Text

Wolf Creek Transmittal of Revision 21 to Updated Safety Analysis Report
	ML081790167
Person / Time
Site:	Wolf Creek
Issue date:	03/04/2008
From:	Hedges S E; Wolf Creek
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	WM 08-0003
	Download: ML081790167 (55)
	v • d • e

WLF CREEK'NUCLEAR OPERATING CORPORATION Stephen E. Hedges Vice President Oversight March 4, 2008 WM 08-0003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Revision 21 of the Wolf Creek Updated Safety Analysis Report Gentlemen:

Pursuant to the updating requirements of 10 CFR 50.71(e), Wolf Creek Nuclear Operating Corporation (WCNOC) is providing its Updated Safety Analysis Report (USAR), Revision 21.This submittal satisfies the Final Safety Analysis Report (FSAR) updating requirements of the aforementioned regulation.

Attachment I to this letter provides information relative to changes in regulatory commitments.

This information is provided in accordance with the guidance of Nuclear Energy Institute (NEI)99-04, "Guidelines for Managing NRC Commitments," Revision 0, July 1999.Attachment II to this letter describes specific technical changes that have been processed since issuance of the Updated Safety Analysis Report (USAR), Revision 20. In addition to these technical changes, several editorial changes have been made and are included in Revision 21.Attachment III to this letter provides a discussion of changes made in Revisions 30 through 33 of the Technical Requirements Manual (TRM).Enclosure I to this letter provides the CD-ROM submittal of the Wolf Creek Updated Safety Analysis Report (USAR), Revision 21. This submittal satisfies the Final Safety Analysis Report updating requirement of 10 CFR 50.71 (e)(4).Enclosure II to this letter provides a CD-ROM containing the station-controlled drawings that are considered incorporated by reference into the USAR. Per the guidance of Nuclear Energy Institute (NEI) document NEI 98-03, Revision 1, "Guidelines for Updating FSARs," the USAR figures that are identical to controlled drawings were relocated from the USAR in Revision 17.Enclosure II is considered sensitive unclassified information and therefore warrants withholding under 10 CFR 2.390.P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 A oc)An Equal Opportunity Employer M/F/HCNET AJ WM 08-0003 Page 2 of 3 Enclosure III to this letter provides a CD-ROM containing the Fire Hazards Analysis and the Quality Program Manual, both of which are incorporated by reference into the USAR. Per the guidance of Nuclear Energy Institute (NEI) document NEI 98-03, Revision 1, "Guidelines for Updating FSARs", Chapter 17.2, Quality Assurance During the Operation Phase, was relocated from the USAR into the Quality Program Manual in Revision 21.Enclosure IV to this letter provides those changes made to the Wolf Creek Generating Station (WCGS), Unit 1 Technical Requirements Manual (Revisions 30 through 33) and includes a List of Effective Pages. The WCGS TRM is incorporated by reference into the USAR. " WCNOC has historically submitted updates to the USAR on March, 11 of each year to coincide with the date of issuance of the WCGS operating license and to comply with the requirements of 10 CFR 50.71(e)(4).

WCNOC considers that submittals made prior to or on March 11 satisfy the requirements of 10 CFR 50.71(e)(4).

There are no commitments contained in this letter.If you have any questions concerning this matter, please contact me at (620) 364-4190, or Mr.Richard Flannigan, Manager Regulatory Affairs at (620) 364-4117., Sincerely, S ?hen E. Hedges SEH/rlt Attachment I -Commitment Changes Attachment II -USAR Change Requests Attachment III -Revisions to the Technical Requirements Manual (TRM)Enclosure I -Updated Safety Analysis Report Enclosure II -Updated Safety Analysis Report Controlled Drawings Enclosure III -Updated Safety Analysis Report Fire Hazards Analysis and Quality Program Manual Enclosure IV -TRM Replacement Pages cc: E. E. Collins (NRC), w/a, w/e V. G. Gaddy (NRC, w/a, w/e B. K. Singal (NRC), w/a, w/e Senior Resident Inspector (NRC), w/a, w/e WM 08-0003 Page 3 of 3 STATE OF KANSAS )SS COUNTY OF COFFEY )Stephen E. Hedges, of lawful age, being first duly sworn upon oath says that he is Vice President Oversight of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.Byn Stephen g Vice Pre d: dt Oversight SUBSCRIBED and sworn to before me this q/day of M ,2008.,ThF* -Z RHONDA L. TIEMEYER ih onk.7 b 2 1 6/'SEAL. MY COMMISSION EXPIRES Notary Public -January 11, 2010 Expiration Date 6 9/2flhl}2

//, ,965/6 Attachment I to WM 08-0003 Page 1 of 2 COMMITMENT CHANGES Commitment No.: 1997-113 Commitment

==

Description:==

KG&E intends to continue the Quality First program. As in the past, all Quality First files will be retained and will be fully available for review by NRC investigators.

In addition, KG&E will keep the commission aware of concerns the Quality First program receives on "wrong-doing" issues (such as intimidation/harassment, drug or alcohol abuse) where the initial review indicates that the concern may have some merit.Change to Commitment:

This commitment was archived in accordance with NEI 99-04,"Regulatory Commitment Management Guidance." Reason for Change: At the time the commitment was made the Quality First program was a process used by KG&E that the NRC had indicated they thought was a good program. Since that time the program and guidance in this area have matured and now have other industry standards that, when met, will achieve the objectives of the original program. Programs in accordance with these guidelines will meet the intent of the commitment with appropriate access to information by the NRC.The commitment is actually in three portions:

record retention; availability to NRC review; and keeping the Commission aware of concerns brought to the employee concerns program.Regarding the records retention portion of the commitment:

1. Retaining all Quality First (Q-1) or ECP files indefinitely is not necessary based on regulatory requirements.

No regulatory requirements apply to the retention of these confidential files.2. NEI 97-05, "Nuclear Power Plant Personnel

-Employee Concerns Program (ECP)-Process Tools in a Safety Conscious Work Environment", Revision 2, December 2003, provides guidelines on record retention.

Retention periods should be the minimum necessary until the concern is resolved and then as considered appropriate until all potential litigation on which the information has a bearing has been concluded.

The retention period for each type of records should be established and records destroyed as called for in the policy.> ECP practice at many licensees calls for records retention of between. five and seven years.3. NRC Directive 8.8, "Management of Allegations", Section (OAC) (B) (4) (f), page 17 states: "Closed allegation files should be held for 2 years, then retired to the NRC Archives.

Files may be destroyed 10 years after cases are closed. The OAC should contact the Records Management Branch, Office of the Chief Information Officer, for transfer of files to the archives facility."

Attachment I to WM 08-0003 Page 2 of 2 Regarding the second portion of the commitment, records availability to NRC review, the NRC is currently able to review any records as appropriate to their jurisdictional authority.

Deleting this commitment will not change WCNOC's commitment to continue to work cooperatively with the NRC on any issue, including an employee concern.The third portion of the commitment, keeping the Commission aware of concerns brought to the employee concerns program, has only been done, as applicable, when a concern may have been relevant to a NRC requirement.

Employee concerns have been, and will continue to be, kept confidential by the company. Deleting this commitment will not decrease WCNOC's dedication to thoroughly investigate these employee concerns.Review of the commitment in accordance with NEI 99-04 concluded that it was acceptable to change the commitment as described.

However because the issue of addressing employee concerns is clearly an area that the NRC has an interest in the commitment change is included in this report.

Attachment II to WM 08-0003 Page 1 of 6 USAR CHANGE REQUEST DESCRIPTION 07-003 REVISE THE USAR TO DELETE STATEMENT, "THERE IS NO DRAINAGE BETWEEN MAIN STEAM LINE COMPARTMENTS" FROM THE LAST PARAGRAPH ON PAGE 3B-9.Page: 3B-9 07-005 REVISE USAR SECTION 15.7.4 REGARDING THE RELEASE PATHWAY OF RADIOLOGICAL CONSEQUENCES RESULTING FROM A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT.

THIS CHANGE REFLECTS THE CHANGES MADE BY TECHNICAL SPECIFICATION AMENDMENT 146.Page: 15.7-15 Page: 15.7-14 Page: 15.7-11 Page: 15.7-10 07-006 REVISE THE USAR TO INDICATE THAT THE SPECIFIC ACTIVITY USED FOR ACCIDENT ANALYSIS RELEASES IS BASED ON OPERATING WITH 1% FUEL DEFECTS WHICH RESULTS IN A REACTOR COOLANT SYSTEM ACTIVITY LIMIT MORE LIMITING THAN TS 3.4.16. CHANGE DUE TO LICENSE AMENDMENT NO. 170. REVISION IS TO SECTION 11.1.3.Page: 11.1-2 Page: 11.1-1 Table: 11.1-5 07-007 REVISE THE USAR TO CHANGE 3A-15 TO REFLECT THAT THE ALTERNATIVE PROGRAM TO ANSI N18.7-1976 DOES NOT APPLY TO THE EMERGENCY, ALARM RESPONSE, OFF-NORMAL, SEVERE ACCIDENT MANAGEMENT GUIDELINES, AND EMERGENCY PLAN IMPLEMENTING PROCEDURES.

THESE PROCEDURES WILL CONTINUE TO BE REVIEWED EVERY TWO YEARS.Page: 3A-15 07-008 REVISE THE USAR TO REFLECT FUEL CONFIGURATION CHANGES FOR REGION 18 (CYCLE 16 FRESH) FUEL ASSEMBLIES CONSISTENT WITH CONFIGURATION CHANGE PACKAGE (CCP) 011840, "CYCLE 16 RELOAD DESIGN CHANGES." Page: 4.2-16 Page: 4.2-15 Page: 4.2-13 Page: 4.1-3 Page: 4.2-20 Page: 4.2-14 Page: 4.2-12 Page: 4.2-11 Page: 4.1-4 Page: 4.1-2 Table: 4.3-1 Sheet: 1 Table: 4.1-1 Sheet: 4 Table: 4.1-1 Sheet: 3 Figure: 4.2-2E Attachment II to WM 08-0003 Page 2 of 6 USAR CHANGE REQUEST DESCRIPTION 07-009 REVISE THE USAR TO INSERT "BEING" INTO SECTION 10.4.7.2.3."BEING" SHOULD HAVE BEEN INCLUDED WITH THE CHANGES MADE IN USAR CHANGE REQUEST 99-051, BUT THE WORD WAS OMITTED.Page: 10.4-29 07-010 REVISE THE USAR TO REMOVE REFERENCES TO THE SUPERINTENDENT OPERATIONS SUPPORT (TRAINING)

POSITION BECAUSE THE POSITION HAS BEEN DISSOLVED.

