.

Sarctv Analysis:

Initiators for previously analyzcd as. dents are not relevant to CCW isolation valves to the RW/CCW heat exchangers; therefore, the new precaution added to OI-CC-1 does r ot make an initiating mechanism for a previously analyzed accident more likely. He r+cw precaution ensures that CCW bulk return temper 1 a is maintained in the analyzed range after a large LOCA or main steam line break (MSLL) inside containment.

This ensures that the CCW system will perform its safety function so accident consequences are not worsened. De new precaution ensures that CCW bulk return temperature is maintained in the analyzed range after a large LOCA or MSLB inside containment. An equipment malfunction is not made more likely by this change. De new precaution ensures that CCW bulk retum tuiw.ime is maintained in the analyzed range after a large LOCA or MSLB inside containment. We consequences of an equipment malfunction are not worsened by this change. He new precaution ensures that CCW bulk retum tow.ims is maintained in the analyzed range after a large LOCA or MSLB inside containment. An initiating mechanism for a new type of accident is not created by this change. He new precaution ensures that CCW bulk return temperature is maintained in the analyzed range after a large LOCA or MSLB inside containment. A new type of equipment failure malfunction is not created by this change. Technical Specification 2.4 covers the CCW system under the umbrella of containment cooling. His change does not adversely afTect the ability of the CCW system to provide containment cooling after an accident. It protects the CCW system by climinating a possible single failure for higher river temperatures. His is similar to what is already done on the RW system. He margin of safety related to Technical Specification 2.4 concern containment peak pressure. His margin is not reduced by this change because the change does not affect the containment pressure analyses.

34

LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITilOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised PC 53237 Destription: None USQD 98-54 The formula used to derive the Reactor Coolant System low flow trip setpoint was revised. The change is pnmanly based on the method used to account for flow uncertainty in the setpoint equation. The TS and the analytical limit were not adversely impacted.

Safety Analysis:

He revised Reactor Coolant System low flow setpoint formula is based on a reactor trip at 95% of normal 4-pump Reactor Coolant System flow.

His is consistent with TS 1.3 requirements. He analytical limit specified in EA-FC-97-029 " Cycle 18 Loss of Coolant Flow Analysis"is protected since significant margin exists from the trip setpoint. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. He revised RPS setpoint formula maintains the licensing basis and ensures the Reactor Coolant System low flow trip units function as designed. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. A Loss of Coolant Accident has been analyzed in USAR Section 14.6 and updated for Cycle 1S in EA-FC-97-029. No other accident is postulated as a result of the changes. Any postulated equipment malfunction as a result of this change would be consistent with the assumptions and inputs used in the USAR Section 14.6 analysis. Calibration of the RPS low flow trip units will be performed in accordance with IC-ST-RPS-011. Since this change is merely a change

~

in the method used to calculate the trip setpoint, the possibihty of an equipment malfunction is unchanged. He margin of safety as defined in the Basis of TS 13 is not impacted. Reactor Coolant System Low Flow Trip Setpoint Calculation FC05844, Rev. I determined that the total loop uncertainty is consistent with the 2% flow error implied in the Basis section.

CR 199701077 Description :

USQD 98-012 Section 14.12.6 USAR 98-10 Revision was made to the USAR Section 14.12.6 to change the terminology " atmospheric dump valve" to " Air Assisted Main Steam Safety Valve EA-FC-90-037 (MSSV) MS-291 or MS-292." His change is consistent with the supporting analysis EA-FC-90-037.

Safety Analysis :

De Main Steam Line Break radiological consequences analysis EA-FC-90-037 was performed using and industry generic terminology for the valves MS-291 and MS-292 and their function At most plant these valves would be addressed as atmo pheric dump valves. Ilowever, this terminology is inappropriate for FCS because of the design with a manual valve atmospheric dump valve (IICV-1040) downstream of the Main Steam Isolation Valves.

