Text

Specifications to Adopt TSTF-545 Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, (LDCN 18-0021)
	ML19081A173
Person / Time
Site:	Callaway
Issue date:	03/22/2019
From:	Wink R; Ameren Missouri
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LDCN 18-0021, ULNRC-06478
	Download: ML19081A173 (62)
	v • d • e

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WAmeren MISSOURI Callaway Plant

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March 22, 2019 ULNRC-06478 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 10 CFR 50.90 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC COMPANY RENEWED FACILITY OPERATING LICENSE NPF-30 APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-545 REVISION 3, "TS INSERVICE TESTING PROGRAM REMOVAL & CLARIFY SR USAGE RULE APPLICATION TO SECTION 5.5 TESTING," {LDCN 18-0021)

Pursuant to 10 CFR 50.90, '"Application for amendment of license or construction permit," Union Electric Company (DBA Ameren Missouri) herewith transmits an application for amendment to Renewed Facility Operating License NPF-30 for the Callaway Plant. The proposed change revises the Technical Specifications (TS) to eliminate Section 5.5.8, 11 Inservice Test Program," and add the term, "Inservice Testing Progra;u," to the TS Definitions section. This request is consistent with Technical Specification Task Force (TSTF) TS traveler TSTF-545, Revision 3, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."

Essential/supporting information is provided in the attaclm1ents to this letter. Specifically, Attachment 1 provides a description and assessment of the proposed TS changes. Attaclm1eut 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised (clean)

TS pages to show the proposed changes. Attachment 4 provides TS Bases pages marked up to show the associated TS Bases changes, which are provided for information only.

It has been determined that this amendment application does not involve a significant hazards consideration as detennined per 10 CPR 50.92, "Issuance of amendment. ' Further, pursuant to 10 CFR 51.22 , "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,' Section (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

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8315 County Road 459

Steedman, MO 65077 AmerenMissouri.com

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ULNRC-06478 March 22, 2019 Page 2 of 4 This amendment application was reviewed by the Onsite Review Committee. lu accordance with 10 CFR 50 .91 , "Notice for public comment; State consultation." Section (b )(l) a copy of this amendment application is being provided to the designated Missonri State official.

Ameren Missouri requests approval ofthjs license amendment request prior to March 22, 2020. It is anticipated that the license amendment, as approved, will be effective upon issuance and will be implemented within 90 days from the date of issuance.

If you have any questions concerning this matter, please contact me at 573 -310-7025 or Mr. Tom Elwood at 314-225-1905.

It should be noted that this submittal does not contain any new conunitments.

l declare under penalty of pe1jurythat the foregoing is true and correct.

Sincerely, Executed on: 4~ :l ~... d).017 9E~w:!-' Manager, Regulatory Affairs Attachments:

1. Description and Assessment of Proposed Technical Specification Changes.

2. Existing TS Pages Marked Up to Show the Proposed Changes

3. Revised (Clean) TS Pages Reflecting the Proposed Changes

4. TS Bases Pages Marked Up to Show the Associated TS Bases Changes (provided for information only)

ULNRC-06478 March 22, 2019 Page 3 of 4 cc: Mr. Scott A. Morris Regional Administrator U.S. Nuclear Regul'a tory Commission RegionIV 1600 East Lamar Boulevard Arlington TX 76011-45 I 1 Senior Resident Inspector Callaway Resident Office U.S . Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. L. Jolm Klos Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U . S. Nuclear Regulatory Commission Mail Stop 09E3 Washington, DC 20555-0001

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ULNRC-064 78 March 22 2019 Page 4 of 4 Index an.d send hard copy to QA File A160.0761 Hardcopy:

Certrec Corporation 6 I 00 Western Place, Suite 1050 Fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Tech Spec ULNRC Distribution:

F. M. Diya T. E. Henmann B. L. Cox F. J. Bianco S. P. Banker L. H. Kanuckel R. C. Wink T. B. Elwood Corporate Communications NSRB Secretary STARS Regulatory Affairs Mr. Jay Silberg (Pillsbury Winthrop Shaw Pittman LlLP)

Missomi Public Service Commission Mr. Dru Buntin (DNR)

.Mr. Steve Feeler (DNR)

Attachment I to ULNRC-06478 Pagel of 5 ATTACHMENT!

DESCRIPTION AND ASSESSMENT OF PROPOSED TECHNICAL SPECIFICATIONS CHANGES to ULNRC-06478 Page 2 of 5

1.0 DESCRIPTION

The proposed change eJimjnates Section 5.5.8, 1'lnservice Test (1ST) Program," of the Technical Specifications (TS) in order to remove requirements duplicated in American Society of Mechanical Engineers (ASME) Code for Operations and Maintenance of Nuclear Power Plants (OM Code), Case OMN-20, "Inservice Test Frequency," which is approved for use in the Callaway Plant IST program. A new defined term, '1Inservice Testing Program "is added to TS Section 1. l , "Definitions.

The proposed change to the TS is consistent with TSTF-545 Revision 3, TS Inservice Testing PFogram Removal & Clarify SR Usage Rule Application to Section 5.5 Testing."

2. 1 ASSESSMENT 2.2 Applicability of Pub)jshed Safetv Evaluation Union Electric Company (Ameren Missouri) bas reviewed the model safety evaluation provided to the Technical Specifications Task Force in a letter, "Final Model Safety Evaluation of Technical Specifications Task Force Traveler TSTF-545, Revision 3, 'TS lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing,' '1 dated December 11 , 2015 (NRC ADAMS Accession No. ML15317A071) . The review included a review of the NRC staffs evaluation as well as the infonnation provided in TSTF-545 .

Ameren Missouri concluded that the justifications presented in TSTF-545, as well as the model safety evaluation prepared by the NRC staff, are applicable to Callaway Plant, Unit 1 i:md justify tbe requested amendment for incorporation of the changes to the Callaway Plant TS .

CaJlaway Plant, Unit 1 was issued a construGtion pennit on April 16, 1976, such that the provisions of 10 CFR 50.55a(i)(3) are applicable.

2.3 Variations Ameren Missouri is not proposing any variations from the TS changes described in the TSTF-545 or the applicable paits of the NRC staffs model safety evaluation dated December 11 ,

2015.

3.1 REGULATORY ANALYSJS 3.2 No Significant Hazards Consideration Analysis Union Electric Company Ameren Missouri) requests adoption ofthe Technical Specification (TS) changes described in TSTF-545, "TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing," which is an approved change to the Improved Standard Technical Specifications (ISTS), into the Callaway Unit I Technical Specifications, The proposed change revises the TS Chapter 5, Administrative Controls " Section 5.5, "Programs and Manuals," to delete TS 5.5 .8, "Inservice Testing (1ST) Program . Requirements in the 1ST Program described or specified in TS 5.5.8 are removed, as they are duplicative of to ULNRC-06478 Page 3 of 5 requirements in the American Society of Mechanical Engineers (ASME) Operations and Maintenance (OM) Code, as clarified by Code Case OMN~20 Inservice Test Frequency."

Other requirements or provisions in TS 5.5 .8 are eliminated because the Nuclear Regulatory Commission (NRC) has determined their appearance in the TS is contrary to regulations. A new defined term, "Inservice Testing Program, " which references the requiremeHts of Title 10 of the Code of Federal Regu.lations (10 CFR), Part 50, paragraph 50.55a(f), is added to Section 1.1, "Definitions," of the TS . Ameren Missouri has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50 .92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change revises TS Chapter 5, "Administrative Controls, " Section 5.5, "Programs and Manuals," by eliminating the "lnservice Testing Program 11 specification (i.e., TS 5.5.8). Most requirements in the Inservice Testing Program are removed, as they are duplicative of requirements in the ASME OM Code, as clarified by Code Case O:tvlN-20, "lnservice Test Frequency." The remaining requirements in the Section 5.5 1ST Program description are eliminated because the NRC has dete1mined their inclusion in the TS is contrary to regulations. A new defined term, "Inservice Testing Program "

is added to Section 1.1 of the TS, which references the requirements of IO CFR 50.55a(f).

Performance of inservice testing is not an initiator to any accident previously evaluated.

As a result, the probability of occurrence of an accident is not significantly affected by the proposed change. lnservice test frequencies under Code Case OMN-20 are equivalent to the current testing periods allowed by the TS with the exception that test intervals greater than 2 years may be extended by up to 6 months to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance of the required testing. The testing frequency extension will not affect the ability of the components to mitigate any accident previously evaluated., as the components are required to be operable during the testing period extension .

Performance of inservice tests utilizing the allowances in OMN-20 will :not significantly affect the reliability of the tested components. As a result, the availability of the affected components, as well as their ability to mitigate the consequences of accidents previously evaluated., is not affected.

Therefore, the proposed change does not involve a* significant increase in the probability or consequences of an accident previously evaluated.

ttachrnent 1 ito ULNRC-06478 Page 4 of 5 2, Does the pr,oposed change create the possibifity of a new or different kind of accident from any previously evaluated?