REMOVING A SHIFT MANAGER DUE TO HIS RETIREMENT.

Page: 13.1-9 Page: 13.1-19 Page: 13.1-17 Figure: 13.1-2b 07-012 REVISE THE USAR TO CHANGE THE CONTAINMENT RECIRCULATION SUMP LEVEL TRANSMITTERS TO DIFFERENTIAL PRESSURE TRANSMITTERS, CHANGE THE RANGE OF THE CONTAINMENT WATER LEVEL INDICATION, AND ADD THE ACCURACY OF THE RESIDUAL HEAT REMOVAL SUMP LEVEL INDICATOR DURING ACCIDENT CONDITIONS.

Page: 18.2-58 Table: 7A-3 Sheet: 6.2 Table: 3.11(B)-10 Table: 3.11(B)-3Sheet:

34 Table: 3.11(B)-3 Sheet: 33 07-013 REVISE THE USAR TO INCORPORATE EDITORIAL CORRECTIONS.

Page: 3.0-xxi Page: 6.2-77 Page: 9.0-x Page: 9.0-vii Page: 7.0-x Page: 6.0-x Page: 6.0-v Page: 6.0-i Page: 5.0-vi Page: 9.0-xiii Page: 3.0-xxiv Page: 11.0-ii Page: 3.0-xix Page: 3.0-xvi Page: 3.0-xv Page: 3.0-xiv Page: 3.0-xxi Page: 3.0-xix Page: 11.0-vi Page: 10.4-40 Page: 3.0-xii Page: 3.0-xi Page: 3.0-viii Page: 3.0-iii Page: 6.2-81 Page: 5.0-ii Page: 6.2-72 Page: 18.0-i Page: 9.2-25 Page: 3A-61 Page: 3A-5 Page: 3A-4 Page: 9.1-47 to 70 Page: 13.2-21 Page: 1.2-6 Page: 9.0-xi Page: 6.2-75 Page: 11.0-vii Page: 18.0-iv Page: 18.0-ii Page: 15.0-xii Page: 15.0-xi Page: 15.0-vii Page: 13.0-ii Page: 13.0-iii Page: 12.0-i Page: 1.0-v Page: 4.0-i Table: 1.7-3 Sheet: 5 Table: 1.7-2 Table: 1.7-2 Sheet: 2 Table: 1.7-2 Sheet: 6 Table: 1.7-2 Sheet: 7 Table: 1.7-3 Table: 1.6-3 Sheet: 3 Table: 1.7-3 Sheet: 4 Table: 7.4-6 Table: 3.5-3 Table: 6.2.1-56 Table: 13.1-1 Sheet: 3 Attachment II to WM 08-0003 Page 3 of 6 USAR CHANGE REQUEST DESCRIPTION Table: 13.1-1 Sheet: 4 Table: 1.7-3 Sheet: 2 Figure: 2.4-1 Figure: 15.6-38 07-014 REVISE THE USAR TO CORRECT THE QUALIFICATION AND TRAINING REQUIREMENTS FOR THE RADIATION PROTECTION MANAGER IN TABLE 13.1-1. THE WORD "ADDITIONAL" IS BEING ADDED TO FIVE YEARS OF PROFESSIONAL EXPERIENCE.

Table: 13.1-1 Sheet: 7 07-015 REVISE THE USAR TO CHANGE THE AVAILABLE NET PRESSURE SUCTION HEAD FOR THE RESIDUAL HEAT REMOVAL AND CONTAINMENT SPRAY PUMPS AT THE BEGINNING OF THE RECIRCULATION MODE AS A RESULT OF THE MEASURED HEAD LOSS ACROSS THE CLEAN CONTAINMENT RECIRCULATION SUMP STRAINERS.

Table: 6.2.2-7 Table: 6.3-1 Sheet: 2 07-016 REVISE THE USAR TO ABANDON IN PLACE TWO INCORE THERMOCOUPLES (BBT/CO011 AND BBTIC0035).

ALSO CHANGE THE TOTAL NUMBER OF THERMOCOUPLES FROM 50 TO 47 DUE TO PREVIOUSLY REMOVED THERMOCOUPLES.

Page: 18.2-85 Page: 18.2-72 Table: 3.11(B)-3 Sheet: 19 Table: 3.11(B)-3 Sheet: 17 Table: 7A-3 Sheet: 1.3 07-018 REVISE THE USAR TO INCORPORATE REFERENCE TO NRC LETTER 89-00605, TMI ACTION PLAN ITEM II.K.3.5, REACTOR COOLANT PUMP TRIP, ON PAGE 18.2-102.Page: 18.2-102 07-019 REVISE THE USAR TO CHANGE PAGES 13.1-5, 13.1-6 AND 18.1-16.PAGE 13.1-5, SECTION 13.1.1.2.1, CHANGE "THROUGH SITE MANAGEMENT" TO "THROUGH VICE PRESIDENT ENGINEERING AND VICE PRESIDENT OVERSIGHT." PAGE 13.1-5, SECTION 13.1.1.2.3, ADD ON ISEG FUNCTION AS A RESPONSIBILITY.

PAGE 13.1-6, SECTION 13.1.1.2.5, ADD ISEG FUNCTIONS AS A RESPONSIBILITY.

PAGE 18.1-16, SECTION 18.1.7.2, CHANGE "SITE MANAGEMENT IS" TO "THE VICE PRESIDENT ENGINEERING AND VICE PRESIDENT ARE..." Page: 18.1-16 Page: 13.1-6 Page: 13.1-5 Attachment II to WM 08-0003 Page 4 of 6 USAR CHANGE REQUEST DESCRIPTION 07-020 REVISE THE USAR TO REFLECT ORGANIZATIONAL CHANGES. THE MANAGER CHEMISTRY/RADIATION PROTECTION POSITION WAS VACATED BY THE INDIVIDUAL TAKING THE INPO ASSIGNMENT.

THE CURRENT MANAGER OPERATIONS RELOCATED TO THE MANAGER CHEMISTRY/RADIATION PROTECTION POSITION.

THE CURRENT SUPERINTENDENT OPERATIONS RELOCATED TO THE MANAGER OPERATIONS POSITION.

ONE OF THE CURRENT SHIFT MANAGER'S RELOCATED TO THE SUPERINTENDENT OPERATIONS POSITION.Page: 13.1-19 Page: 13.1-16 Page: 13.1-20 Page: 13.1-18 Page: 13.1-17 07-021 REVISE THE USAR TO REFLECT THE FOLLOWING ORGANIZATIONAL CHANGE:THE CURRENT VICE PRESIDENT OPERATIONS/PLANT MANAGER IS NOW THE VICE PRESIDENT OVERSIGHT AND THE CURRENT VICE PRESIDENT OVERSIGHT IS NOW THE VICE PRESIDENT OPERATIONS/PLANT MANAGER.Page: 13.1-7 07-023 REVISE THE USAR TO ADD BRAD NORTON AS THE MANAGER INTEGRATED PLANT SCHEDULING.

Page: 13.1-21 07-024 REVISE THE USAR TO CHANGE THE STEAM GENERATOR FEEDWATER HYDRAZINE LOW SETPOINT IN TABLE 9.3-5 FROM 100 PPB TO 60 PPB TO BE IN AGREEMENT WITH THE CURRENT CHEMISTRY SPECIFICATION AS GIVEN IN AP 02-003, CHEMISTRY SPECIFICATION MANUAL.Table: 9.3-5 07-025 REVISE THE USAR TO REFLECT THE 2007 INSTALLATION OF THE UPGRADED PLANT PROCESS COMPUTER.Page: 8.3-36 Page: 8.2-6 Page: 7.7-35 Page: 5.2-41 Table: 7A-3 Table: 7A-3 Table: 7A-3 Table: 7.5-2 Table: 7.5-1 Page: 12.3-30 Page: 8.1-15 Page: 7.5-8 Page: 2.3-40 Sheet: 17.5(CONT)

Sheet: 12.3 Sheet: 2.8 Page: 9.4-53 Page: 7.7-15 Page: 7.3-10 Page: 1.2-15 Table: 7A-3 Table: 7A-3 Table: 7A-3 Page: 9.4-48 Page: 7.7-11 Page: 5.2-42 Sheet: 17.5 Sheet: 4.3 Sheet: 1.1 Attachment II to WM 08-0003 Page 5 of 6 USAR CHANGE REQUEST DESCRIPTION 07-026 REVISE THE USAR TO CHANGE TABLE 12.5-1 TO ELIMINATE INSTRUMENTATION AND TO UPDATE AVAILABLE QUANTITIES OF INSTRUMENTS AND EQUIPMENT TO REFLECT CURRENT NUMBERS AND PRACTICES.

Table: 12.5-2 Sheet: 2 Table: 12.5-2 Table: 12.5-1 07-027 REVISE THE USAR TO UPDATE TABLE 3.11(B)-3, SHEET 35, AND TABLE 9.5.5-1, SHEET 2, TO REFLECT EMERGENCY DIESEL INTERCOOLER RE-TUBING OF HEAT EXCHANGERS EKJ03A AND EKJ03B.Table: 9.5.5-1 Sheet: 3 Table: 7.4-6 Sheet: 23 Table: 9.5.5-1 Sheet: 2 Table: 3.11(B)-3 Sheet: 35 07-028 REVISE THE USAR TO REFLECT THE INSTALLATION OF NEW WASTE GAS ANALYZER RACKS (HA162). THE RACKS ANALYZE THE OXYGEN AND HYDROGEN CONTENT OF THE GAS ENTERING AND EXITING HYDROGEN RECOMBINER SHA01B.Page: 11.3-10 Figure: 11.3-4 07-029 REVISE THE USAR TO ALLOW THE USE OF CODE CASES N-411, N-318-3, N-391, N-392, AND 1606-1. THE CODE CASES ARE APPLICABLE TO WOLF CREEK AND ARE ALLOWED UNDER REG GUIDE 1.84, REVISION 31.Page: 5.2-45 Page: 5.2-2 Table: 3.9(B)-7 07-030 REVISE THE USAR TO REMOVE INCORRECT SENTENCES REGARDING FIGURE 8.3-6. ONE SENTENCE IS UNASSOCIATED WITH ITS REFERRAL IN THE USAR AND ONE SENTENCE REFERS TO FIGURE 8.3-6, SHEET I THAT WAS INCORPORATED BY REFERENCE IN REVISION 17.Page: 8.1-3 Page: 8.3-38 Page: 8.3-22 Table: 8.3-4 Attachment II to WM 08-0003 Page 6 of 6 USAR CHANGE REQUEST DESCRIPTION 07-031 REVISE THE USAR TO REFLECT TOM DOUGAN REPLACING JOHN FLETCHER AS THE SUPERVISOR QUALITY ASSURANCE.