The USAR change does not alter the engineering analysis or its assumption. Dere are no physical changes to the plant. De change is a text clarification supported by the existing design basis. De operation of the subject nlves was not changed by this change. Herefore, this activity did not increase the probability of occmTence or consequences of an accident previously evaluated in the USAR. The activity has no effect on the availability or performance of the MSSVs. This activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. He availability and reliability of the MSSVs were not changed by this activity.

His activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. No Technical Specification safety limits or margins have been changed.

f ,

35 l

l

LIC-98-0139 Attachment A CllANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March I,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s), i Section(s), Table (s),

or Figure (s) Revised ECN 96-462 Description : Section 5.2.3.2 [

USQD 98-065 USAR 98-040 USAR Section 5.2.3.2 was revised to include an alternate material specification for Containment tendon bushings.

Safety Acalysis: i ne bushing material specification originally designated in Section 5.2.3.2 is no longer available. He subject ECN 96-462 determined that the ,

new material specification has equivalent or greater critical design characteristics of the original material specification.

The USAR change does not alter the current design basis. Here are no physical changes to the plant. He change is a text addition of the AISI code supported by the existing design basis, ne operation of the subject valves was not changed by this change. nerefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. He activity has no efTect on the availability or performance of the containment structure, This activity did not increase the probability of occurrence or the consequences of a ,

malfunction of equipment important to safety previously evaluated in the USAR. The availability and reliability of the containment and its systems were not changed by this activity. His activity did not create the possibility of an accident or malfunction of equipment important to safety of a dilTerent type than any previously evaluated in the USAR.

CR 199701556

Description:

Section 9.8.4.1  ;

USQD 98-023 USAR 98-13 USAR Section 9.8.4.1 was revised to delete the discussion regarding manual operation of the raw water outlet valve. f i

Safety Analysis:

i~

ne discussion was deleted because the raw water outlet valves are not manually operated and there are no design bases requirement to balance raw water flow through the raw water / component cooling water heat exchangers using the raw water outlet valves.

He USAR change is for clarification only and does not alter the current design bases of the raw water system or the component cooling system. ,

Dere are no physical changes to the plant. De change is a text change supported by the existing design basis. He operation of the subject raw  ;

water outlet valves was not changed by this change. nerefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. De activity has no efTect on the availability or performance of the raw water system or the i component cooling system This activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. De availability and reliability of the raw water system and the component cooling water ,.

system were not changed by this activity. His activity did not create the possibility of an accident or malfunction of equipment important to

, safety of a difTerent type than any previously evaluated in the US AR.

t t

r i

i 36

[

i

LIC-98-0139 Atterhment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITilOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Seuree Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised ECN 97-211 Deser;ption: None USQD 98-051 Onfice plates were removed from the pressure switch PS-6559 sensing line. PS-6559 controls the start function of the electric motor-driven fire pump (FP-1 A). He 1/2" brass sensing line has two unions installed on it that have integral orifices. He orifices provide system damping to ensure proper switch action. He orifices were replaced by a pair of normally closed gate valves with holes drilled in the discs. His allows sensing pressure when the valves are closed and also allows sand to be flushed from the sensing pipe when the valves are opened enhancing reliability of the start function of FP- I A.

Safety Analysis:

No malfunction of FP- 1 A or the fire suppression water system increases the probability of occurrence of an accident previously evaluated in the USAR. FP-1 A has no direct interactions with accident mitigating safety-related components until a fire occurs. Ilowever, there are no accidents in USAR Section 14 that invoh e a fire concurrent with a design basis accident. Herefme, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since the fire suppiession water system is not assumed to be in service during the operation of safety-related equipment during an accident, this activity did not increase the probability of occurrence or consequences of a malfunction of equipment important to safety. Here were no changes to che fire suppression water system in materials of construction, modes of operation, setpoints of control devices, or testing conditions. Derefore, no new failure modes were created and the interaction with safety-related systems was not changed. A failure of the fire pump to start due to a sens3ng line malfunction has no effect on safety-related devices that operate during design basis accidents. De fire suppression water system is not required to operate during a design basis accident. He configuration change does not affect the design redundancy of the second pump (FP- 1 B) because the modified sensing line serves only FP-1 A. Here are no safety limits or margins for the fire suppression water system in the Technical specifications.