Response: No The proposed change does not alter the design or configuration of the plant. The proposed change does not involve a physical altei-ation of the plant; no new or different kind of equipment wil l be installed. The proposed change does not alter the types of inservice testing performed. In most cases, the frequency of inservice testing is unchanged. However the frequency of testing would not result in a new or different kind of accident from any previously evaluated since the testing methods are not altered.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated .

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change eliminates some requirements from the TS in lieu ofrequirements in the ASME Code, as modified by use of Code Case OMN-20. Compliance with the ASME Code is required by 10 CFR 50.55a. The proposed change also allows inservice tests with test intervals greater than 2 years to be extended by 6 months (consistent with code case OMN-20) to facilitate test scheduling and consideration of plant operating conditions that may not be suitable for performance oftheJ*equired testing. The testing frequency e}..1:ension will not affect the ability of the components to respond to an accident as the components are required to be operable during the testing period extension. The proposed change also eliminates a statement that nothing in the ASME Code should be construed to supetsede the requirements of any TS. The NRC has determined that statement to be incorrect. However, elimination of the statement will have no effect on plant operntion or safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Ameren Missouri concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CPR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to insta1lation or use of a facility component located within the restricted area, as defined in 10 CFR 20 or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a sign ificant hazards consideration, (ii) a significant change in the types or significant increase in to ULNRC-06478 Page 5 of 5 the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CPR 51.22(c)(9). Based on the above, pursuant to 10 CPR 5 l .22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

to ULNRC-064 78 Page l of 15 ATTACHMENT 2 EXISTING TS PAGES MARKED UP TO SHOW THE PROPOSED CHANGES to ULNRC-064 78 Page 2 of 15 CHAPTER TABLE OF CONTENTS CHAPTER 5. 0 ADM INISTRATIVE CONTROLS Section 5.1 Responsibility ... ................. ........... .. .. ....... ........ .......... .. ..... .. .... .... .. ..... .... .... .......... 5.0-1 5.2 Organ ization ...... .... .... ...... .... .. .... ................. ......... ...... ........ ....... .. ... ......... .... .. ..... .. 5.0-2

5. 2.1 Onsite and Offsite Organizations ..................... .......................... ................. .. 5.0-2 5.2.2 Unit Staff ..... ...... .. ......... ...... ......................... .. ............ .. ............ .......... .. ........... 5.0-2 5.3 Unit Staff Qualifications .. .. .... ............................. ............ ............ ... ...... ... ............. . 5.0-4 5.4 Procedures ....... .................. .... ..... .. ...................... ...... ...... .. ........ .... .............. ........ 5.0-5 5.5 Programs and Manuals ... .. ........ ..... ........ .. .. ...... .... .......... .... ... ..... .... .. ... .... ... .. ........ 5.0-6 5.5 .1 Offsite Dose Calculation Manual (ODCM) ....... ... ............ ............................... 5.0-6 5.5.2 Primary Coolant Sources Outside Containment .............. ......... ...... .... ........... 5.0-7 5.5.3 Not Used .. ... .... .... .... .. .... ........ ... .... ... ..... ,......... ... ...... ... ...... ... .... ...... .. ...... ... ... .. 5.0-7 5.5.4 Radioactive Effluent Controls Program .... .... .... ... .......................................... 5.0-7 5.5 .5 Component Cyclic or Transient Limit .... ......... ...... ...... ... ....... .. .................... ... 5.0-9 5.5.6 Containment Tendon Surveillance Program ..... .. ... ......... .. ........ ... ................. 5.0-9 5.5.7 Reactor Coolant Pump Flywheel Inspection Program .. ........ .............. ...... ..... 5.0-9 5.5.8 lnservice Testing ProgramNot Used .... ... ...... .. .......... ............. ........ ...... .. .. ... 5.0-10 5.5.9 Steam Generator (SG) Program ..... ..... ..... ................................... .. ....... ....... 5.0-10 5.5.10 Secondary Water Chemistry Program ...... ...... ...... .. ..... .... .... ,.... ...... ......... ... . 5.0-1 4 5.5.1 1 Ventilation Filter Testing Program (VFTP) ...................... ...... ...................... 5.0-14 5.5.12 Explosive Gas and Storage Tank Radioactrvity Monitoring Program .. ...... .. 5.0-16 5.5. 13 Diesel Fuel Oil Testing Program .. ........... .. .................. ........ .. ...... .... ... ......... 5.0-17 5.5.14 Techn ical Specifications (TS) Bases Control Program .... .... .. .. ................ ... 5.0-18 5.5.15 Safety Function Determination Program (SFDP) ........................................ 5.0-18 5.5. 16 Containment Leakage Rate Testing Program ... .. .... ...... ............. ..... .... ... .. ... 5.0-19 5.5. 17 Control Room Envelope Habitability Program ... ......... ............. .. ......... ...... ... 5.0-21 5.5.1 8 Surveillance Frequency Control Program .................. ... .. .. ......... .. ........ ..... ... . 5.0-22 5.6 Reporting Requirements ...... .......... .............. .. ... .... .. ...... ........ ................... ......... 5.0-23 5.6.1 Not Used . .. .... .. ........... ......... .. .................... .. ..... ...... ... .. ....... ......................... 5.0-23 5.6.2 Annual Radiological Environ mental Operating Report .................. .............. 5.0-23 5.6.3 Radioactive Effluent Release Report ......... ......... ........................................ 5.0-23 5.6.4 Not used . .. ,......... ...... .. ..... ....... ........ .......... ... .... .. .... .......... ...... ...... ....... ... ... .. . 5. 0-23 CALLAWAY PLANT 5.0-i to ULNRC-06478 Page 3 of 15 Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-87, Kr-88, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specifiic noble gas nuclide is not detected , it should be assumed to be present at the minimum detectable activity.

The determination of DOSE EQUIVALENT XE-133 shall be performed using the effective dose conversion factors for air submersion listed in Table 111.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081 , "External Exposure to Radionuclides in Air, Water, and Soil", 1993.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE(ESF)RESPONS E when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i .e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any serfes of sequential, overlapping, or total steps so that the entire response time is measured . In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requireme nts of 10 CFR 50.55a(O .

LEAKAGE LEAKAGE shall be :

a. Identified LEAKAGE

1. LEAKAGE , such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water leakoff) , that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operatioh of leakage detection systems or not to be pressure boundary LEAKAGE ; or (continued)

CALLAWAY PLANT 1.1-3 Amendment 178 to ULNRC-064 78 Page4 of 15 Pressurizer Safety Valves 3.4.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verify each pressurizer safety valve is OPERABLE in In accordance with accordance with the lnservice Testing the INSERVICE ProgramlNSERVICE TESTING PROGRAM. TESTING Following testing, lift settings shall be within +/- 1% of PROGRAM lnservi 2460 psig. cc Testing Program CALLAWAY PLANT 3.4-21 Amendment No. 213 l

A ttachment 2 to ULNRC-06478 Page 5 of 15 Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11 .1 - - - - NOTE - - - -

Not required to be performed with block valve closed in accordance with the Required Actions of this LCO.

Perform a complete cycle of each block valve. In accordance with the Surveillance Frequency Control Program SR 3.4.11 .2 Perform a complete cycle of each PORV. In accordance with the INSERVICE TESTING PROGRAM lnservl ce Tosl-ing Progrom CALLAWAY PLANT 3.4-25 Amendment No. 213 I to ULNRC-06478 Page 6 of 15 RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 ----------------- NOTES - - - -*- - - - -

1. Not required to be performed in MODES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is equivalent to In accordance with

$: 0.5 gpm per nominal inch of valve size up to a the INSERVICE maximum of 5 gpm at an RCS pressure 2:'.' 2215 psig TESTING and ~ 2255 psig. PROGRAMIAservr cc Testing Program ,

In accordance with the Surveillance Frequency Control Program CALLAWAY PLANT 3.4-35 Amendment No. 213 I to ULNRC-06478 Page 7 of 15 ECCS - Operating 3 .5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the INSERVICE developed head, TESTING PROGRAM~

ce testing pro§ram SR 3.5.2.5 Verify eact.i ECCS automatic valve in the flow path In accordance that is hot locked, sealed, or otherwise secured in with the position , actuates to the correct position on an actual Surveillance or simula_ted actuation signal. Frequency Control Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance actual or simulated actuation signal. with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, In accordance each mechanical position stop is in the correct with the position . Surveillance Valve Number Frequency Control Program EMV0095 EMV0107 EMV0089 EMV0096 EMV0108 EMV0090 EMV0097 EMV0109 EMV0091 EMV0098 EMV0110 EMV0092 SR 3.5.2.8 Verify, by visual inspection, each ECCS train In accordance containment sump suction inlet is not restricted by with the debris and the suction inlet strainers show no Surveillance evidence of structural distress or abnormal corrosion. Frequency Control Program CALLAWAY PLANT 3.5-6 Amendment No. 202 to ULNRC-06478 Page 8 of 15 Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

StJRVEILLANCE FREQUENCY SR 3.6,3.4 - - - - - - - - - - - - NOTE - - - -- -- - - - - -

Valves and blind flanges in high radiation areas may be verified by use of administrative means_

Verify each containment isolation manual valve and Prior to entering blfnd flange that is located inside containment and not MODE 4 from locked, sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed , except performed within for containment isolation valves that are open under the previous administrative controls. 92 days SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with operated containment isolation valve is within limits. the INSERVICE TESTJNG PROGRAMlnservi ce Testing Program SR 3.6.3.6 - - - - - - -. NOTE------------

Only required to be performed when containment shutdown purge valve blind flanges are installed.