Page: 13.1-16 07-032 REVISE THE USAR TO ADD PRESTON LAWSON AS A NEW SHIFT MANAGER IN SECTION 13.1.Page: 13.1-19 07-033 REVISE THE USAR TO UPDATE TEXT IN SECTION 15.6.3.1.2 TO REFLECT ASSUMPTIONS USED IN THE STEAM GENERATOR TUBE RUPTURE ACCIDENT ANALYSIS DOCUMENTED IN CALCULATION AN-99-1025, REV. 1.Page: 15.6-11 Page: 15.6-10 07-034 REVISE THE USAR TO REFLECT THE INSTALLATION OF WELD OVERLAYS ON THE PRESSURIZER (TBB03) SPRAY, RELIEF, SAFETY AND SURGE NOZZLES.Page: 5.2-45 Table: 5.2-2 08-001 Page: 5.4-60 Page: 5.4-49 Sheet: 2 REVISE THE USAR TO REMOVE CHAPTER 17.2 FROM THE USAR AND TO INCORPORATE IT BY REFERENCE.

INCLUDE A CROSS-REFERENCE TO SHOW WHERE THE SECTIONS OF 17.2 ARE LOCATED IN THE NEW QUALITY PROGRAM MANUAL. REMOVE REFERENCES TO CHAPTER 17.2 THROUGHOUT THE USAR AND REPLACE WITH REFERENCES TO QUALITY PROGRAM MANUAL.Page: 17.2-0 Page: 14.2-16 Page: 13.4-1 Page: 3A-14 Table: 17.2-3 Table: 17.2-1 Figure: 17.2-1 Page: 17.0-i Page: 14.2-12 Page: 13.1-5 Page: 1.6-1 Page: 17.0-0 Page: 14.2-8 Page: 7.1-9 Page: 17.2-1 to 55 Page: 13.4-2 Page: 3A-15 Table: 17.2-2 Attachment III to WM 08-0003 Page 1 of 3 REVISIONS TO THE TECHNICAL REQUIREMENTS MANUAL (TRM)1. Technical Requirement (TR) Bases, TRB 3.4.17, Required Actions A.1 and A.2, were revised to clarify the determination of component or system OPERABILITY when a through-wall flaw is identified.

Regulatory Issue Summary (RIS) 2005-020, "Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability," was issued September 26, 2005. WCNOC incorporated the guidance of RIS 2005-020 into the Technical Requirements Manual (TRM) with the issuance of Revision 27 (DRR 06-0799) on May 18, 2006. An event on October 7, 2006 and a second event on January 3, 2007 (PIR 2007-000022), identified an inconsistency in the RIS 2005-020 guidance of Section C.1 1, Flaw Evaluation, and Section C.12, Operational Leakage From Code Class 1, 2, and 3 Components, and this guidance was incorporated into the TR 3.4.17 Bases.The change specifies that only the component is initially required to be declared inoperable when pressure boundary leakage is occurring.

If a reasonable expectation of OPERABILITY does not exist for the system, the system should also be declared inoperable.

Further flaw characterization and evaluation is necessary and if the evaluation determines the flaw does not meet structural integrity requirements, the system is declared inoperable.

The system may be considered OPERABLE with the component isolated as long as the system is capable of performing its specified safety function.

The proposed change is consistent with the intent of the guidance in RIS 2005-020.2. Technical Surveillance Requirement (TSR) 3.4.3.3 Bases is revised to include a discussion of the Note to the TSR. During the development of the TRM for consistency with the improved Technical Specification format, the TRM format was also revised and detailed Bases were developed for the TRM as well. In the Improved Standard Technical Specification Writers Guide, there is no specific discussion about ensuring that the Notes in the TSs are discussed in the TS Bases (WCNOC utilizes the same format/content guidance for the TRM). However, in general, it is believed that it was the intent to include in the TR Bases discussions on Notes in the TR. TSRs 3.4.3.1 and 3.4.3.2 have similar Notes in the TSR and the Notes are discussed in the TR Bases. Adding a discussion regarding the use of the Note is considered an administrative change as the TR Bases should contain sufficient information to ensure an adequate basis for an understanding of the requirements.

3. TSR 3.3.16.3 Bases are revised to incorporate changes based on the implementation of Design Change Package (DCP) 10399, "Replace HA161 & HA162 Gas Analyzer Racks." The Waste Gas Hydrogen and Oxygen Analyzers that are part of the Gaseous Radwaste System (GRWS), located in the Radwaste Building, have had maintenance and design issues that are causing system operational problems.

The analyzers were not designed to compensate for the presence of Helium (a by product of the neutron absorption of Boron 10) in the sample and consequently will not read correctly.

The helium increase over the original design is due to the extended 18-month cycle and the increase in the reactor coolant system (RCS) boron concentration to support the cycle. As a result, the analyzers read conservatively high. This conservatively high reading makes the system hard to operate during a Volume Control Tank (VCT) purge and during Waste Gas System Hydrogen recombination.

Operations has had to maintain personnel in constant Attachment III to WM 08-0003 Page 2 of 3 attendance when recombining gas from the VCT. To correct this problem, new flow panels and new analyzers that automatically compensate for helium have been installed in Gas Analyzer Racks HA1 61 and HAl 62.4. TR 3.7.20 and associated Bases is being revised to incorporate changes for the addition of new Technical Specification Limiting Condition for Operation (LCO) 3.0.8 approved in Amendment No. 173. Amendment No. 173 modifies the TS requirements for inoperable snubbers by adopting the NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification (STS) Change Traveler TSTF-372, "Addition of LCO 3.0.8, Inoperability of Snubbers." The amendment adds (1) a new LCO 3.0.8 addressing situations where one or more required snubbers are unable to perform their associated support function(s) (i.e., the snubber is inoperable) and (2) a reference to LCO 3.0.8 in LCO 3.0.1, which describes when LCOs shall be met. Changes to the TR and TR Bases are being made consistent with the approval of Amendment No. 173.5. Technical Requirement Manual (TRM) Revision 27 revised TR 3.4.17, "Structural Integrity," and associated TR Bases based on the guidance in Regulatory Issue Summary (RIS) 2005-20: Revision to Guidance Formerly Contained in NRC Generic Letter 91-18,"Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability." RIS 2005-020 was issued to inform licensees that it has revised the guidance in NRC Inspection Manual, Part 9900, Technical Guidance -this guidance supersedes the guidance previously provided in Generic Letter 91-18 and Revision 1 to Generic Letter 91-18. Appendix C, Section C. 11 (Flaw Evaluation) and Section C.12 (Operational Leakage From Code Class 1, 2, and 3 Components) of Part 9900 primarily impact TR 3.4.17. TR 3.4.17 and the associated TR Bases were revised to reflect the revised guidance provided in Part 9900. Specifically, this revision required, in part, declaring the component inoperable upon identification of a through wall leak. Revision 30 to the TRB 3.4.17, Required Actions A.1 and A.2, clarified the determination of component or system OPERABILITY when a through-wall flaw is identified.

Subsequently, on June 22, 2007, an NRC memorandum entitled, "Inspection Manual Part 9900 Operability Guidance Involving Structural Integrity of ASME Code Class 2 and 3 Piping," was issued providing guidance that indicates that the NRC agrees with the past position (pre-RIS 2005-20) on assessing OPERABILITY of ASME Code Class 2 and 3 piping with through wall leakage. TR 3.4.17 and associated Bases were revised based on the guidance in the NRC memorandum.

6. TR 3.4.16, "RCS Chemistry," and associated Bases were revised to provide an exception to allow TSR 3.4.16.1 to not be performed until a time period after entry into MODE 6 from defueled.

A Note is added to TSR 3.4.16.1 indicating that the surveillance is not required to be met for Chloride and Fluoride until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after entry into MODE 6 from a defueled state and footnote (b) is added to Table TR 3.4.16-1 to indicate that the limits are not applicable with no fuel in the reactor vessel. A 72-hour exception is reasonable based on the Frequency of the TSR being 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The word "Applicable" in Table TR 3.4.16-1 was misspelled and is being corrected.

Additionally, the Frequency for TSR 3.4.16.1 is revised from "Once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " to "72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months " per Section 1.4, "Frequency." The use of "once" implies a one-time performance Frequency and does not qualify for the 25% extension allowed by SR 3.0.2 as discussed Attachment III to WM 08-0003 Page 3 of 3 in Example 1.4-2. The Frequency specifies an interval (72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ) during which the associated TSR must be performed at least once.The RCS water chemistry is selected to minimize corrosion.

A periodic analysis of the coolant chemical composition is performed to verify that the reactor coolant quality meets the requirements of TR 3.4.16. This TR places limits on the dissolved oxygen, chloride and fluoride content of the RCS to minimize corrosion.

Both chlorides and fluorides have been shown to cause stress corrosion if present in the RCS in sufficiently high concentrations at high pressure and temperature conditions.

Stress corrosion can lead to either localized leakage or catastrophic failure of the RCS. The associated effects of exceeding the dissolved oxygen, chloride, and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the steady state limits, up to the transient limits, for the specified limited time intervals without having a significant effect on the structural integrity of the RCS. When the RCS is drained to a reduced inventory condition with all fuel removed from the reactor vessel and all loops drained, a representative sample of the RCS for chlorides and fluorides is not possible.

Sampling is not possible since sample flow through the RCS loop sampling taps or instrument taps is not possible.

Fluoride and chloride concentration limits are based on corrosion at high temperature

(> 500'F), and are not a concern from a materials standpoint at refueling temperatures of less than 150 0 F.