CR 199701643

Description:

Sectiona3 USAR 98-44 USQD 98-062 USAR Section 4.3.14 was updated to reflect actual leak detection methodology. His was done to correct a statement that cumulative run tirre of the containment sump pumps were used as a means to detect reactor coolant leakage when containment sump level indicators were actually the instruments used.

Safety Analysis:

De USAR change was for clarification; no new accident initiating instruments were installed and existing Reactor Coolant System leakage detection instruments were not modified. He USAR change did not afTect barriers or analysis parameters of previously evaluated accidents.

Containment sump level monitoring during normal or accident conditions was not modified. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the US AR. He USAR change did not modify reactor coolant leak detection equipment in containment and the probability of an equipment malfunction was not affected. Here were no equipment changes or methods ofleak detection and the consequences of a malfunction of equipment important to safety were not afTected. Therefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. No new accident imtiating equipment was installed or existing modified. Existing equipment important to safety was not modified and no new equipment malfunction failure modes were created. Herefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. Technical Specifications 2.1.4(4) and the Basis regarding the reactor coolant leak detection equipment, controls, setting, and operability requirements were not changed. Herefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

37

LIC-95-0139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITilOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised USAR 98-50

Description:

Section 14.23 USQD 98-063 EA-FC-94-013 De toxic gas accident analysis (EZ 94-012), was revised to account for changes proposed by the FCS Chemistry Department, and for new, lower immediate danger to life and health (IDLll) limits promulgated by NIOSil. He analysis demonstrates that the control room will remain habitable following an onsite spill of ethanolamine (ETA), hydrazine, or morpholine. Since control room habitability is maintained, nuclear safety is not affected and no margin of safety is reduced. He changes afTect the concentration of chemicals stored on site, but do not increase the likelihood of an onsite spill.

The analysis includes key data and assumptions concerning the quantity of chemicals stored onsite:

1. Ilydrazine is injected from a semi-fixed 365 gallon tote mounted in the turbine building. A second 365-gallon tote is stored on top of the primary tote. One additional tote may be stored in the warehouse.

2. ETA is injected from a 365-gallon tote stored in the turbine building. An additional 280-gallon tote may be stored in the warehouse. He analysis conservatively assumes a 365-gallon tote volume. He assumption bounds current practice while providing operating margin to accommodate future changes.

3. The FCS Chemistry Department currently plans to use one 55-gallon drum of morpholine in the turbine building. Up to four additional drums may be stored in the warehouse. The analysis conservatively assumes that one 365-gallon tote will be used in the turbine building and that one additional tote may be stored in the warehouse. De assumption bounds the current plan while allowing operating margin for future changes.

De analysis postulaTs that one of these chemicals (i.e., ETA, hydrazine, or rnorpholine)is spilled outside during handling. Key assumptions used in the engineering analysis were incorporated into the text of USAR Section 14.23.

Safety Analysis:

USAR Section 14.23 evaluates the impact of a toxic gas accident (TGA), including onsite chemical spills. He changes affect the types and concentrations of chemicals stored onsite, but the control room will remain habitable (within des;gn basis limits) following an on site spill of hydrazine, ETA, or morpholine. Herefore, this activity does not increase the probability of an occurrence of a TG A. He revised analysis demonstrates that the control room will remain habitable after an onsite spill of ETA, morpholine, or hydrazine, which ensures that reactor safety is not affected. The changes reduce secondary plant corrosion and steam generator sludge buildup, thereby reducing the chance of equipment failure due to corrosion and will limit the severity of any failure that does occur. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The changes are consistent with EPRI guidelines and do not introduce new hazards or the possibility of a new type of accident. Derefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. The Technical Specifications do not require monitoring or other actions related to chemicals stored onsite. TS Section 2.22 is based on the SER related to Amendment 183. Amendment 183 allowed Fort Calhoun Station to delete all toxic gas monitors except those for ammonia. The SER concluded that the " licensee has provided acceptable deterministic justifications for deleting the requirements for monitoring hydrazine, ethanolamine, and sulfuric acid" EA-FC-94-012 contains the deterministic justifications referenced in the SER. The EA concluded that onsite chemical spills do not jeopardire control room habitability since in all cases, the control room concentration resulting from the spill is less than the toxicity limit for the chemical involved. De concentration of ethanolamine used and stored onsite has been changed and morpholine was reintroduced. The revised analysis concludes that the control room concentration resulting from an onsite spill is less than the applicable toxicity limit and thus the margin of safety is not reduced.