Perform leakage rate testing for containment In accordance shutdown purge valves with resilient seals amd with the associated blind flanges. Surveillance Frequency Control Program Following each reinstallation of the blind flange (continued)

CALLAWAY PLANT 3.6-14 Amendment No. 202 to ULNRC-06478 Page 9 of 15 Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.3 Verify each containment cooling train cooling water IR accordance flow rate is ~ 2200 gpm. with the Surveillance Frequency Control Program SR 3.6.6.4 Verify each containment spray pump's developed In accordance with head at the flow test point is greater than or equal to the INSERVICE the required developed head. T ESTING PROGRAM lnser1i ce Test:lng Pffi~

SR 3.6.6.5 Verify each automatic containment spray valve in the In accordance flow path that is not locked1, sealed , or otherwise with the secured in position , actuates to the correct positior.i on Surveillance an actual or simulated actuation signal. Frequency Control Program SR 3.6.6.6 Verify each containment spray pump starts In accordance automatically on an actual or simulated actuation with the signal. Surveillance Frequency Control Progra m SR 3.6.6.7 Verify each containment cooling train starts In accordance automatically and minimum cooling water flow rate is with the established on an actual or simulated actuation Surveillance signal. Frequency Control Program (contintied)

CALLAWAY PLANT 3.6-20 Amendment No. 213 to ULNRC-06478 Page 10 of 15 MSSVs 3.7.1 ACTIONS (continued)

CONDITION COMPLETION REQUIRED ACTION TIME B. (continued) B.2 - - - - NOTE - - - - -

Only required in Mode 1.

Reduce the Power 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Range Neutron Flux -

High Trip setpoir.its to less than or equal to the Maximum Allowable %

RTP specified in Tabl e 3.7.1-1 for the number of OPERABLE MSSVs.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not r:net. AND OR C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators witli 2".. 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1. 1 --------- ----- NOTE-----------

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per Table In accordance with 3.7.1-2 in accordance with the INSERVICE TESTING the INSERVICE PROGRAMlnservice Testing Program . Following TESTING testing , lift setting shall be within +/- 1 %. PROGRAM lnsGFVi ce Testing Pfe§fciffi CALLAWAY PLANT 3.7-2 Amendment No. 133 to ULNRC-06478 Page 11 of 15 MS1Vs, MSIVBVs, and MSLPDIVs

3. 7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify isolation time of each MSIV is within lirnits. In accordance with the INSERVICE TESTING PROGRAM lnservi ce Testing Program SR 3.7 .2.2 Verify each MSIV, each MSJVBV, and each MSLPDIV In accordance actuates to the isolation position on an actual or with the simulated actuation signal. Surveillance Frequency Control Program SR 3.7.2.3 Verify isolation time of each MSIVBV and MS LP DIV is In accordance with within limits. the INSERVICE TESTING PROGRAM Inservi ce Testing:

Program CALLAWAY PLANT 3.7-8 Amendment No. 202 to ULNRC-06478 Page 12 of15 MFIVs 1 MFRVs, and MFRVBVs 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FR~QUENCY SR 3.7.3.1 ------------- * - - --- - NOTE---~-------

Only required to be performed 1in MODES 1 and 2.

Verify the closure time of each MFRV and MFRVBV 1is In accordance with within lirnits. the INSERVICE TESTING PROGRAM lnservi ce Testing Program SR 3.7.3.2 ---- - **---------NOTE - - - - - - -- -

For the MFRVs and MFRVBVs , only required to be performed in MODES 1 and 2.

Verify each MFIV, MFRVand MFRVBV actuates to In accordance the isolation position on an actual or simulated with the actuation signal. Surveillance Frequency Control Program SR 3.7.3.3 Verify the closure time of each MFIV is within limits. In accordance with the INSERVICE.

TESTING PROGRAMlnservi cc Testing Program CALLAWAY PLANT 3.7-11 Amendment 202 t o ULNRG-06478 Page 13 of15 ASDs 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 ---- -- NOTE------ - - - - - - -

Only required to be performed in MODES 1 and 2.

Verify one complete cycle of each ASD . In accordance with the INSERVICE TESTING PROGRAM lnservi ce Testing Program SR 3.7.4.2 Verify one complete cycle of each ASD manual In accordance with isolation valve. *the INSERVICE TESTING PROGRAMI nserv1 ce Te~ing Program CALLAWAY PLANT 3.7-14 Amendment No. 198 I to ULNRC-06478 Page 14 of 15 SSIVs 3.7 .19 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. Required Action and 8.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.19.1 Verify the iso'lation time of each automatic SSIV is In accordance with within limits. the INSERVICE TESTING PROGRAM lnservi ceTest1ng-Program SR 3.7 .19.2 Verify each automatic SSIV in the flow path actuates In accordance to the isolation position on an actual or simulated with the actuation signal. Surveillance Frequency Control Program CALLAWAY PLANT 3.7-47 Amendment No. 202 to ULNRC-06478 Page 15 ofl5 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 lnservice Testing ProgramNot Used This program provides sontrols fur inservice testing of ASM!i: Code Class 1, 2, a

and eomponents. The program shall include the fullowing:

a. Testing frequencies app licable to the ASME Code for Operation and Maintenance of ~uclear Power Plants (ASME OM Code) and applicable

/\ddenda as fullows :

/\SME. OM Gode and applisable /\ddenda terminology for Required i;::requencies inservice testing for performing inservice activities testing activities V\leekly At least once per 7 days Monthly /\t least once per 31 days Quarterly or every 3 months /\t least once per 92 days Semiannually or evel)1 0 months At least once per 184 days Evel)' 9 months At least once per 270 days Yearly or annually At least once per 300 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities; C. The provisions of SR 3.0.3 are applicable to inservioe testing aotivities; aM

d. ~Jothing in the ASME OM Gode shall be construed to supersede the requirements of any TS.

5.5.9 Steam Generator (SG} Program A Steam Generator Program shall be established and implemer;ited to ensure that SG t1:.1be integrity is maintained. In addition , the Steam Generator Program shall include the forlowing :

a. Provisions for condition monitoring assessments . Condition monitoring assessment means an evaluation of the "as found " condition of the tubing (continued)

CALLAWAY PLANT 5.0-10 Amendment No. 215

Attaclm1ent 3 to ULNRC-06478 Page 1 of 15 ATTACHMENT 3 REVISED (CLEAN) TS PAGES TO SHOW THE PROPOSED CHANGES