Enclosure I to WM 08-0003 Page 1 of 1 Subject Enclosed is the CD-ROM submittal of the Wolf Creek Updated Safety Analysis Report (USAR), Revision 21.Contact Name Mailing Address E-Mail Address Phone Number Richard Flannigan Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 riflannc)wcnoc.com 620-364-4117 Document Components:

The CD-ROM labeled "Wolf Creek Updated Safety Analysis Report" contains the following files: 001_NUSAR.pdf 002_NUSARCOl .pdf 003_NUSARC02.pdf 004_NUSARC02FIGURES.pdf 005_NUSARC03.pdf 006_NUSARC04.pdf 007_NUSARC05.pdf 008_NUSARC06.pdf 009_NUSARC07.pdf 010_NUSARC08.pdf 011iNUSARC09.pdf 012_N USARC1O.pdf 013_NUSARC 11.pdf 014 NUSARC12.pdf 015 NUSARC13.pdf 016 NUSARC14.pdf 017 NUSARC15.pdf 018 NUSARC16.pdf 019_NUSARC17.pdf 020 NUSARC18.pdf 021_NUSARNRCQ.pdf 022_USAR Rev. 21-loep.pdf 2.52 MB, publicly available 661 KB, publicly available 33.4 MB, publicly available 32.6 MB publicly available 22.0 MB, publicly available 3,209 KB, publicly available 1,572 KB, publicly available 16.6 MB, publicly available 1.91 MB, publicly available 1.02 MB, publicly available 14.3 MB, publicly available 875 KB, publicly available 1.64 MB, publicly available 909 KB, publicly available 335 KB, publicly available 370 KB, publicly available 9.70 MB, publicly available 58 KB, publicly available 92 KB, publicly available 985 KB, publicly available 308 KB, publicly available 446 KB, publicly available Enclosure II to WM 08-0003 Page 1 of 1 Subject Enclosed is the CD-ROM submittal of the station-controlled drawings that are considered incorporated by reference into the Wolf Creek Updated Safety Analysis Report (USAR). In accordance with 10 CFR 2.390, this enclosure is considered sensitive unclassified information and therefore warrants withholding.

Contact Name Mailing Address E-Mail Address Phone Number Richard Flannigan Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 riflann(cwcnoc.com 620-364-4117 Document Components:

The CD-ROM labeled "Updated Safety Analysis Report Controlled Figure Drawings Only" contains the following files: 001_Chapter 1.pdf 002_Chapter 2.pdf 003_Chapter 5.pdf 004_Chapter 6.pdf 005_Chapter 7.pdf 006_Chapter 8.pdf 007_Chapter 9.pdf 008_Chapter 10.pdf 009_Chapter 11 .pdf 010_Chapter 12.pdf 011_Chapter 18.pdf 012_Index Removed Figure List.pdf 18.6 MB, sensitive unclassified information 2.33 MB, sensitive unclassified information 3.07 MB, sensitive unclassified information 4.14 MB, sensitive unclassified information 1.68 MB, sensitive unclassified information 2.01 MB, sensitive unclassified information 39.4 MB, sensitive unclassified information 16.6 MB, sensitive unclassified information 3.12 MB, sensitive unclassified information 2.66 MB, sensitive unclassified information 207 KB, sensitive unclassified information 90.1 KB, sensitive unclassified information Enclosure III to WM 08-0003 Page 1 of 1 Subject Enclosed is the CD-ROM submittal of the station Fire Hazards Analysis and Quality Program Manual that are considered incorporated by reference into the Wolf Creek Updated Safety Analysis Report (USAR).Contact Name Mailing Address E-Mail Address Phone Number Richard Flannigan Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 riflann(cwcnoc.com 620-364-4117 Document Components:

The CD-ROM labeled "Updated Safety Analysis Report Fire Hazards Analysis and Quality Program Manual" contains the following files: 001_Quality Program Manual.pdf 002_E-19900.pdf 003_E-19905.pdf 004_E-19910.pdf 005_WIP-E-9910.pdf 006_XX-E-013-Rev 1.pdf 007_XX-E-013-Rev.

1 -CN001 .pdf 008 XX-E-013-Rev.

1 -CN002. pdf 009 'XX-E-013-Rev.

1-CN003. pdf 010_XX-E-013-Rev.1-CN005.pdf 011 XX-E-013-Rev.1-CN006.pdf 012 XX-E-013-Rev.1-CN007.pdf 013 XX-E-013-Rev.1-CN008.pdf 014 XX-E-013-Rev.1-CN009.pdf 015 XX-E-013-Rev.1-CN01O.pdf 016_M-663-00017A W03 CS to B1-98.pdf 017_M-663-00017A W03 B1-99 to B2-408.pdf 018_M-663-00017A W03 B2-41 to B6-2.pdf 019_M-663-00017A W03 B6-3 to B8-3.pdf 020_M-663-00017A W03 B8-4 to B8-147.pdf 021_M-663-00017A W03 B8-148 to B 13-25. pdf 022-M-663-00017A W03 B13-26 to G1A-63.pdf 023_M-663-00017A W03 G1A-64 to G2B-61.pdf 024_M-663-00017A W03 G2B-62 to G3D-50.pdf 025_M-663-00017A W03 ATT G4 to ATT H.pdf 263 KB, publicly available 2.21 MB, publicly available 0.98 MB, publicly available 1.83 MB, publicly available 57.5 KB, publicly available 1.35 MB, publicly available 167 KB, publicly available 192 KB, publicly available 155 KB, publicly available 120 KB, publicly available 214 KB, publicly available 174 KB, publicly available 135 KB, publicly available 107 KB, publicly available 139 KB, publicly available 43.1 MB, publicly available 41.6 MB, publicly available 49.9 MB, publicly available 46.7 MB, publicly available 48.4 MB, publicly available 39.5 MB, publicly available 46.3 MB, publicly available 47.7 MB, publicly available 34.0 MB, publicly available 35.6 MB, publicly available Enclosure IV to WM 08-0003 TRM Replacement Pages RCS Chemistry TR 3.4.16 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.16 RCS Chemistry TR 3.4.16 RCS chemistry shall be maintained within the limits specified in Table TR 3.4.16-1.APPLICABILITY:

According to Table TR 3.4.16-1.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more chemistry A.1 Restore parameter to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months parameters

> steady state within steady state limit.limit and < transient limit in MODES 1, 2, 3, and 4.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 Completion Time of Condition A not met. AND OR B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months One or more chemistry parameters

> transient limit in MODES 1, 2, 3, and 4.(continued)

Wolf Creek -Unit 1 -TRM 3.4-5 Revision 3 RCS Chemistry TR 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. ---------NOTE ---------C.1 Initiate action to reduce Immediately Required Action C.2 must pressurizer pressure to be completed whenever <500 psig.Condition C is entered.---------

AND Steady state limit C.2 Determine RCS is Prior to increasing exceeded for Chloride or acceptable for continued pressurizer Fluoride for > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months operation.

pressure > 500 while in other than psig MODES 1, 2, 3, and 4.OR OR Prior to entering Transient limit exceeded MODE 4 for Chloride or Fluoride while in other than MODES 1, 2, 3, and 4.TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.4.16.1 -----------------

NOTE --.-.---------------....---------

Not required to be met for Chloride and Fluoride until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after entry into MODE 6 from a defueled state.Verify RCS chemistry parameters within limits 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided in Table TR 3.4.16-1.Wolf Creek -Unit 1 -TRM 3.4-6 Revision 33 RCS Chemistry TR 3.4.16 Table TR 3.4.16-1 RCS Chemistry Limits APPLICABLE MODES OR OTHER SPECIFIED STEADY STATE TRANSIENT PARAMETER CONDITIONS LIMIT LIMIT 1. Dissolved Oxygen At all times (a) < 0.10 ppm < 1.00 ppm 2. Chloride At all times(b) < 0.15 ppm 1.50 ppm 3. Fluoride At all times(b) < 0.15 ppm 1.50 ppm (a) Not required with Tavg less than or equal to 250 0 F.(b) Limit not applicable with no fuel in the reactor vessel.Wolf Creek -Unit 1 -TRM S3.4-7 Revision 33 Structural Integrity TR 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.17 Structural Integrity TR 3.4.17 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained.

APPLICABILITY:

MODES 1, 2, 3, 4, 5, and 6.ACTIONS Separate Condition entry is allowed for each component.

CONDITION REQUIRED ACTION COMPLETION TIME A. --------NOTE ---------------

A.1 Declare the affected Immediately Not applicable to component(s) inoperable.

moderate-energy Class 2 or 3 piping. AND A.2 Enter applicable Immediately One or more ASME Code Conditions and Required Class 1,2, or 3 Actions of Technical component(s) contain(s)

Specification or Technical through-wall flaw. Requirement for the affected component(s).

OR Structural integrity of one or more ASME Code Class 1, 2, or 3 component(s) not maintained.(continued)

I Wolf Creek -Unit 1 -TRM3R 3.4-8 Revision 33 Structural Integrity TR 3.4.17 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. ASME Code Class 2 or 3 B.1 Verify structural integrity is 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months piping in moderate-energy still maintained.

fluid system(s) contains through-wall flaw. OR B.2 Isolate the affected 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months component(s) from service.C. Required Action and C.1 Declare the affected Immediately associated Completion component(s) inoperable.

Time of Condition B not met.OR Structural integrity of ASME Code Class 2 or 3 piping in moderate-energy fluid system(s) containing through-wall flaw not maintained.

TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.4.17.1 Perform inservice inspection of ASME Section XI In accordance with Code Class 1, 2, and 3 components.

Inservice Inspection Program Wolf Creek -Unit 1 -TRM 3.4-9 Revision 33 Snubbers TR 3.7.20 3.7 PLANT SYSTEMS 3.7.20 Snubbers TR 3.7.20 All required snubbers shall be OPERABLE, and no system, or portion thereof, has experienced an unexpected, potentially damaging transient.

APPLICABILITY:

MODES 1, 2, 3, and 4, MODES 5 and 6 for snubbers located on systems required OPERABLE.ACTIONS-----------------------------

NOTE ------------------------------

Separate Condition entry is allowed for each affected system.CONDITION REQUIRED ACTION COMPLETION TIME A. ------- NOTE ---------A.1 Declare the system Immediately Required Action A.3 must inoperable if it is known or be completed whenever indeterminate that the this Condition is entered. attached inoperable

snubber is required for system OPERABILITY.

One or more required snubbers inoperable.

AND A.2 Refer to Technical Immediately Specification LCO 3.0.8.AND A.3 Perform an engineering 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months evaluation per section 5 of Table TR 3.7.20-4 on the attached component.(continued)

Wolf Creek -Unit 1 -TRM 3.7-9 Revision 32 Snubbers TR 3.7.20 ACTION (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. A system, or portion B.1 Perform an inspection of Within 6 months thereof, has experienced required snubbers following the event an unexpected, potentially affected by transient in damaging snubber accordance with Table TR transient.