38

LIC-98-0139 Attachment A C11ANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s), I Section(s), Table (s),

or Figure (s) Revised CR 199701517

Description:

Section 9.5.1.2 USAR 98 43 USQD 98-064 De USAR was revised to remove an outdated statement concerning a historical interlock on spent fuel pool handling machine (Fli-12) between the Region I and Region 2 racks. He USAR was also clarified to state that a fuel bumup determination with appropriate independent verification is performed by surveillance test prior to fuel movement from either the reactor core or from a Region I rack to a Region 2 rack. These discrepancies were identified during the USAR verification project of Spent Fuel Pool Cooling & Spent Fuel Storage Systems.

Safety Analysis:

The USAR change affirms the previously approved and NRC accepted administrative controls for transfer of spent fuel from the teactor core to Region 2 of the spent fuel pool. De probability of occurrence of a misplaced fuel assembly has not increased. De worst-case misplaced fuel assembly accident is for an unirradiated assembly and not an irradiated assembly. Criticality control is maintained even for the worst case accident by the TS 2.8(12) provision for 500 ppm boron concentration in the pool. He consequences of an accident for a misplaced fuel assembly are unchanged. De passive spent fuel racks are the only pertinent (IE) equipment and will not malfunction. The fuel racks are not affected by this clarification of administrative controls as the approved method of controlling the movement of a spent fuel assembly into the Region 2 portion of the spent fuel pool. Derefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previously evaluated in the USAR. The only credible accident remains the misplacement of a fuel assembly, which was previously evaluated. The activity does not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. He detailed administrative instructions and TS 2.8 operational requirements (specifically the 500 ppm boron concentration) provide assurance that no reduction in the margin of safety for fuel reactivity occurs. He clarification of administrative controls by the USAR change does not reduce this margin of safety. Therefore. this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

39

LIC-98-0139 Attachment A CIIANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March I,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised .

• USAR 98-03

Description:

Sections 1.2,1.6, 4.2, MR-FC-97-005 4.3,I4.10,14.15 The existing orifice plates attached to each steam generator (SG) tube sheet (hot leg plenum side) were removed. His allows eddy current testing of those tubes located tader the plate with the latest probe technology. He comer plate bolt heads were modified to allow eddy current and tube plugging equipment access to those tubes adjacent to the corner plates. Removal of the plates decreases hydraulic resistance in the steam generators and results in approximately 5% increase in primary flow rate and a slightly lower (2.6 *F) T hot.

Safety Analysis:

Orifice plate removal changes primary and secondary parameters. The resulting operating conditions were evaluated with respect to plant operation, system design, and plcnt normal and accident procedures and none is postulated to increase the probability of an accident previously evaluated in the USAR. The impact on fuel integrity, steam generator tube degradation was included in the analysis. He USAR Section 14 accident analysis was revised to reflect the change in operating conditions. The changes do not increase the consequences of an accident previously evaluated in the USAR. Orifice plate removal retums the Reactor Coolant System to its original design configuration and climinates the potential for mechanical failures and the resulting loose parts. De resultant increase in primary flow and lower T hot is within the anticipated range of components and systems for which the plant was designed and licensed. ABB/CE performed an evaluation of the operational conditions ,