Attachment 3 to ULNRC-064 78 Page 2 of 15 CHAPTER TABLE OF CONTENTS CHAPTER 5.0 ADMINISTRATIVE CONTROLS Section 5.1 Responsibility ........... ....... ... .... ......... ...... ..... ... ........................ ......... ....... ..... .... .... . 5.0-1 5.2 Organization ........ ........... .. .. ... ....... ..... ..... ..................................... ......... ..... .. ..... ... 5.0-2 5.2.1 Onsite and Offsite Organizations .................. .. .... ... .. ......... ............................ 5.0-2 5.2.2 Unit Staff ...... ..... ...... .... ... .. .... .. ............... ....... ........ .............................. ..... ....... 5.0-2 5.3 Unit Staff Qualifications .... .... ... .. .. ...... ...... .... ... ..... ........ .. ... ......... .. ... ... ......... .. ....... 5.0-4 5.4 Procedures ..... ... .. ........ .... .. .. ............ ... ...... .... ........... ....... .... .......... ... ............ ........ 5.0-5 5.5 Programs and Manuals ....... ... ................ .. .... .,...... .. ... ......... .... ........ .... ............ .. .... 5.0-6 5.5.1 Offsite Dose Calcu lation Manual (ODCM) ......................... ............................ 5.0-6 5.5.2 Primary Coolant Sources Outside Containment ....... ........ ... .. .... .. ......... .. .. ..... 5.0-7 5.5.3 Not Used .. ... .. .... .. .... ........ .. ...... ... ......... .... ...... ...... .......... ............. .... .... ... ...... .. 5.0-7 5.5.4 Radioactive Effluent Controls Program ....... ...... ... .... .. .... .. .. .... .... ..... ... ... ..... ... 5.0-7 5.5.5 Component Cyclic or Transient Limit ..... ... ... ......... ....... .. .... .... ... .... ........... .. ... 5.0-9 5.5.6 Containment Tendon Surveillance Program ............ ......... ....................... .. ... 5.0-9 5.5.7 Reactor Coolant Pump Flywheel Inspection Program ....... ...... ... .. ........... .. .. .. 5.0-9 5.5.8 Not Used .... ............... ..... ...... .. ........ .. ... .. .... .. .... ... ..... ... ..... .. .. .......... ... ........... 5.0-10 5.5.9 Steam Generator (SG) Program , ............ ......... ................................. ........ ... 5.0-10 5.5.1 0 Secondary Water Chemistry Program .. ................ .. ... .. .................... .. ....... ... 5.0-14 5.5 .11 Ventilation Filter Testing Prog ram (VFTP) .. ........... ........ .. ... .. ............... .... .. . 5.0-14 5.5.12 l;xplosive Gas and Storage Tank Radioactivity Monitoring Program .... ... .. . 5.0-16 5.5.13 Diesel Fuel Oil Testing Program ....... .. ...... ... .......... ............... ........... .. ......... 5.0-17 5.5.14 Technical Specifications (TS) Bases Control Program ..... ................ .. ........ 5.0-18 5.5.15 Safety Function Determination Program (SFDP) ... .. ... ........ .. ........... .. ...... .. . 5.0-18 5.5.16 Containment Leakage Rate Testing Program ........................................ ..... 5.0-19 5.5.17 Control Room Envelope Habitability Program ...... .... ................. ..... ... .. ........ 5.0-21 5.5.18 Surveillance Frequency Control Program ...... ....... .. ...... .......................... ...... 5.0-22 5.6 Reporting Requ irements .. ...... ... ... ..... .. ... ........ .. ... ........ ............. .. ........ ........ .. .. ... 5.0-23 5.6.1 Not Used. ................... ............... .. ....... ... ............. ... ................. ... .. ............. .. . 5.0-23 5.6.2 Annual Radiological Environmental Operating Report .. ....... .. .......... ... ..... ... 5.0-23 5.6.3 Radioactrve Effluent Release Report .. .... ... .... .. .. .. ......... .. .......... ............ ...... 5.0-23 5.6.4 Not used . .. ........... .. .... ... .. .. .. ..................... ...... .... .. .... .. ........ ................ .......... 5.0-23 CALLAWAY PLANT 5.0-i to ULNRC-064 78 Page 3 ofl 5 Definitions 1.1 1.1 Definitions (continued)

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-1,33 (microcuries per gram) that alone would produce the same aeute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-87, Kr-88 , Xe-133m , Xe-133, Xe-135m , Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should' be assumed to be present at the minimum detectable activity.

The determination of DOSE EQUIVALENT XE-133 shall be performed using the effective dose conversion factors for air submersion listed In Table 111.1 of EPA Federal Guidance Report No. 12, EPA-402-R-93-081 , 1o External Exposure to Radionuclides in Air, Water, and Soil", 1993.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE(ESF)R ESPONSE when the monitored parameter exceeds its ESF actuation TIME setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions , pump discharge pressures reach their required values , etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping , or total steps so that the entire response time is measured . In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been ,previously reviewed and approved by the NRC .

INSERVICE TESTING The IN SERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requ irements of 10 CFR 50.55a(f) .

LEAKAGE LEAKAGE shc:111 be:

a. Identified LEAKAGE

1. LEAKAGE , such as that from pump sec1ls or valve packing (except reactor coolant pump (RCP) seal water leakoff) , that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and l<nown either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE i or (continued)

CALLAWAY PLANT 1.1-3 Amendment XXX to ULNRC-06478 Page 4 of l 5 Pressurizer Safety Valves 3.4.1 O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4 .1 0.1 Verify each pressurizer safety valve is OPERABLE in In accordance with accordance with the IN SERVICE TESTING the INSERVICE PROGRAM . Following testing , lift settings shall be TESTING within +/- 1% of 2460 psig. PROGRAM CALLAWAY PLANT 3.4-21 Amendment No. XXX to ULNRC-06478 Page 5 of 15 Pressurizer PORVs 3A.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11 .1 - - - - - - - - - - - - - NOTE -

Not required to be performed with block valve closed in accordance with the Required Actions of this LCO .

Perform a complete cycle of each block valve. In accordance with the Surveillance Frequency Control Program SR 3.4.11 .2 Perform a complete cycle of each PORV. In accordance with the INSERVICE TESTING PROGRAM CALLAWAY PLANT 3.4-25 Amendment No. XXX to ULNRC-064 78 Page 6 of 15 RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.14.1 - - - - - - - - - - - - - NOTES - - - - - - - - - - --

1. Not required to be performed in MODES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in th e shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided .

Verify leakage from each RCS PIV is equivalent to In accordance with

$ 0.5 gpm per nominal inch of valve size up to a the INSERVICE maximum of 5 gpm at an RCS pressure~ 2215 psig TESTING and $ 2255 psig. PROGRAM ,

AND fn accordance with the Surveillance Frequency Control Program (continued)

CALLAWAY PLANT 3.4-35 Amendment No. XXX to ULNRC-06478 Page 7 of15 ECCS - Operating 3.5. 2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3,5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the INSERVICE developed head. TESTING PROGRAM SR 3.5.2.5 Verify each ECCS automatic valve in the Oow path In accordance that is not locked, sealed, or otherwise secured in with the position , actuates to the correct position on an actual Surveillance or simulated actuation signal. Frequency Control Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an In accordance actual or simulated actuation signal. with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, In accordance each mechanical position stop is in the correct with the position. Surveillance Valve Number Frequency Control Program EMV0095 EMV0107 EMV0089 EMV0096 EMV0108 EMV0090 EMV0097 EMV0109 EMV0091 EMV009f3 EMV0110 EMV0092 SR 3.5.2.8 Verify, by visl!lal inspection, each ECCS train In accordance containment sump suction inlet is not restricted by with the debris and the suction inlet strainers show no Su rveilla nee evidence of structural distress or abnormal corrosion. Frequency Control Program CALLAWAY PLANT 3.5-6 Amen dment N.o. XXX to ULNRC-064 78 Page 8 of 15 Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.4 ----- - -*- - - - - - - - NOTE-------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve and Prior to entering.

blind flange that is located inside containment and not MODE 4from locked , sealed, or otherwise secured and required to MODE 5 if not be closed during accident conditions is closed , except performed within for containment isolation valves that are open under the previous administrative controls . 92 days SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with operated containment isolation valve is within limits. the INSERVICE TESTING PROGRAM SR 3.6.3.6 ------------------ NOTE - - - - - ~ - - ---

Only required to be performed when containment shutdown purge valve blind flanges are installed.

Perform leakage rate testing for containment In accordance shutdown purge valves with resilient seals and with the associated blind flanges. Surveillance Frequency Control Program F ollowlng each reir.istallation of the brltid flange (contimied)

CALLAWAY PLANT 3.6-1 4 Amendment No. XXX to ULNRC~06478 Page 9 of 15 Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.3 Verify each containment cooling train cooling water In accordance flow rate is 2: 2200 gpm. with the Surveillar;ice Frequency Control Program SR 3.6.6.4 Verify each containment spray pump's developed Jn accordance witl1 head at the fJow test 1poJnt is greater than or equal to the INSERVICE the required developed head. TESTING PROGRAM SR 3.6.6.5 Verify each automatic containment spray valve in the In accordance flow path that is not locked, sealed , or otherwise With the secured in position , actuates to the correct position on Surveillance an actual or simulated actuation signal. Frequency Control Program SR 3.6.6.6 Verify each containment spray pllmp starts In accordance automatically on an actual or simulated actuation with the signal. Surveillance Frequency Control Program SR 3.6.6.7 Verify each containment cooling train starts In accordance automatically and minimum cooling water flow rate is with the establishe.d on an actual or simulated actuation Surveillance signal. Frequency Control Program (continued )

CALLAWAY PLANT 3.6-20 Amendment No, XXX to ULNRC-064 78 Page 10 of 15 MSSVs 3.7.1 ACTIONS (continued)

CONDITION COMPLETION REQUIRED ACTION TIME B. (continued) B.2 --------- NOTE ------

Only required in Mode 1.

Reduce the Power 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Range Neutron Flux -

High Trip setpoints to less than or equal to the Maximum Allowable %

RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

C. Required Action and C .1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators with ~ 4 MSSVs inoperable.

SURVEILLANCE REQUIREMENTS SURVEI Ll!..ANCE FREQUENCY SR 3.7 .1.1 -------------- --NOTE-- . - - - - - - - -

Only required to be performed in MODES 1 and 2.