3.7.20-1.C. Required Action and C.1 Declare attached system Immediately associated Completion inoperable.

Time not met.TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.7.20.1 Perform visual inspections of each required snubber In accordance with in accordance with Table TR 3.7.20-2.

Table TR 3.7.20-3 TSR. 3.7.20.2 -----------

NOTE-- ---------This surveillance shall not be performed in MODES 1 and 2.Perform a functional test on a representative sample 18 months of each type of snubber in accordance with Table TR 3.7.20-4.TSR 3.7.20.3 Verify that the service life of mechanical snubbers is In accordance with not exceeded.

Snubber Service Life Program Wolf Creek -Unit 1 -TRM 3.7-10 Revision 321 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL PAGE (I) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -Title Page TRM Title Page TAB -Table of Contents i 14 DRR 02-1459 2/12/03 ii 25 DRR 05-1996 9/28/05 iii 25 DRR 05-1996 9/28/05 TAB -1.0 USE AND APPLICATION 1.0-1 21 DRR 04-1535 12/1/04 TAB -3.0 APPLICABILTY 3.0-1 20 DRR 04-1410 10/7/04 3.0-2 20 DRR 04-1410 10/7/04 3.0-3 12 DRR 02-0515 5/2/02 3.0-4 20 DRR 04-1410 10/7/04 TAB -3.1 REACTIVITY CONTROL SYSTEMS 3.1-1 3 DRR 99-1581 12/18/99 3.1-2 3 DRR 99-1581 12/18/99 3.1-3 11 DRR 02-0412 4/5/02 3.1-4 11 DRR 02-0412 4/5/02 3.1-5 11 DRR 02-0412 4/5/02 3.1-6 11 DRR 02-0412 4/5/02 3.1-7 11 DRR 02-0412 4/5/02 3.1-8 13 DRR 02-1063 9/26/02 3.1-9 11 DRR 02-0412 4/5/02 3.1.10 11 DRR 02-0412 4/5/02 TAB -3.3 INSTRUMENTATION 3.3-1 3.3-2 3.3-3 3.3-4 3.3-5 3.3-6 3.3-7 3.3-8 3.3-9 3.3-10 3.3-11 3.3-12 3.3-13 3.3-14 3.3-15 3.3-16 3.3-17 3.3-18 3.3-19 3.3-20 3.3-21 3.3-22 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 23 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 DRR 05-0708 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 4/20/05 Wolf Creek- UnIt 1 Revision 33 Explosive Gas Monitoring Instrumentation TR B 3.3.16 BASES TECHNICAL TSR 3.3.16.2 SURVEILLANCE REQUIREMENTS A CHANNEL OPERATIONAL TEST is to be performed every 31 days on (continued) each required channel to ensure the entire channel \MIl perform the intended function.

This test verifies the capability of the explosive gas monitoring instrumentation to detect levels of oxygen and hydrogen prior to them reaching flammable levels. The Frequency is based on operating experience.

TSR 3.3.16.3 A CHANNEL CALIBRATION is performed every 92 days. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameters with the necessary range and accuracy.

The Frequency is based on operating experience.

For the Inlet Hydrogen Monitor, the CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal 1% by volume hydrogen, at least 10% by volume helium, and balance nitrogen mixture; a nominal 4% by volume hydrogen and balance nitrogen mixture;and a 100% helium gas. For the Outlet Hydrogen Monitor, the CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal 1 % by volume hydrogen, at least 10% by volume helium, and balance nitrogen mixture; a nominal 4% by volume hydrogen and balance nitrogen mixture; and a 100% helium gas.For the Inlet Oxygen Monitor, the CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal 3% by volume oxygen, balance nitrogen mixture, and a nominal 5% by volume oxygen, balance nitrogen mixture. For the Outlet Oxygen Monitor, the CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal 25 ppm by volume oxygen, balance nitrogen mixture, and a nominal 75 ppm by volume oxygen, balance nitrogen mixture.REFERENCES 1 .USAR, Section 11.3.1.2. WCAP-1 1618, "MERITS Program-Phase 11, Task 5, Criteria Application," including Addendum 1 dated April, 1989.3. USAR Change Request 97-204.4. Technical Specification 5.5.12, "Explosive Gas and Storage Tank Radioactivity Monitoring Program." Wolf Creek -Unit 1 -TRM B331- eiin3 B 3.3.16-5 Revision 31 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -3.3 INSTRUMENTATION (continued) 3.3-23 23 DRR 05-0708 4/20/05 3.3-24 23 DRR 05-0708 4/20/05 3.3-25 23 DRR 05-0708 4/20/05 3.3-26 23 DRR 05-0708 4/20/05 3.3-27 23 DRR 05-0708 4/20/05 3.3-28 23 DRR 05-0708 4/20/05 3.3-29 26 DRR 0050 2/28/06 3.3-30 23 DRR 05-0708 4/20/05 3.3-31 23 DRR 05-0708 4/20/05 3.3-32 23 DRR 05-0708 4/20/05 TAB -3.4 REACTOR COOLANT SYSTEM (RCS)3.4-1 3 DRR 99-1581 12/18/99 3.4-2 3 DRR 99-1581 12/18/99 3.4-3 13 DRR 02-1063 9/26/02 3.4-4 6 DRR 00-1428 10/12/00 3.4-5 3 DRR 99-1581 12/18/99 3.4-6 33 DRR 07-1554 9/28/07 3.4-7 33 DRR 07-1554 9/28/07 3.4-8 33 DRR 07-1554 9/28/07 3.4-9 33 DRR 07-1554 9/28/07 3.4-10 3 DRR 99-1581 12/18/99 3.4-11 3 DRR 99-1581 12/18/99 TAB -3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3.5-1 3 DRR 99-1581 12/18/99 TAB -3.6 CONTAINMENT SYSTEMS 3.6-1 17 DRR 04-0452 5/26/04 3.6-2 3 DRR 99-1581 12/18/99 3.6-3 17 DRR 04-0452 5/26/04 TAB -3.7 PLANT SYSTEMS 3.7-1 3.7-2 3.7-3 3.7-4 3.7-5 3.7-6 3.7-7 3.7-8 3.7-9 3.7-10 3.7-11 3.7-12 3.7-13 3.7-14 3.7-15 3.7-16 3.7-17 3 28 25 25 25 25 26 25 32 32 25 25 25 25 25 25 25 DRR 99-1581 DRR 06-1349 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 06-0050 DRR 05-1996 DRR 07-1058 DRR 07-1058 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 05-1996 12/18/99 7/24/06 9/28/05 9/28/05 9/28/05 9/28/05 2/28/06 9/28/05 07/10/07 7/10/07 9/28/05 9/28/05 9/28/05 9/28/05 9/28/05 9/28/05 9/28/05 Wolf Creek -Unit 1 ii Revision 33 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -3.7 PLANT SYSTEMS (continued) 3.7-18 25 DRR 05-1996 9/28/05 3.7-20 25 DRR 05-1996 9/28/05 3.7-21 25 DRR 05-1996 9/28/05 3.7-22 25 DRR 05-1996 9/28/05 3.7-23 25 DRR 05-1996 9/28/05 3.7-24 25 DRR 05-1996 9/28/05 3.7-25 25 DRR 05-1996 9/28/05 3.7-26 25 DRR 05-1996 9/28/05 3.7-27 25 DRR 05-1996 9/28/05 3.7-28 25 DRR 05-1996 9/28/05 3.7-29 25 DRR 05-1996 9/28/05 TAB -3.8 ELECTRICAL POWER SYSTEMS 3.8-1 22 DRR 05-0167 2/16/05 3.8-2 22 DRR 05-0167 2/16/05 3.8-3 26 DRR 06-0050 2/28/06 3.8-4 26 DRR 06-0050 2/28/06 TAB -3.9 REFUELING OPERATIONS 3.9-1 3 DRR 99-1581 12/18/99 3.9-2 3 DRR 99-1581 12/18/99 3.9-3 3 DRR 99-1581 12/18/99 TAB -3.10 EXPLOSIVE GAS AND STORAGE TANK RADIOACTIVITY MONITORING 3.10-1 3 DRR 99-1581 12/18/99 3.10-2 3 DRR 99-1581 12/18/99 3.10-3 3 DRR 99-1581 12/18/99 3.10-4 3 DRR 99-1581 12/18/99 3.10-5 3 DRR 99-1581 12/18/99 3.10-6 3 DRR 99-1581 12/18/99 TAB -5.0 ADMINISTRATIVE CONTROLS 5.0-1 3 DRR 99-1581 12/18/99 5.0-2 11 DRR 02-0412 4/5/02 5.0-3 24 DRR 05-1376 6/28/05 5.0-4 24 DRR 05-1376 6/28/05 5.0-5 17 DRR 04-0452 9/28/05 5.0-6 25 DRR 05-1996 2/16/05 5.0-7 16 DRR 03-1498 11/4/03 Note 1 The page number is listed on the center of the bottom of each page.Note 2 The revision number is listed in the lower right hand corner of each page.number will be page specific.The Revision Note 3 The change document will be the document requesting the change. Therefore, the change document should be a DRR number in accordance with AP 26A-002.Note 4 The date effective or implemented is the date the Technical Requirement pages are to be issued by Document Control.Wolf Creek -Unit 1 iii Revision 33 Pressurizer Pressure/Temperature (P/T) Limits TR B 3.4.3 BASES ACTIONS A.1 and A.2 (continued)

In addition to restoring operation to within limits, an engineering evaluation of the structural integrity of the pressurizer is required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to determine if operation may continue.

This may require event-specific stress analyses or inspections.

WCAP-14717 (Ref. 6) provides guidance for engineering evaluation of transients.

A favorable evaluation must be completed before continuing operation.

The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is consistent with that allowed in Technical Specification 3.4.3,"RCS Pressure and Temperature Limits." A Note is provided to clarify that Required Action A.2 must be completed whenever this Condition is entered. The Note emphasizes the need to perform the engineering evaluation of the effects of the excursion outside the allowable limits. Restoration to within limits is insufficient without the evaluation of the structural integrity of the pressure boundary of the pressurizer.