with respect to the original design values and determined that adequate code allowable margins exist. Orifice plate removal returns the plant to its original design configuration. The change in operating parameters was evaluated with respect to fuel integnty, Section 14 events, and SG tube degradation. The alTected operational parameters do not change plant operation or response such that it is postulated to increase the consequences of a malfunction of equipment important to safety. Reactor Coolant System flow rate is increased by virtue ofdecreased hydraulic resistance through the steam generators. He resulting change in operating parameters (decrease in T hot and increase in primary flow) was evaluated and it was determined not to create the possibility of an accident of a different type than any previously evaluated. De resulting operating conditions were evaluated by ABB/CE with respect to design values for the Reactor Coolant System and found to be within code allowables for which the i Reactor Coolant System was designed. Herefore, the revised operating conditions do not adversely impact original design margins or the bounding conditions assumed in accident analysis. He change does not introduce an initiating mechanism for a different type ofequipment  ;

failure. The change in operating conditions is not postulated to create the possibility of a malfunction of equipment important to safety of a '

different type than any previously evaluated in the USAR. Orifice plate removal results in a 5% increase in Reactor Coolant System flow rate.

His increases the margin for minimum Reactor Coolant System flow rate specified in TS 2.7. Therefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

I I

?

40

LIC-98-0139 Attachment A-CHANGES, TESTS, AND EXPERIMENTS CARRIED OUT WfnlOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

er Figure (s) Revised USAR 98-23

Description:

Sections 9.7.2 &

MR-FC-97-007 9.7.8 Changes made to component cooling water (CCW) surge tank relief setpoints and CCW surge tank pressure alarm under a 1996 temporary modifica* ion was made permanent by this modification. CCW surge tank pressure and level indicator were relocated to the control room to provide easier access for the operatets. A larger local pressure gauge was provided for the CCW surge tank. Setpoints were increased on ten CCW thermal relief valves to preclude possible lifting during a post-DBA transient. Demineralized water (DW) and nitrogen gas (NG) makeup lines to the CCW surge tank were modified to (1) provide additional isolation capability, and (2) relocate valves to a more convenient location for operation and maintenance, and (3) facilitate periodic testing.

Safety Analysis:

His modification did not increase the probability of occurrence of an accident previously evaluated in the USAR. There are no accident initiators associated with CCW surge tank relief valves, CCW thermal relief valves, CCW surge tank control room instrumentation, or the DW and NG makeup lines to the CCW surge tank. He changes made in this modification don not adversely affect the ability of the CCW system to perform its credited accident function. The consequences of a previously evaluated accident are not worsened by these changes because following a DBA, the CCW system will operate the same as before implementation of the modification. De changes made under this modification do not increase the likelihood of a failure of safety-related equipment previously evaluated in the USAR. CCW thermal relief valves, CCW surge tank control room .

instrumentation, and the DW and NG makeup lines to the CCW surge tank are not associated with mechanisms causing an analyzed safety-related equipment malfunction. Equipment failure mechanisms were not made worse by this modification. It was confirmed that the CCW system will withstand a post-DBA transient without a pressure boundary failure. Herefore, this activity did not increase the probability of occurrence or the consequences of a malfunction of equipment important to safety previoisly evaluated in the USAR. This modification did not create a failure mechanism that could cause a different type of safety-related equipmen*: failure. The CCW surge tank relief setpoints are within the design pressure of the tank and will serve to limit system pressure to 110% ofdesign during a post-DBA transient He new thermal relief valve setpoints ensure that the respective valves will not lift and cause a loss of CCW inventory during a post-DBA transient, while still providing protection against pressure boundary failure within their respective thermal overpressure isolation boundaries. He changes to the DW and NG makeup lines to the CCW surge tank provides redundant isolation capability and does not introduce a new equipment failure mechanism. Therefore, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a different type than any previously evaluated in the USAR. Technical Specification 2.4 includes operability requirements for CCW system components. The design aspects of this modification do not reduce the margin of safety as defined in the basis for TS 2.4 because the design does not reduce the number of operating CCW components related to the credited nuclear safety function in removing heat after an accident. Minimum operability requirements for equipment were not changed. Therefore, this activity did not reduce the margin of safety es defined in the basis for any Technical Specification.