Verify each required MSSV lift setpoint per Table In accordance with 3.7 .1-2 in accordance with the INSERVICE TESTING the INSERVICE PROGRAM . Following testing , lift setting shall be TESTING within +/- 1%. PROGRAM CALLAWAY PLANT 3.7-2 Amendment No. XXX to ULNRC-064 78 Page 11 of 15 MSIVs, MS1V8Vs , and MSLPDIVs 3.7.2 SURVEIL;LANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify isolation time of each MSIV is within limits. In accordance with the INSERVICE TESTING PROGRAM SR 3.7.2 .2 Verify each MSIV, each MSIVBV, and each MSLPDIV In accordance actuates to the isolation position on an actual or with the simulated actuation signal. Surveillance Frequency Control Program SR 3.7.2 .3 Verlfy isolation time of each MSIVBV and MSLPDIV is In accordance with within limits. the INSERVICE TESTING PROGRAM CALLAWAY PLANT 3.7-8 Amendment No. XXX to ULNRC-064 78 Page 12 of1 5 MFIVs, MFRVs, and MFRVBVs 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 ----------------NO TE----* - - - - - - -

Only required to be performed in MODES 1 and 2.

Verify the closure time of each MFRVand MFRVBV is In accordancewrth within limits. the INSERVICE TESTING PROGRAM SR 3.7 .3.2 - - - - - - - - - - - - - - NOTE------------

For the MF RVs and MFRVBVs, only required to be performed in MODES 1 and 2.

Verify each MFIV, MFRV and MFRVBV actuates to In accordance the isolation position on an actual or simulated with the actuation signal. Surveinance Frequency Control Program SR 3.7.3.3 Verify the closure time of each MFIV is withfn limits. In accordance with the INSERVICE TESTING PROGRAM CALLAWAY PLANT 3.7-11 Amendment XXX to ULNRC-06478 Page 13 of 15 ASDs 3.7.4 SURV~ILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4 .1 --------------------- NOTE Only required to be performed in MODES 1 and 2.

Verify one complete cycle of each ASD . In accordance with the INSERVICE TESTING PROGRAM SR 3.7.4.2 Verify one complete cycle of each ASD manual In accordance with isolation valve. the INSERVICE TESTING PROGRAM CALLAWAY PLANT 3.7-14 Amendment No. XXX

ttachm ent 3 to ULNRC-064 78 Page 14 of 15 SSIVs 3.7.19 ACTIONS (continued)

CONDITION COMPLETION REQUIRED ACTION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1 9.1 Verify the isolation time of each automatic SSIV is In accordance with within limits. the INSERVICE TESTING PROGRAM SR 3.7.19.2 Verify each automatic SSIV in the flow path actuates In accordance to the isolation position on an actual or simulated with the actuation signal. Surveillance Frequency Control Program CALLAWAY PLANT 3.7-47 Amendment No. XXX

Attaohment 3 to ULNRC-06478 Page 15 of 15 Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.8 Not Used 5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented -to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for conditron monitoring assessments. Condition mon itoring assessment means an evaluation of the "as found " condition of the tubing (continued)

CALLAWAY PLANT 5.0-10 Amendment No. XXX to ULNRC-064 78 Page 1 of22 ATTACHMENT 4 TS BASES PAGES MARKED lJP TO SHOW THE ASSOCIATED TS BASES CHANGES (provided for information only)

LCO Appltcability 8 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.8 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated. SR 3.0.2 and SR 3.0. 3 apply in Chapter 5 only when invoked by a Chapter 5 Specification.

LCO 3.0.1 LCO 3.0.1 establishes the .Applicability statement within each individual Specification as the requirement for when the LCO is required to be met 1

(i.e ., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCO 3,0.2 establishes that upon discovery of a failure to meet an LCO ,

the associated ACTIONS shall be met. rhe Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:

a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification ; and

b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.

There are two basic types of Req1:1ired Actions . The first type of Required Action specifies a time limit in which tl:le LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits , If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS .) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not f1:1rther restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

(continued)

CALLAWAY PLANT B 3.0-1 Revision 10

SR Applicability B 3.0 BASES SR 3.0.1 to be met and performed ln accordance with SR 3.0.2, prior to returning (continued) equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE . This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2 . Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function . This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed .

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per .. . " interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating cond itions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities) .

When a Section 5.5. "Programs and Manuals, specification states that the provisions of SR 3.0.2 are applicable. a 25% extension of the testing interval. whether stated in the Specification or incorporated by reference, is permitted.

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.

These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS . Therefore, when a test interval is specified in the regulations directly or by reference, the test interval cannot be extended by the TS ... and t11e SR or program req1:J!rerncnts include a ~late stating "SR a.0.2 is not applicable." /\n (continued)

CALLAWAY PLANT B 3.0-12 Revision 10

SR Applicability B 3.0 BASES SR 3.0.2 example of an exception when the test interval is not specified ln the (contiriued) regulations is the Note in the Containment Leakage Rate Testing Program

, "SR a.0.2 is not applicable. " This 9*ception is provided because the program already ineludes extension of test intervals. Examples of where SR 3.0.2 does not apply are the Containment Leakage Rate Testing Program (Section 5.5.16) as required by 10 CFR 50, Appendix J. and the testing of pumps and valves in the INSERVICE TESTING PROGRAM established per the American Society of Mechanical Engineers Operations and Maintenance (OM) Code. as required by 10 CFR 50.55a.

Extensions of test intervals under these programs are only as allowed by the programs and thei r regulations and not by any provisions in the Technical Specifications.

As stated in SR 3.0.2. the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ... " basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action . whether it is a particular Surveillance or some other remedial action , is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limi.ts when a Surveillance has not been completed within the specified Frequency.._

including the allowance of SR 3.0.2 as applicable. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency. whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2 , and not at the time that the specified Frequency was not met.

These provisions also apply to testing requirements in Section 5.5, "Programs and Manuals." In such cases. when the Section 5.5 Specification states that the provisions of SR 3.0.3 are applicable. the flexibility to defer declaring the testing requirement not met in accordance (continued)

CALLAWAY PLANT B 3.0-13 Revision 10

SR Applicability B 3.0 BASES SR 3.0.3 with SR 3.0.3 when the testing has not been completed with in the testing (continued) interval. including the allowance of SR 3. 0.2 (if invoked by the Section 5.5 Specification) . similarly applies.

+flt&The allowed delay period provides adequate time to complete Surveillances that have been missed. +ffiSThe delay period permits the completion of a Surveillance before complying with Required Actfons or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of unit conditions, adequate planning , availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements . When a Surveillance with a Frequency based not on time intervals , but upon specified unit conditions, operating situations , or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading , or in accordance with 10 CFR 50 ,

Appendix J , as modified by appr0ved exemptions , etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance.

However, since there is not a time interval specified , the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, ln addition to unit conditions , planning , availability of personnel, and the time required to perform the Surveillance. lihis risk impact should be managed through the program in place to implement 10 OFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and (continued)

CALLAWAY PLANT B 3.0-14 Revision 10

Pressurizer Safety Valves B 3.4.10 BASES ACTIONS B.1 and 8.2 (continued)

If the Required Action of A.1 cannot be rnet within the required Completion Time or if two or more pressurizer safety valves are inoperable, the plant must be brought to a MODE in which the requiremeht does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 with any RCS cold leg temperature::; 275°F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The allowed Completion Times are reasonable, based on operating ei<perience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. With any RCS cold leg temperatures at or below 275°F, overpressure protection is provided by the COMS . The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three pressurizer safety valves .

SURVEILLANCE SR 3.4.10.1 REQUIREMENTS SRs are specified in the lnservice +cstln§ PrograF11 .This SR requires testing specified for the pressurizer safety valves in the INSERVICE TESTING PROGRAM. Pressurizer safety valves are to be tested in accordance with the requirements of the ASME Code (Ref. 4 ), which provides the activities and Frequencies necessary to satisfy #lethis SRs .

No additional requirements are specified.

The pressurizer safety valve setpoint is +/-2% for OPERABILITY; however, the valves are reset to +/-1 % during the Surveillance to allow for drift.

REFERENCES 1. ASM~ , Boiler ahd Pressure Vessel Code, Section Il l.

2. FSAR, Chapter 15.

3. WCAP-7769 , Rev. 1, .'June 1972.

4. ASME Code for Operation and Maintenance of Nuclear Power Plants.

CALLAWAY PLANT B 3.4.10-4 Revision 8

Pressurizer PORVs 83.4.11 BASES ACTIONS G.1 and G.2 (continued)

If the Required Actions of Condition F are not met, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion nmes are reasonable , based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 1, 2, 3, and 4 (with any RCS cold leg temperature s 275°F), 5, and 6 (with the reactor vessel head on), automatic PORV OPERABILITY is required . See LCO 3.4.12 for requirements in MODES 4 , 5, and 6.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be opened and closed .