B. 1 In the event that the Required Action and associated Completion Time are not met, Required Action B.1 requires initiation of a Performance Improvement Request (PIR) immediately to address why the pressurizer pressure/temperature limit was not restored to OPERABLE status within the Completion Time. As part of the initiation of the PIR, action shall be implemented in a timely manner to place the unit in a safe condition as determined by plant management.

The PIR should provide an accurate description of the problem, the Required Action and associated Completion Time not complied with. The intent of this Required Action is to utilize the corrective action program to assure prompt attention and adequate management oversight to minimize the additional time pressurizer temperature is not within limits.TECHNICAL TSR 3.4.3.1 SURVEILLANCE REQUIREMENTS This TSR verifies the rate of heatup is within limit. The 30 minute Frequency is considered reasonable based on the instrumentation available in the control room to monitor the status of the RCS. The TSR has been modified by a Note which requires the TSR to be performed only during pressurizer heatup.Wolf Creek -Unit 1 -TRM B 3.4.3-3 Revision 10 Pressurizer Pressure/Temperature (P/T) Limits TR B 3.4.3 BASES TECHNICAL SURVEILLANCE REQUIREMENTS (continued)

TSR 3.4.3.2 TSR 3.4.3.2. verifies the rate of cooldown is within limits. "Step-wise" cooling must be avoided as discussed in Reference

4. The Frequency of 30 minutes is considered reasonable based on the instrumentation available in the control room to monitor the status of the RCS. The TSR has been modified by a Note, which requires the TSR to be performed only during pressurizer cooldown.TSR 3.4.3.3 This TSR verifies that the maximum spray water temperature differential does not exceed 5830F. This is to guard against subjecting the spray nozzle to undue thermal stresses.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is considered adequate based upon operating experience.

The TSR has been modified by a Note, which requires the TSR to be performed only during auxiliary spray operation.

REFERENCES

1. 10 CFR 50.2, "Definitions." 2. 10 CFR 50.55a, "Codes and Standards." 3. ASME Boiler and Pressure Vessel Code,Section III, Appendix G.4. Westinghouse letter SAP-90-263, "Reactor Coolant System Accelerated Cooldown," dated November 5, 1990.5. USAR, Section 5.3.3.1.6. WCAP-14717, Rev. 1, Supplement 1, "Basis and Application of Pressurizer Heatup and Cooldown Limits," November 1999.Wolf Creek -Unit 1 -TRM B 3.4.3-4 Revision 31 RCS Chemistry TR B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)TR B 3.4.16 RCS Chemistry BASES BACKGROUND The Reactor Coolant System (RCS) water chemistry is selected to minimize corrosion.

A periodic analysis of the coolant chemical composition is performed to verify that the reactor coolant quality meets the specification (Ref. 1). This Technical Requirement places limits on the dissolved oxygen, chloride and fluoride content of the RCS to minimize corrosion.

Limiting dissolved oxygen content of the RCS limits the amount of general corrosion and reduces the possibility of stress corrosion.

General corrosion is a contributing factor in reactor coolant activity and must be controlled for ALARA (as low as reasonably achievable) considerations as well as structural integrity considerations.

Both chlorides and fluorides have been shown to cause stress corrosion if present in the RCS in sufficiently high concentrations at high pressure and temperature conditions.

Stress corrosion can lead to either localized leakage or catastrophic failure of the RCS. The associated effects of exceeding the dissolved oxygen, chloride, and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the steady state limits, up to the transient limits, for the specified limited time intervals without having a significant effect on the structural integrity of the RCS.APPLICABLE SAFETY ANALYSES The limitations on RCS chemistry ensure that corrosion of the RCS is minimized and reduces the potential for RCS LEAKAGE or failure due to stress corrosion.

Maintaining the chemistry within the steady state limits provides adequate corrosion protection to ensure the structural integrity of the RCS over the life of the plant. It is not, however, a consideration in the analyses of Design Basis Accidents.

The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the steady state limits, up to the transient limits, for the specified limited time intervals without having a significant effect of the structural integrity of the RCS.Wolf Creek -Unit 1 -TRM B 3.4.16-1 Revision 3 RCS Chemistry TR B 3.4.16 BASES TR TR 3.4.16 establishes the steady state and transient limits on concentration of dissolved oxygen, chloride and fluoride in the RCS.These limits ensure that dissolved oxygen, chloride and fluoride concentrations are maintained at levels low enough to prevent unacceptable degradation of the RCS pressure boundary.APPLICABILITY Concentrations of dissolved oxygen, chloride and fluoride in the RCS must be maintained within limits at all times as specified in Table TR 3.4.16-1.

Applicability is modified by a Note on Table TR 3.4.16-1 indicating that dissolved oxygen is not required with T 2 vg:! 250OF and the chloride and fluoride limits are not applicable when there is no fuel in the reactor vessel.ACTIONS A. 1 If one or more chemistry parameters are not within steady state limits in MODES 1, 2, 3, or 4, the parameter(s) must be restored to their steady state limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This allows time to take corrective actions to restore the contaminant concentrations to within the steady state limits.B.1 and B.2 With one or more chemistry parameters not within transient limits in MODES 1, 2, 3, or 4, or if the Required Action of Condition A is not met within the associated Completion Time, the plant must be placed in a condition where the limit is not applicable or where corrosion rates are reduced. This is accomplished by placing the plant in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . In MODE 5, the dissolved oxygen limit is not applicable and stress corrosion rates are reduced. The 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allotted to reach MODE 3 is a reasonable time, based on operating experience, to shutdown the plant from full power in an orderly manner and without challenging plant systems. The Completion Time of 36 allows for a reasonable time to reach MODE 5 from full power conditions in and orderly manner and without challenging plant systems.If high chloride or fluoride concentrations are the reason for entering Condition B, and the condition is not corrected before entering MODE 5, Required Actions C.1 and C.2 must be performed.

Wolf Creek -Unit 1 -TRM B341- eiin3 B 3.4.16-2 Revision 33 RCS Chemistry TR B 3.4.16 BASES ACTIONS (continued)

C.1 If RCS chloride or fluoride concentration are not within steady state limits for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any condition other than MODES 1, 2, 3, or 4, or if RCS chloride or fluoride concentration are not within transient limits for any amount of time in any condition other than MODES 1, 2, 3, or 4, action must be immediately initiated to reduce pressurizer pressure to< 500 psig unless it is already below 500 psig. The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.A Note is added to Condition C stating that Required Action C.2 must be completed whenever this Condition is entered.C.2 In addition to Required Action C.1, an engineering evaluation must be performed to determine the effects of the out-of-limit condition on the structural integrity of the RCS and that the RCS remains acceptable for continued operation.

These actions must be taken prior to increasing pressurizer pressure above 500 psig or prior to entry in MODE 4. These evaluations are necessary because of the time/temperature/concentration dependency of the effects of exceeding the limits. Corrosion evaluations for conditions outside the limits are made on a case by case basis.TECHNICAL SURVEILLANCE REQUIREMENTS TSR 3.4.16.1 TSR 3.4.16.1 requires through a representative sample that the chemistry parameters are within their limits of Table TR 3.4.16-1.

The TSR is modified by a Note stating that the chloride and fluoride limit is not required to be met until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after entry into MODE 6 from a defueled state. Representative samples for chloride and fluoride concentration can not be obtained during defueled conditions when the RCS and Residual Heat Removal System is not recirculating.

Therefore, the Note allows entry into MODE 6 when a representative sample can be obtained.

A Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provides adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.REFERENCES

1. USAR, Section 5.2.3.2.Wolf Creek -Unit 1 -TRM B 3.4.16-3 Revision 33 Structural Integrity TR B 3.4.17 B 3.4 REACTOR COOLANT SYSTEM TR B 3.4.17 Structural Integrity BASES BACKGROUND The quality group classification for each water- and steam- containing pressure component is shown in USAR Table 3.2-1. The components are classified according to their safety significance as dictated by service and functional requirements and by the consequences of their failure.The quality group classifications and code requirements for the quality of plant process systems meet the intent of Regulatory Guides 1.26 and 1.143 (Ref. 1).The design, fabrication, inspection, and testing requirements of each classification provide the required degree of conservatism in assuring component pressure integrity and OPERABILITY (Ref. 1).The Code requirements applicable to each quality group classification are identified in USAR Table 3.2-2. The quality group classifications and the interfaces between classifications in a system having components of different classifications are indicated on the piping and instrumentation diagram or flow diagram of that system (Ref. 1).APPLICABLE Certain components which are designed and manufactured to the SAFETY ANALYSES requirements of specific sections of the ASME Boiler and Pressure Vessel Code are part of the primary success path and function to mitigate DBAs and transients.

However, the OPERABILITY of these components is addressed in the relevant specifications that cover individual components.

TR TR 3.4.17 requires the structural integrity of ASME Code Class 1,2, and 3 components be maintained.

Structural integrity is the functional capability (i.e., pressure retaining capability) of ASME Code Class 1,2, and 3 systems during all design and operational conditions.

In those areas where conflict may exist between the Technical Requirements and the ASME Boiler and Pressure Vessel Code, the Technical Requirements Manual takes precedence.

APPLICABILITY The structural integrity of the ASME Code Class 1, 2, and 3 components is required during MODES 1, 2, 3, 4, 5, and 6. This TR applies whenever Code Class 1, 2, or 3 components are OPERABLE, not just during the performance of inservice inspection examinations.

Wolf Creek -Unit 1 -TRM B 3.4.17-1 Revision 3 Structural Integrity TR B 3.4.17 BASES ACTIONS A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Requirement may be entered independently for each component.

The Completion Time(s) for the inoperable component will be tracked separately for each component starting from the time the Condition was entered for that component.

A.1 and A.2 Condition A applies to one or more degraded or non-conforming ASME Code Class 1, 2, or 3 component(s) (excluding moderate-energy Class 2 or 3 piping) containing the specific degradation of a through-wall flaw or degraded or non-conforming ASME Code Class 1, 2, or 3 component(s), not including through-wall flaws, for which structural integrity is not maintained.

Condition A is modified by a Note which limits the applicability of the Condition of one or more ASME Code Class 1, 2, or 3 component(s) containing a through-wall flaw by excluding moderate-energy Class 2 or 3 piping. With one or more ASME Code Class 1, 2, or 3 component(s) containing a through-wall flaw (excluding moderate energy Class 2 or 3 piping) or one or more ASME Code Class 1, 2, or 3 components, not including through-wall flaws (no limitation on component applicability) for which structural integrity is not maintained, Required Actions A.1 and A.2 require the affected component(s) to be declared inoperable and the applicable Conditions and Required Actions of the Technical Specifications or Technical Requirement be entered for the affected component.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.Moderate-energy fluid systems are systems that during normal plant conditions meet both of the following conditions:

a. maximum operating temperature is 200OF or less, and b. maximum operating pressure is 275 psig or less.In accordance with 10 CFR 50.55a(g), structural integrity must be maintained in conformance with ASME Code Section XI for those components that are subject to Code requirements.