41

LIC-984139 Attachment A CilANGES, TESTS, AND EXPERIMENTS CARRIED OUT WITHOUT PRIOR COMMISSION APPROVAL March 1,1998 through October 31,1998 Source Description / Summary of Safety Analysis USAR Page(s),

Section(s), Table (s),

or Figure (s) Revised USAR 98-22

Description:

Section 8.3.4.2 MR-FC-97-001 A new circuit was added to both DC busses, which provides voltage indication between each leg of the DC bus and ground. The voltage readings are proportional to the resistance between each leg and ground and enable the operators to determine the magnitude of any ground fault that may occur.

Safety Analysis:

The new circuit is fused to provide electrical separation if an internal fault should occur. He equipment is scismically mounted to address any two over one concern. Herefore, this activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR. Since the new circuit is fused to provide electrical seoaration if an internal fault should occur, this activity did not increase the probability of occurrence or the consequences of a malfunction of equiprnent important to safety previously evaluated in the USAR. Based on the above, this activity did not create the possibility of an accident or malfunction of equipment important to safety of a difTercnt type than any previously evaluated in the USAR. De new circuit is not desenbed in the Technical Specifications and has been designed so that it will not have an adverse impact on any circuit described in the TS. Derefore, this activity did not reduce the margin of safety as defined in the basis for any Technical Specification.

USAR 96 54

Description:

Sections 5.9.5, MR-FC-91-039 14.14.1 Dis modification provides Operations personnel with the ability to override the containment isolation actuation signal (CI AS) closure 6f steam generator (SG) blowdown isolation valves llCV-2506A&B and ilCV-2507A&B for a preset time during a steam generator tube rupture (SGTR) "

event that results in a CIAS. This enables Chemistry personnel to take SG blowdown samples to identify and isolate the affected SG within the time period stated in the USAR.

Safety Analysis:

His modification involves changes to mimmire the consequences of a previously evaluated accident (SGTR) by earlier identification and isolation of the affected SG. It does not affect the plant during any other mode of operation. Since a SGTR is an analyzed event and this change enhances the ability to identify and isolate the affected SG during a SGTR, the probability of occurrence of an accident previously evaluated in the USAR is not increased. Failure of Operations personnel to follow the abnormal operating procedures ( AOP)/ cmcrgency operating procedures (EOP) could result in the release of at most 60 gallons of sampling wa cr to the river. Ilowever, radiological analysis concludes that such a release does not exceed 10 CFR 20 or 10 CFR 100 limits. Herefore, this activity does not increase the consequences of an accident previously evaluated in the USAR. The control circuits of SG blowdown isolation salves IICV-25%A&B and IICV-2507A&B were modified to override a CIAS in the event of a SGTR. His does not afTect the plant during any other mode of operation. De enhanced ability to identify the affected SG during a SGTR does not increase the probability of occurrence or the comequences of a malfunction of equipment important to safety previously evaluated in the USAR. De change is specific to a SGTR accident and the consequences of overriding the CIAS are bounded by the LOCA analysis. De existing switch and timer and a new tripping relay is used to enable the operator to override CIAS from SG blowdown sampling isolation valves during a ,

SGTR event that results in a CIAS. Although a malfunction of the timer could cause the isolation valves to be open for more than two hours, there would not be a continuous release since the chemist closes the valve after taking the grab sample. Therefore, this activity does not create the possibility of an accident or malfunction of equipment important to safety of a difTerent type than any previously evaluated in the USAR. His modification involves changes to minimize the consequences of a previously evaluated accident (SGTR) by earlier identification and isolation of the afTected SG. As such, this activity does not reduce the margin of safety as defined in the basis for any Technical Specification.

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a i Attachment B 1

! LIC-98-0139  !

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USAR Changes Other than Those Resulting from 10 CFR 50.59  :

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I LIC-98-0139 AttachmentD

- USAR CHANGES OTHER THAN 'DIOSE RESULTING FROM 10 CFR 50.59 March I,1998 through October 31,1998 Source Description USAR PAGE(S), ,

SECTION(S) TABLE (S)

OR FIGURE (S)

REVISED -

USQD The USAR was revised to incorporate mformation from a NRC SER granting an exemption from instalhng RCP lube oil Sections 4.3,4.7, & 9.1 l 98-053, collection equipment on certain locations on the RCP motors. The ;Av.yvision of the SER into the USAR documents OPPD's USAR 98- ***vl. w with 10 CFR 50, Appendix R.