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

The Note modifies this SR by stating that it is not required to be performed with the block valve closed, in accordance with the Required Actions of this LCO. Oper:iing the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.

SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. Operating experience has shown that these valves usually pass the Surveillance when performed at the required INSERVICE TESTING PROGRAM lnseP1ice Testing Program frequency. The Frequency is acceptable from a reliability standpoint.

REFERENCES 1. FSAR Section 15.5.1.

2. Regulatory Guide 1.32, February 1977.

3. FSAR, Section 15.2,

4. ASME Code for Operation and Maintenance of Nuclear Power Plants .

CALLAWAY PLANT B3.4.11-8 Revision 10

COMS B 3.4.12 BASES SURVEILLANCE SR 3.4.12.1 . SR ~.4.12 .2. and SR 3..4.12.3 (continued)

REQUIREMENTS operators (Refs. 10 and 11 ). Verification that each accumulator is isolated is only required when accumulator isolat,ion is required as stated in Note 3 to the LCO.

The sc:1fety injection pumps and one ECCS centrifugal charging pump are rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control.

An alternate method of cold overpressure protection control may be employed using at ileast two independent means to render a pump lncapable of injecting into the RCS such that a single failure or single action will not result in an injection into the RCS . This may be accomplished by placing the pump control switch in pull to lock and closing at least one valve in the discharge flow path , or by closing at least one valve in the discharge flow path and removing power from *the valve operator. or by closing at least one manual valve in the discharge flow path under administrative controls.

The Surveillance Frequency is based on operating experience , equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.4.1 2.4 Each required RH R suction relief valve shall be demonstrated OPERABLE by verifying its RHR suction isolation valves are open and by testing it in accordance with the lnservice Testing ProgramlNSERVICE TESTING PROGRAM. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO .

The RHR suction jsolation valves are verified to be opened . The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

The ASME Code (Ref. 8), test per lnserviee Testing PregramlNSERVICE TESTING PROGRAM verifies OPERABILITY by provlng proper relief valve mechanical motion and by measuring and , if required , adjusting the lift setpoint.

(continued)

CALLAWAY PLANT B 3.4.12-11 Revision 12

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.4 REQUIREMENTS (continued) Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve . The ECCS pumps are required to develop the following differential pressures on recirculation flow: 1) ECCS centrifugal chargif19 pumps 2'. 2400 psid; 2) safety injection pumps ~ 1445 psid; and 3) RHR pumps 2'.165 psid . This verifies both that the measured performance is within an acceptable tolerance of the original pump than or equal to the performance assumed in the plant safety analysis. SRs are specified in the applicable portlohs of the INSERVICE TESTING PROGRAM lnseF¥ice Testing Program , which encompasses the ASME Code. The ASME Code provides the activities and Frequencies necessary to satisfy the requirements .

SR 3.5.2.5 and SR 3.5 .2.6 These Surveillances demonstrate that each automatic,ECCS valve actuates to the required position or:i an actual or simulated SI signal or on an actual or simulated RWST Level Low-Low 1 Automatic Transfer signal coincident with an SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. The containment recirculation sump to RHR pump isolation valves (EJHV8811A/B) automatically open upon receipt of an actual or simulated RWST Level Low-Low-1 Automatic Tr-ansfer signal coincident with an SI signal. In addition to testing that automatic function , SR 3.5.2.5 demonstrates that the RWST to RHR pump suction isolation valves (BNHV8812A/B) are capable of automatic closure after the EJHV8811A/B valves are fully open . The valve interlock functions are depicted in Reference 10. Thls Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrc1tive controls . The Surveillance Frequency is based on operating experience , equipment reliability, -and plant risk and is controlled under the Surveillance Frequency Control Program. The actuation logic is tested as part of ESF Actuation System testing, c1nd equipment performance is monitored as part of the lnservice Testing Pro@fafnlNSERVICE TESTING PROGRAM .

SR 3.5.2.7 The correct position of throttle valves in the flow path is necessary for proper ECCS performance. These valves have mechanical stops to allow proper positioning for restricted flow to a ruptured cold leg , ensuring that (continued}

CALLAWAY PLANT B 3.5.2-12 Revision 14

Containment Isolation Vc:ilves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.4 (continued)

REQUIREMENTS these valves were verified to be in the current position upon locking ,

sealing, or securing .

A Note has been added that allows valves and bllnd flanges located in high radiation areas to be verified closed by use of administrative means.

Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 , for ALARA reasons . Therefore, the probability of misalignment of these containment isolation valves, once they have been verrfred to be in their proper position, is smc:ill.

SR 3.6.3.5 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. An automatic power operated containment isolation valve is a containment isolc:ition valve which is closed by an automatic (i.e., other than operator mam~al) actuation signal and is powered by other than manual actuation (e.g . by a pneumatic, solenoid, or motor operator). The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses. ifhe isolation time and Frequency of this SR are in accordance with the lnservice Testing Pro9raA'1INSERVICE TESTING PROGRAM .

SR 3.6.3.6 Leakage integrity tests with a maximum allowable leakage rate for containment shutdown purge supply and exhaust isolation valves will provide early indication of resilient material seal degradation and will allow opportunity for repair before gross leal<age failures could develop.

This SR is modified by a Note indicating that the SR is only required to be performed when the containment shutdown purge valve blind flanges are installed.

If the blind flange is installed, leakage rate testing of the valve and its associated blind flange must be performed. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. Leakage rate testing must also be performed following each reinstallation of the blind flange .

(continued)

CALLAWAY PLANT B 3.6.3-13 Revision 12

Containment Spray and Cooling Systems B 3. 6.6 BASES SURVEILLANCE SR 3.6.6.1 (continued)

REQUIREMENTS should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication With the operators in the control room . This indlvidual will have a method to rapidly close the system vent flow path if directed, SR 3.6.6.2 Operating each containment cooling train fan unit for~ 15 minutes ensures that all fan units are OPERABLE . It also ensures that abnormal cond itions or degradation of the fan unit can be detected for correctlve action. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6.6.3 Verifying that each containment cooling train ESW cooling flow rate is

~ 2200 gpm provides assurance that the design flow rate assumed in the safety analyses will be achieved (Ref. 3). The Surveillance Frequency is based on operatililg experience , equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.6 .6.4 Verify ing each containment spray pump's developed head at the flow test point is greater than or eq1.1al to the required developed head ensures that spray pump performance has not degraded during the cycle . Flow and differential pressure are normal tests of centrifugal pump performance requ ired by the ASME Code (Ref. 5). Since the containment spray pumps cannot be tested with flow through the spray headers , they are tested on recirculation flow. This test ensures that each pump develops a discharge pressure of greater than or equal to 250 psig. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the lnseP,1ioe Testing ProgramlNSERVICE TESTING PROGRAM .

(continued)

CALLAWAY PLANT B 3.6.6-8 Revision 13

MSSVs B 3 .7 .1 BASES APPLICABLE that the maximum RCS pressure does not exceed 110% of the design SAFETY pressure. All cases analyzed demonstrate that the MSSVs maintain main ANALYSES steam system integrity by limiting the maximum steam pressure to less (continued) than 110% of the steam generator design pressure. In some circumstances it is necessary to limit the primary side heat generation that can be achieved during an AOO by reducing the setpoint of the Power Range Neutron Flux-High reactor trip function. For example, if morce than one MSSV on a single steam generator is inoperable, an uncontrolled RCCA bank withdrawal at power event occurring from a partial power level may result in an increase in reactor power that exceeds the combined steam flow capacity of the turbine and the remaining OPERABLE MSSVs. Thus, for multiple inoperable MSSVs on the same steam generator it rs necessary to prevent this power increase by lowering the Power Range Neutron Flux-High setpoint to an appropriate value. When the Moderator Temperature Coefficient (MTC) is positive ,

the reactor power may increase above the initial value during an RCS heatup event (e.g. , turbine trip). Thus, for any number of inoperable MSSVs it is necessary to reduce the trip setpoint if a positive MTC may exist at partial power conditions.

The MSSVs are assumed to have two active and one passive failure modes. ifhe active failure modes are spurious opening, and failure to reclose once opened. The passive failure mode is failure to open upon demand.

The MSSVs satisfy Criterion 3 of 10 CFR 50.36 (c)(2)(ii).

LCO The accident analysis requires that five MSSVs per steam generator be OPERABLE to provide overpressure protection for design basis transients occurring at 102% RTP. The LCO requires that five MSSVs per steam generator be OPERABLE in compliance with Reference 2 and the OBA .analysis.

The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances to relieve steam generator overpressure, and reseat when pressure has been reduced . The OPERABILITY of the MSSVs is deteFmined by periodic surveillance testing in accordance with the lnser,lce Testing Program lNSERVICE TEST ING PROGRAM.