Depending on the type of degraded or nonconforming condition, structural integrity can be determined by meeting applicable ASME Section III requirements,Section XI acceptance standards,Section XI IWB-3600 analytical Wolf Creek -Unit 1 -TRM B 3.4.17-2 Revision 33 Structural Integrity TR B 3.4.17 BASES ACTIONS A.1 and A.2 (continued) evaluation procedures, or applicable NRC approved ASME Code Cases.In addition, NRC Inspection Manual, Part 9900, Technical Guidance (Regulatory Issue Summary 2005-020 (Ref. 2), Appendix C, Section C.10 (Piping and Pipe Support Requirements), C. 11 (Flaw Evaluation), and C.12 (Operational Leakage From Code Class 1, 2, and 3 Components) contain specific NRC guidance and acceptable evaluation procedures than can be used to determine structural integrity and OPERABILITY for ASME Code Class 1, 2, and 3 components.

Modified NRC OPERABILITY guidance was provided in Reference 7.If a flaw is discovered by any means (including surveillance, maintenance activity, or inservice inspection) in a system subject to Code requirements (whether during normal plant operation, plant transition, or shutdown operation), the flaw must be promptly evaluated to assure structural integrity is maintained.

If the flaw results in through-wall leakage of a Class 1, 2, or 3 component (excluding moderate-energy Class 2 or 3 piping), the component containing the flaw is inoperable.

If a flaw evaluation determines that structural integrity is not maintained, the component and associated system containing the flaw are inoperable.

The system may be considered OPERABLE with the component isolated as long as the system is capable of performing its specified safety function.

If the flaw evaluation determines that structural integrity is maintained, the component and associated system is OPERABLE.However, a determination should be made as to how long the flawed component will remain OPERABLE before the flaw grows to exceed acceptable limits and criteria.Non Throuqh-Wall Flaws Use of Generic Letter (GL) 90-05 was included in NRC Inspection Manual, Part 9900, Technical Guidance (Ref. 2) (Appendix C, Section C.1 1) addressing flaws in Class 3 piping not resulting in through-wall leaks. GL 90-05 was originally issued to address temporary non-Code repairs. It also provided NRC approved guidance in assessing structural integrity in moderate-energy Class 3 piping as part of the licensee relief request when a Code repair was deferred.

10 CFR 50.55a(b)(2)(xiii)(Ref. 3) modified the GL 90-05 guidance in 1999 by approving ASME Section Xl Code Case N-513 and Case N-523-1 (Ref. 6). The Federal Register issuing this amendment noted "These Code Cases were developed to address criteria for temporary acceptance of flaws (including through-wall leaking) of moderate energy Class 3 piping where a Section Wolf Creek -Unit 1 -TRM B 3.4.17-3 Revision 33 Structural Integrity TR B 3.4.17 BASES ACTIONS A.1 and A.2 (continued)

XI Code repair may be impractical for a flaw detected during plant operation (i.e., a plant shutdown would be required to perform the Code repair). In the past, licensees had to request NRC staff approval to defer Section XI Code repair for these Class 3 moderate energy (200 deg. F, 275 psig) piping systems ....... The use of Code Case N-513, with the limitations, and Code Case N-523-1 will obviate the need for licensees to request approval for deferring repairs; thus saving NRC and licensee resources." The NRC has approved ASME Code Case N-513-1 in Regulatory Guide 1.147, Rev. 14, with limitations that must be implemented with the Case.Code Case N-513-1 has been adopted for use at WCGS. Code Case N-523-1 has been incorporated into Appendix IX of the ASME Section XI edition and addenda used at WCGS. As a result, Appendix IX is used in lieu of Code Case N-523-1 for resolution of degraded and non-conforming conditions (including through-wall leaks) by installation of a structural mechanical clamp to restore structural integrity.

Therefore, when Code Case N-513-1 is applicable (i.e., the scope of Code Case N-513-1 and the NRC limitations applied to the use of the Case allow the Case to be used), Case N-513-1 should be considered before use of GL 90-05 to temporarily accept degraded and nonconforming conditions in Class 2 and 3 moderate-energy piping. GL 90-05 guidance on temporary non-Code repairs remains applicable.

In addition, if Code Case N-513-1 is not applicable to a degraded or nonconforming condition in moderate-energy Class 3 piping, GL 90-05 should be considered recognizing the associated need to request relief from the NRC.Alternative evaluation procedures and/or acceptance criteria may also be considered for confirming structural integrity of degraded or nonconforming conditions.

However, when alternative evaluation procedures and/or acceptance criteria are used as a basis for structural integrity and acceptable continued service, the component and associated system containing the degraded or nonconforming condition is considered inoperable until NRC approval of procedures and criteria is obtained (Ref. 2).Wolf Creek -Unit 1 -TRMB 7 B 3.4.17-4 Revision 33 Structural Integrity TR B 3.4.17 BASES ACTIONS A.1 and A.2 (continued)

Through-Wall Flaws The NRC also provided guidance in NRC Inspection Manual, Part 9900, Technical Guidance, (Regulatory Issue Summary 2005-020 (Ref. 2)) for addressing specific degraded and nonconforming conditions consisting of through-wall pressure boundary leaks. In NRC Inspection Manual, Part 9900, Technical Guidance, Appendix C, Section C.12, the NRC stated that "Upon discovery of leakage from a Class 1, 2, or 3 pressure boundary component (pipe wall, valve body, pump casing, etc.) the licensee must declare the component inoperable." Evidence of leakage from the pressure boundary indicates the presence of a through-wall flaw.The NRC further stated that because the size of the through-wall flaw on the inside surface is unknown, the component is declared inoperable while methods such as ultrasonic examination are performed to characterize the actual geometry of the through-wall flaw. However, Reference 7 provided modified NRC guidance for moderate-energy Class 2 or 3 piping with through-wall leakage.In Reference 2, Appendix C, Section C.12, the NRC noted that structural mechanical clamps may be used on high-energy Class 2 or 3 piping 2 inches and smaller to restore structural integrity of through-wall leaks.However, as discussed above, this piping must be declared inoperable when the through-wall leakage is discovered and must remain inoperable until the Section Xl Appendix IX requirements are met for the installation of the mechanical clamp.If NRC approved guidance or generically approved alternatives are not available for resolution of the through-wall leaks in ASME Code Class 1, 2, or 3 components (excluding moderate-energy Code Class 2 or 3 piping) the component(s) must have a repair/replacement activity performed in accordance with ASME Section XI, or relief from Code requirements must be requested and approval obtained from the NRC prior to declaring the component(s)

OPERABLE.B.1 and B.2 Condition B applies to ASME Code Class 2 or 3 piping in moderate-energy fluid system(s) containing a through-wall flaw. Moderate-energy fluid systems are systems that during normal plant conditions meet both of the following conditions:

a. maximum operating temperature is 200°F or less, and b. maximum operating pressure is 275 psig or less.Wolf Creek -Unit 1 -TRM B 3.4.17-5 Revision 33 Structural Integrity TR B 3.4.17 BASES ACTIONS B.1 and B.2 (continued)

The NRC provided guidance in Reference 2 for addressing the specific degraded and non-conforming condition consisting of through-wall pressure boundary leaks. However, Reference 7 provided modified NRC guidance for moderate-energy Class 2 or 3 piping with through-wall leakage. The modified guidance did not require moderate-energy Class 2 or 3 piping with through-wall leakage to be declared inoperable upon discovery of the leakage. The modified guidance stated that through-wall leakage immediate OPERABILITY determination should be based on reasonable expectation of OPERABILITY in accordance with the guidance of Reference

2. It further stated that through-wall leakage prompt OPERABILITY determination should be based on actual NDE results to characterize the flaw(s) dimensions causing the through-wall leakage and engineering analysis methods acceptable to the NRC, i.e., NRC approved Code Cases and GL 90-05.Required Action B.1 stipulates verifying within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that structural integrity is still maintained with a through-wall flaw. Structural integrity for through-wall leakage is determined in accordance with applicable ASME Code Cases and NRC Inspection Manual, Part 9900, Technical Guidance (Ref. 2), Appendix C, Section C.11 (Flaw Evaluation), and C.12 (Operational Leakage From Code Class 1, 2, and 3 Components), as modified by Reference

7. These contain specific NRC guidance and acceptable evaluation procedures to be used in determining structural integrity and OPERABILITY for moderate-energy ASME Code Class 2 or 3 piping.For Class 3 moderate-energy piping, the structural integrity of the piping may be evaluated by fully characterizing the extent of the through-wall flaw using volumetric methods and evaluating the flaw using the criteria of paragraph C.3.a of Enclosure 1 to GL 90-05 (Ref. 4). If the flaw meets the criteria, the piping is still considered OPERABLE but degraded until relief from the applicable Code requirement or requirements is obtained from the NRC. The structural integrity of leaking Class 2 or 3 moderate-energy piping may also be evaluated using criteria of Code Case N-513-1. The NRC has approved Code Case N-513-1 (Ref. 5) in Regulatory Guide 1.147, Rev. 14, with limitations that must be implemented along with Code Case N-513-1. If the piping meets the criteria of ASME Code Case N-513-1, with NRC limitations, the piping may be deemed OPERABLE and continued temporary service of the degraded piping is permitted.