28 USAR 98-21 This was a trivial USAR change to clanfy the description of the diesel generators (DG). The change noted that rather than bemg Section 8.4 identical, the DGs are similar in design. There are minor differences in the intake damper configuration and only DG-2 is configured for fire safe shutdown capability. There were also minor revisions to the description of the wall mounted day tank in the diesel room USAR 98-1I lhe USAR was revise to reflect the results of EA-FC-98-015, which documents the results of Cycle 17 end-of-cycle moderator Table 3.4-14 EA-FC toup.e .. coefTicient testing.

015 USAR 98-26 In cuid w with NRC Inspection and Enforcement Bulletm (IEB) 79-14, NRC stafT reviewed a rehef request concerning Pipe Appendix F USQD 98-50_ Supports SlH-3 and RCH-13. Because ofALARA concems, the NRC granted relief from modifying these pipe supports. USAR  ;

Appendix F was revised to .un--us the Esvuv .ying NRC SER.

USAR 98-45, lhe USAR was revised to make Section 9.2.1 consistent with Section 6.1.2.3 which states that the chemical and volume contml Section 9.2.1  !

CR system (CVCS) and the charging pumps are not engineered safeguards equipment.

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t Attachment C LIC-98-0139 i

i Regulatory Commitment Changes l

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LIC-98-0139 Attachment C REGULATORY COMMITMENT CllANGES The following Regulatory Commitments were revised based upon evaluations made in accordance with the NEI Guideline for Managing NRC Commitments. His NEI Guideline was accepted in SECY-95-300.

SOURCE OF REVISION TO COMMintENT COMMITMENT GL-88-17 in response to GL-88-17 OPPD committed to include the following precautions in the Reactor Coolant System Draining precedure: "De w ater level in the reactor coolant system must remain above the middle of the hot leg to provide a suction to the low pressure safety injectien pump for shutdown cooling re-circulation. If RCS water level is above the top of the hot leg nozzle, decrease level until the steam gencrator U tubes have dumped, prior to removing tool access flanges, the reactor vessel head, or work on the RCP seals. "

He response to this GL only needed to address inventory control when in reduced inventory conditions. Reduced inventory conditions begin below 1010 feet elevation. Changing this commitment would allow work on the tool access flanges, heated junction thermocouples and instrument bullet noses prior to dumping the steam generator "U" tubes. The low est of these openings is at 1020 feet elevation,10 feet above reduced inventory conditions. Opening of these flanges will not cause partial dumping of the SG "U tubes. Ilowever if the tubes were to lose vacuum, shutdown cooling would be unaffected and the RCS would remain 10 feet above the reduced inventory conditions.

Using the NEI guidelines this precaution was revised to remove the prerequisites conceming the tool access flanges and to read:

"The water level in the reactor coolant system me st remain above the middle of the hot leg to provide a suction to the low pressure safety injection pump for t shutdown cooling re-circulation. If RCS water level is above the top of the hot leg nozzle, decrease level until the steam generator U tubes have dumped, prior to removing the reactor vessel head or working on the RCP seals. This is to ensure that no unexpected rise in RCS water level takes place should the vacuum be inadvertently lost in the steam generator U-tubes. Level will increase on LI-197 when the U-tubes dump."

Amendment I83 In the request for amendment and the NRC Safety Evaluation Report issued by the NRC grantmg Amendment 183 to the FCS Operating License, the specific quantity and concentration " Forty percent ethanolamine solution is stored inside the plant in a 365 gallon container . ."

Using the NEI guidelines, which included a no significant hazards consideration, this Regulatory Commitment regarding the concentration of ethanolamine was revised to read:

One hundred percent (100 %) ethanolamine solution is stored inside the plant in a 365 gallon container . "

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