This LCO provides assurance that the MSSVs wlll perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB or Main Steam System Integrity.

(continued)

CALLAWAY PLANT B3.7.1-2 Revision 8

MSSVs B 3.7 .1 BASES (Continued)

SURVEILLANCE SR 3.7.U REQUIREMENTS This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with the lnservlce Tes.ting ProgramlNSERVICE TESTING PROGRAM .

The ASME Code (Ref. 5) specifies the activities and frequencies necessary to satisfy the requirements. Table 3.7 .1-2 allows a +3%/-1 %

setpoint tolerance for OPERABILITY; however, the valves are reset to

+/- 1% during the Surveillance to allow for drift. The lift settings pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions , the lift setting pressure shall be corrected , if necessary, to ambient conditions of the valve at operating temperature and pressure.

REFERENCES 1. FSAR Section 10.3.2, Main S1eam Supply System - System Description.

2. ASME, Boiler and Pressure Vessel Code , Section Ill ,

Article NC-7000, Class 2 Components.

3. FSAR , Section 15.2, Decrease in Heat Removal by the Secondary System.

4. NRC !~formation Notice94-60, "Potential Overpressurization of the Main Steam System," August 22 , 1994.

5. ASME Code for Operation and Maintenance of Nuclear Power Plants.

6. Westinghouse LetterSCP-99-129, dated July 7, 1999.

7. WCAP-16265-P, "Callaway Replacement Steam Generator Program NSSS Licensing Report," September 2004.

CALLAWAY PLANT B3.7.1-5 Revision 8

MSIVs , MSIVBVs, and MSLPIDIVs B 3.7.2 BASES (Continued)

SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR verifies that the closure time of each MSIV is within the limits of Figure B 3.7.2-1 from each actuator train when tested pursuant to the lnseNice Test ProgramlNSERVICE TESTING PROGRAM . The MSIV isolation time is assumed in the accident and containment analyses.

Figure B 3.7.2-1 is a curve of the M-S IV isolation time as a function of steam generator pressure, since there is no pressure indication available at the MSIVs. The acceptance curve for the MSIV stroke time is conservative enough to account for potential pressure differential between the steam generator pressure indication and pressure at the MS IVs.

Meeting the MSIV isolation times ih Figu re B 3.7.2-1 ensures that the evaluation performed in Reference 7 remains valid. This Surveillahce is normally performed upon returning the unit to operation following a refueling outage. The MS IVs should not be tested at power, since even a part stroke exercise increases the risk of a valve closure when the unit is generating power.

The Frequency is in accordance with the lnservice +est:ing ProgramlNSERVICE TESTING PROGRAM .

SR 3.7.2.2 This SR verifies that each MSIV, each MSIVBV, and each MSLPDIV is capable ofclosure on an actual or simulated actuation signal. The manual fast close handswitch in the Control Room 1provides an acceptable actuation signal. For the MS IVs each actuation train must be tested separately. This Surveillance ls normally performed upon returning the unit to operation following a refueling outage in conjunction with SR 3.7.2.1. However, it is acceptable to perform this surveillance iodividually. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program .

SR 3.7.2.3 This SR verifies that the closure time of each MSIVBV and MSLPDIV is

~ 15 seconds when tested pursuant to the lnser.*ice Tosting ProgramlNSERVICE TESTING PROGRAM . This Is consistent with the assumptions used in the accident and containment analyses.

For the MSIVBVs and MSLPDIVs, this Surveillance is performed routinely during plant operations (or as required for post-maintenance testing) , but it (continued)

CALLAWAY PLANT B 3.7.2-12 Revision 11

MSIVs , MSIVBVs , and MSLPDIVs B 3.7.2 BASES SURVEILLANCE SR 3.7.2.3 (continued)

REQUIREMENTS may also be required to be performed upon returning the unit to operation following a refueling outage.

The Frequency for this SR is in accordance with the lnservice Testing ProgramlNSERVICE TESTING PROGRAM.

REFERENCES 1. FSAR, Section 10.3, Main Steam Supply System .

2. FSAR , Section 6.2, Containment Systems.

3. FSAR , Section 15.1 .5, Steam System Piping Failure.

4. 10 CFR 100.11 .

5. FSAR 6.2 .1.4.3.3, Containment Pressure - Temperature Results .

6. Amendment 172 to Facility Operating License No. NPF-30 ,

(NRC Safety Evaluation included) Callaway Unit 1, dated June 16, 2006 .

7. Westinghouse Letter, SCP-07-26 , dated March 6, 2007 .

CALLAWAY PLANT B 3.7.2-13 Revision 11

MFIVs, MFRVs, and MFRVBVs B 3.7.3 BASES ACTIONS C.1 and C.2 (continued) ensure that the assumptions in the safety analysis remain valid. The 7 day Completion Time is reasonable, based on engineering judgment, in view of valve status indications available in the control room , and other administrative controls, to ensure that these valves are closed or fsolated .

If the MFRVBVs are closed and de-activated , or closed and isolated by a closed manual valve , or isolated by two closed manual valves, this LCO does not apply as discussed in the Applicability section of these Bases.

Two inoperable valves in the same flow path is treated the same as loss of the isolation capability of this flow path. For each feedwater line there are two flow paths, defined as flow through the MFRV/MFIV and flow through the MFRVBV/MFIV. Because the MFIV, MFRV, and MFRVBV are of different designs, a common mode failure of the valves in the same flow path is not likely. However, under these conditions, affected valves in each flow path must be restored to OPERABLE status, or the affected flow path isolated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This action returns the system to the condition where at least one valve in each flow path is performing the required safety function . The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is reasonable , based on operating experience, to complete the actions required to close the MFIV or MFRV and MFRVBV, or otherwise isolate the affected flow path.

E.1 and E.2 If the MFIV(s) and MFRV(s) and MFRVBV(s) cannot be restored to OPERABLE status, or closed, within the associated Completion Time, the unit must be placed fn a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable , based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURV:EILLANCE SR 3.7.3 .1 REQUIREMENTS This SR verifies that the closure time of ec1ch MFRV and MFRVHV is s 15 seconds when tested pursuant to the lnservice.1es.ting ProgramlNSERVICE TE STING PROGRAM. The MFRV and MFRVBV closure ttme is assumed in the accident and (continued)

CALLAWAY PLANT B 3.7.3-10 Revision 1 *1

MFIVs, MFRVs, and MFRVBVs B 3.7.3 BASES SURVEILLANCE SR 3.7.3.1 (continued)

REQUIREMENTS containment analyses. For the MF RVs, performance of this surveillance involves simultaneously stroking the MFRVs closed Using nitrogen at or below the required accumulator pressure (400 psig) and verifying that the valves close within the required limit. (Ref. 9)

For the MFRVs this Surveillance is normally performed upon returning the unit to operation following a refueling, outage, or it may be performed as required for post-maintenance testing under appropriate conditions during applicable MODES. The MFRVs should normally not be tested at power since even a partial stroke exercise increases the risk of a valve closure with the unit generating power. However, when the plant is operating using 1he MFRVBVs (steam generator level maintained solely by flow through the MFRVBVs) , the surveillance for the MFRVs may be performed for post-maintenance testing during such conditions without increasing plant risk.

For the MFRVBVs, this Surveillance is performed routinely during plant operation (or as required for post-maintenance testing), but it may also be required to be performed upon returning the unit to operation following a refueling outage.

For verifying valve closure time when returning the unit to operation followlng a refueling outage, the SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with *t hose under which the acceptance criterion was _generated . Test conditions are with the unit at nominal operating temperature and pressure , as dicussed in Reference 8.

Per Reference 4, if it is necessary to adjust stem packing to sto,p packing leakage and if a required stroke test is not practical in the current plant MODE, it should be shown by analysis that the packing adjustment is within torque limits specified by the manufacturer for the existing configuration of packing , and that the performance parameters of the valve are not ad~er,sely affected. A confirmatory test must be performed at the first available opportunity when plant conditions allow testing. Packing adjustments beyond the manufacturer's limits may not be performed without (1) an engineering analysis and (2) input from the manufacturer, unless tests can be performed after adjustments.

The Frequency for this SR is in accordance with the lnserviee Testing Pro9ramlNSERVICE TESTING PROGRAM.

(continued)

CALLAWAY PLANT B 3.7.3-11 Revision 11

MFIVs, MFRVs , and MFRVBVs B 3.7.3 BASES SURVEILLANCE SR 3.7.3.2 REQUIREMENTS (continued) This SR verifies that each MFIV, MFRV, and MFRVBV is capable of closure on an actual or simulated actuation signal. For the MFIVs tl:!e manual fast close handswitch in the Control Room provides an acceptable actL:Jation signal. Each MFIV actuation train must be tested separately. For the MFRVs and the MFRVBVs, actuation of solenoids locally at the MFRVs and MFRVBVs constitutes an acceptable simulated actuation signal.

This Surveillance is normally performed for the MFIVs and MFRVs upon returning the unit to operation following a refueling outage in conjunction with SR 3.7.3:1. The SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR for the MFRVs and MFRVBVs. This allows a delay of testing until MODE 3 to establish conditions consistent with those necessary to perform SR 3.7.3.1 and SR 3.7.3 .2 concurrently for the MFRVs and for the MFRVBVs, as necessary.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.7.3.3 This SR verifies that the closure time of each MFIV is within the limits of Figure 8 3.7 .3-1 from each actuation train when tested pursuant to the lnservice Testing ProgramlNSERVICE TESTING PROGRAM . The MFIV closure time is assumed in the accident and containment analyses. Figure B 3.7.3-1 is a curve of the MFIV isolation time limit as a function of steam generator steam pressure, since there is no pressure indication available at the MFIVs. The acceptance curve for fhe MFIV stroke time conservatively accounts for the potential pressure differential between the steam generator pressure indication and the pressure at the MFIVs.

Meeting the MFIV isolation times in Figure B 3.7.3-1 ensures that the evaluations performed in Reference 2 and Reference 7 remain valid . This Surveillance is normally performed upon returning the unit to operation following a refueling outage. These valves should not be tested at power since even a partial stroke exercise increases the rl sk of a valve closure with the unit generating power.

The F.requency for this SR is in accordance with the IAservice TostiAg PrograrnlNSERVICE TESTING PROGRAM .

(continue )

CAL.LAWAY PLANT B 3.7.3-12 Revision 11

ASDs B 3.7.4 BASES ACTIONS ~ (continued) required ASD lines to OPERABLE stat1:1s. Since the manual isolation valve can be closed to isolate an ASD , some repairs may be possible with the unit at power. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is reasonable to repair inoperable ASD lines , based on the availability of the Condenser Steam Dump System and MSSVs, and the low probability of an event occurring during this period that would require the ASD lines.

D.1 and D.2 Requiring a 30 day limit for restoring an ASD valve to OPERABLE status from inoperable, d1:1e to excessive seat leakage from the valve , provides assurance that the required number of ASDs will be available for plant cooldown. This action limits the period in wh1ch a manual isolation valve is closed due to excessive seat leakage of the ASD and minimizes the delay associated with manually opening a closed manual isolation valve (due to excessive seat leakage of the ASD) .

E.1 and E.2 If the required ASD line(s) cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within

12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditlons in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a controlled cooldown of the RCS , th!: ASDs must be able to be opened remotely and throttled through their full range. This SR ensures that the ASDs are tested through a full control cycle as described in the lnservice Test PrograFT1INSERVICE TESTING PROGRAM .

The conditions that best verify the operability of the ASDs is with the manual isolation valve open and nominal steam line operating pressure and temperature. The ASDs are designed such that steam line pressure acts on top of the valve pl1..1g. When the valve is required to move to the open position the actuator must act against steam line pressure. For tl1is surveillance requirement to best verify the operational readiness of the (continued)

CALLAWAY PLANT B 3.7.4-4 Revision 11

ASDs B 3 .7.4 BASES SURVEILLANCE SR 3.7.4.1 (continued)

REQUIREMENTS ASDs, it should be performed at nominal SG operating temperature and pressure , which is in the upper portion of MODE 3 (Ref. 5) .

Use of an ASD during a unit cooldown may satisfy this requirement.

Operating experience has shown that these components usually pass the Surveillance when performed at the required IRservice TostiRg Program lNSERVICE TESTING PROGRAM Frequency. The Frequency is acceptable from a reliability standpoint.

This Surveillance Requirement is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which acceptance criterion was generated (Ref. 5) .

SR 3.7.4.2 The function of the manual isolation valve is to Isolate a failed open or leaking ASD . Cycling the manual isolation valve both closed' and open demonstrates its capability to perform this function . Performance of inservice testing or use of the manual isolation valve during unit cooldown may satisfy this requirement. Operating experience has shown that these components usually pass the Surveillance when performed in accordance with the lnservice Testing PrograrnlNSERVICE TESTING PROGRAM Frequency. The Frequency is acceptable from a reliability standpoint.

REFERENCES 1. FSAR, Section 10.3, Main Steam Supply System.

2. FSAR, Chapter 15.2 , Decrease in Heat Removal by the Secondary System.

3. FSAR , Section 15.6.3, Steam Generator Tube Failure .

4. FSAR , Section 9.3.1, Compressed Air System.

5. Operating License Amendments 45 and 59.

6. Operating License Amendment 131.

CALLAWAY PLANT B 3.7.4-5 Revision 11

AFW System B 3.7.5 BASES SURVEILLANCE SR._3.7.5.1 (continued)

REQUIREMENT such as check valves and relref valves. Additionally, vent and drain valves are not within the scope of this SR .

This SR is modified by a Note indicating that the SR is not required to be performed for the AFW flow control valves until the AFW system is placed in automatic control or when Thermal Power is above 10% RTP.

In order for the TDAFP and MDAFPs to be OPERABLE during normal power operation while the AFW system is in automatic control or above 10% RTP, the applicable discharge flow control valves (ALHV0005 and ALHV0007 for PAL01 B; ALHV0009 and ALHV001 1 for PAL01 A; ALHV0006 , ALHV0008 , ALHV0010 , and ALHV0012 for PAL02) shall be OPERABLE and in the fully open positior.i. The TDAFP and MDAFPs remain OPERABLE with the applicable discharge flow control valves throttled to maintain steam generator levels during plant heatup, cooldown 1 or if started due to an Auxiliary Feedwater Actuation Signal (AFAS) or manually started in anticipation of an AFAS.

The Surveillance Frequency is based on operating experience , equipment reliab11ity, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.7.5.2 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME Code (!Ref. 2). Because it is undesirable to introduce cold AFW into the steam generators while they are operating ,

th is testing is performed on redrculation flow. Such lnservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

Performance of inservice testing discussed in the ASME Code (Ref. 2)

(only required at 3 month intervals) satisfies this requ iremen t. The test Frequency in accordance with the lnservice Testing PrograrnlNSERVJCE TESTING PROGRAM results in testing each pump once every 3 months, as required by Reference 2.

The required differential pressure for the AFW pumps when tested in accordance with the lnser.<<ice Testing ProgramlNSERVICE TESTING PROGRAM is:

(continued)

CALLAWAY PLANT B 3.7 .5-9 Revision 13

Secondary System Isolation Valves (SSIVs)

B 3.7.19 BASES SURVEILLANCE SR 3.7.19.1 REQUIREMENTS This SR verifies that the isolation time of each automatic SSIV ls within limits when tested pursuant to the lnservice Testing (1ST) Program . The specific limits are documented in the 1ST Program. The SSIV isolation times are less than or equal to those assumed in the accident and containment analyses. This surveillance does not include verifying a closure time for the steam generator chemical addition injection isolation valves since it is. not applicable. to normally locked closed manual valves .

H1ese valves are not included in the 1ST program because the valves are passive (not required to actuate to their safety position) and they contain a locking device and a check valve in their flow path.

For the SSIVs, performance of this surveillance may be done during plant operation (or as required for post-maintenance testing) , but it may also be required to be performed upon returning the unit to operation following, a refueling outage.

The Frequency for this SR is 1n accordance with the IRsePJioc Testing Program lNSERVICE TESTING PROGRAM .

SR 3.7.19.2 This SR verifies that each automatic SSIV in the flow path is capable of closure on an actual or simulated actuation sfgnal. This surveillance is routinely performed during plant operation, but may also be performed upon returning the unit to operation following a refueling outage.

The Surveillance Frequency ts based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

(continued)

CALLAWAY PLANT B 3.7.19-6 Revision 11

Containment Penetrations B 3.9.4 BASES SURVEILLANCE SR 3 .9.4.3 REQUIREMENTS (continued) Th is Surveillance demonstrates that each containment purge isolation valve actuates to its isolation position on manual initiation. The Surveillance Frequency is based on operating experience, eqwipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance wi*th the INSERVIC E TESTING PROGRAMlnseRricc Tosting Pro§faffl requirements. These Surveillances will ensure that the valves are capable of being manuafly closed after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

REFERENCES 1. Amendment 114 to Facility Operating License No. NPF-30 ,

Callaway Unit 1, dated July 15, 1996 .

2. FSAR , Section 15.7.4.

3. NUREG-0800 , Section 15.7.4, Rev. 1, July 1981 .

4. Amendment 138 to Facrlity Operating License No, NPF-30, Callaway Unit 1, dated September 26, 2000.

CALLAWAY PLANT B 3.9.4-8 Revision 1 *1

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ULNRC-06478, Specifications to Adopt TSTF-545 Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, (LDCN 18-0021)

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ULNRC-06478, Specifications to Adopt TSTF-545 Revision 3, TS Inservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing, (LDCN 18-0021)

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