Wolf Creek -Unit 1 -TRM B 3.4.17-6 Revision 33 Structural Integrity TR B 3.4.17 BASES ACTIONS B.1 and B.2 (continued)

Additional Bases description for use of GL 90-05 and Code Case N-513-1 is included in Required Actions A.1 and A.2 above.In Reference 2, Appendix C, Section C.12, the NRC noted that structural mechanical clamps may be used on 6 inch and smaller moderate-energy Class 2 or 3 piping to restore structural integrity of through-wall leaks.However, if the structural integrity has not been verified by use of GL 90-05 or Code Case N-513-1, the piping containing the through-wall leak is inoperable until the ASME Section Xl, Appendix IX requirements are met for the installation of the mechanical clamp.If NRC approved guidance or generically approved altematives are not available for resolution of through-wall leaks in moderate-energy Class 2 or 3 piping, the piping containing the leak must have a repair/replacement activity performed in accordance with ASME Section XI, or relief from Code requirements must be requested and approval obtained from the NRC prior to declaring the piping OPERABLE.The Completion Time of Required Action B.1 is expected to be a maximum time limit. The guidance of Reference 2 and Reference 7 would expect an earlier Completion Time if possible, consistent with the safety significance of the component(s) containing the flaw. With this clarification, the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is reasonable and consistent with the guidance in References 2 and 7.Alternatively, Required Action B.2 allows the affected component(s) to be isolated within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .C.1 If the Required Actions and associated Completion Times of Condition B are not met or if structural integrity of the Class 2 or 3 piping (in moderate-energy fluid system(s))

containing a through-wall flaw is not maintained, the affected component(s) should be immediately declared inoperable.

Wolf Creek -Unit 1 -TRM B 3.4.17-7 Revision 33 Structural Integrity TR B 3.4.17 BASES TECHNICAL TSR 3.4.17.1 requires performing inservice inspections of ASME Section SURVEILLANCE XI Code Class 1, 2, and 3 components in accordance with the Inservice REQUIREMENTS Inspection Program described in TR 5.5.6, "Inservice Inspection Program." Inservice inspection of ASME Code Class 1, 2, and 3 components are performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code (Ref. 1) and applicable Addenda, as required by 10 CFR 50.55a(g), to ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. Exception to these requirements apply where relief has been granted by the Commission pursuant to 10 CFR 50.55a(a)(3) and (g)(6)(i).

The surveillance intervals specified in Section XI of the ASME Code apply.During design and construction, components of the Reactor Coolant System were designed to provide access to permit inservice inspection in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975.REFERENCES

1. USAR, Section 3.2.2 2. Nuclear Regulatory Issue Summary 2005-020 (including NRC Inspection Manual, Part 9900, Technical Guidance, "Operability Determinations

& Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety").3. 10 CFR 50.55a(b)(2)(xiii), "Flaws in Class 3 Piping," Federal Register (Vol. 64, No. 183), September 22, 1999.4. Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping," June 15, 1990; and August 16, 1990, NRR letter by J. E. Richardson,"Follow-up on Generic Letter 90-05 Regarding Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2, and 3 Piping," (letter 90-02380).

5. Code Case N-513-1, "Evaluation Criteria for Temporary Acceptance of Flaws in Class 2 and 3 Piping." 6. ASME Section XI, Appendix IX, "Mechanical Clamping Devices for Class 2 and 3 Piping Pressure Boundary" (previously contained in Code Case N-523-1).Wolf Creek -Unit 1 -TRM B 3.4.17-8 Revision 33 Structural Integrity TR B 3.4.17 BASES REFERENCES

7. NRC Memorandum, "Inspection Manual Part 9900 Operability (continued)

Guidance Involving Structural Integrity of ASME Code Class 2 and 3 Piping," June 22, 2007.Wolf Creek -Unit 1 -TRM B 3.4.17-9 Revision 33 Snubbers TR B 3.7.20 BASES ACTIONS A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Technical Requirement may be entered independently for each affected system. The Completion Time(s) of the inoperable snubber will be tracked separately for each affected system starting from the time the Condition was entered for that system as a result of discovery of an inoperable snubber.A.1, A.2, and A.3 If one or more required snubbers have been declared inoperable, declare the system inoperable if it is known or indeterminate that the attached inoperable (nonfunctional) snubber is required for system OPERABILITY.

The system may be declared OPERABLE while utilizingTechnical Specification (TS) LCO 3.0.8. Technical Specification LCO 3.0.8 provides an allowance under which systems are not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s).

The following are applicable for utilizing LCO 3.0.8: a. Determine whether a TS system is rendered inoperable by a inoperable snubber (1) If it is determined that the supported TS system(s) do not require the snubber(s) to be OPERABLE in order to support OPERABILITY of the system(s)

LCO 3.0.8 is not needed.(2) If the LCO(s) associated with any supported TS system(s)are not currently applicable (i.e., the plant is not in a MODE or other specified condition in the Applicability of the LCO), LCO 3.0.8 is not needed.(3) If the support TS system(s) are inoperable for reasons other than snubbers, LCO 3.0.8 cannot be used.b. Determine the design basis of the inoperable snubber. When a snubber is to be rendered inoperable for testing or maintenance or is discovered to be inoperable, the design function of the snubber must be determined in order to determine if LCO 3.0.8 may be used.Wolf Creek -Unit 1 -TRM B 3.7.20-3 Revision 32 Snubbers TR B 3.7.20 BASES ACTIONS A. 1, A.2, and A.3 (continued)

(1) If the design function of the snubber is to react to only seismic loads, LCO 3.0.8 may be applied.(2) If the design function of the snubber includes both seismic loads and non-seismic loads, any TS systems supported by the inoperable snubber must be able to remain OPERABLE if subjected to the non-seismic loads with the snubber removed. If the supported TS system will remain OPERABLE when subjected to non-seismic loads, LCO 3.0.8 may be applied.(3) If the design function of the snubber includes only non-seismic loads, LCO 3.0.8 cannot be used.c. When LCO 3.0.8.a is used, at least one train of the Auxiliary Feedwater (AFW) System not associated with the inoperable snubber must be OPERABLE.d. When LCO 3.0.8.b is used, at least one train of the AFW System not associated with the inoperable snubber must be OPERABLE.If the inoperable snubber(s) supports all trains of the AFW System or if the AFW System becomes inoperable due to an emergent condition, some alternative means of core cooling must be available.

e. The LCO 3.0.8 requirement to assess and manage risk is met by programs to comply with the requirements of paragraph (a)(4) of the Maintenance Rule, 10 CFR 50.65, to assess and manage risk resulting from maintenance activities.

LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issue properly addressed.

The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function.f. A record of the implementation of any applicable Tier 2 restrictions, and the associated plant configuration, shall be available on a recoverable basis for NRC inspection.

Wolf Creek -Unit 1 -TRM B 3.7.20-4 Revision 32 Snubbers TR B 3.7.20 BASES ACTIONS A.1, A.2, and A.3 (continued).

Condition A is modified by a Note that requires that Required Action A.3 be completed whenever Condition A is entered. Thus, if the snubber is restored to OPERABLE status, the Condition will require the completion of an engineering evaluation per section 5 of Table TR 3.7.20-4.The engineering evaluation is performed to: a. Determine the cause of the failure As a result of this evaluation, the need for testing other snubbers will be considered.

The results from the testing will be used to consider expanded functional testing and cause examination with consideration of manufacturing and design deficiency.

b. Determine the impact on the supported component This evaluation shall determine if the inoperable snubber has adversely affected the attached component.

The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on engineering experiences and is reasonable, considering the time it will take to identify the problem and take the proper corrective actions.B. 1 If the plant has experienced an unexpected, potentially damaging snubber transient, an inspection per Table TR 3.7.20-1 is performed on all snubbers attached to sections of systems that have experienced the transient.

The potential impact of the transient is assessed by reviewing operating data and by visually inspecting the associated system. In addition to the visual inspection, the freedom-of-motion of the mechanical snubber(s) is verified per Table TR 3.7.20-1.Wolf Creek -Unit 1 -TRM B 3.7.20-5 Revision 32 1 Snubbers TR B 3.7.20 BASES ACTIONS B.1 (continued)

The Completion Time of 6 months has been assigned based upon industry practice.C.1 If Required Actions and associated Completion Times of Condition A or B are not met, the supported system or component is immediately declared inoperable.

TECHNICAL Surveillance Testing is performed in accordance with the applicable SURVEILLANCE requirements of ASME Section XI, "Rules for Inservice Inspection of REQUIREMENTS Nuclear Power Plant Components" (Ref. 1).NRC letter dated June 2, 2006 (Ref. 10), approved the proposed alternative to use TRM, Section 3.7.20, for snubber visual inspection and functional testing in lieu of the applicable ASME Code requirements specified in Section Xl, Article IWF-5000 for the third 10-year inservice inspection interval.

The NRC safety evaluation specifies that changes to the TRM snubber visual inspection and functional testing requirements shall be submitted to the NRC for authorization pursuant to 10 CFR 50.55a(a)(3) or as an exemption pursuant to 10 CFR 50.12.Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the completion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions.

In order to establish the inspection frequency for each type of snubber on a safety related system, it was assumed that the frequency of snubber failures and initiating events is constant with time and that the failure of any snubber could cause the system to be unprotected and to result in failure during an assumed initiating event. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection.

However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval.

Any inspection whose results require a shorter inspection interval will override the previous schedule.Wolf Creek -Unit 1 -TRM B 3.7.20-6 Revision 32 1 Sn ubbers TR B 3.7.20 BASES TECHNICAL TSR 3.7.20.1 SURVEILLANCE REQUIREMENTS TSR 3.7.20.1 comprises a visual inspection of the snubbers.

A pre-fuel load visual inspection and functional test has been performed on each snubber using the acceptance criteria listed in Table TR 3.7.20-2.

The baseline takes into account that the snubbers have experienced thermal cycling and normal operating service as a result of previous hot functional testing. The initial inservice inspection has been performed on the snubbers prior to completion of the first refueling outage. The frequency of subsequent surveillances depends on the number of snubbers found inoperable from each previous inspection as provided in Table TR 3.7.20-3 and the Inservice Inspection Program as described in TR 5.5.6. The acceptance criteria and corrective actions are listed. in Table TR 3.7.20-2.The visual inspections are designed to detect obvious indications of inoperability of the snubbers.

Removal of insulation or direct contact with the snubbers is not required initially.

However, suspected causes of inoperability are to be investigated and all snubbers of the same type and all snubbers subjected to the same failure mode are to be inspected more frequently.

The visual inspection frequency is based upon maintaining a constant level of snubber protection during an earthquake or severe transient and the number of unacceptable snubbers found during the previous inspection.

As a result, the required inspection intervals vary inversely with the number of inoperable snubbers found during an inspection.

If a snubber fails the visual acceptance criteria, the snubber is declared unacceptable and cannot be declared OPERABLE via functional testing.However, if the cause of rejection is understood and remedied for that type of snubber and for any other type of snubbers that may be generically susceptible and OPERABILITY verified by testing, that snubber may be reclassified acceptable for the purpose of establishing the next surveillance interval.Snubbers may be categorized according to accessibility as noted in the Notes to Table TR 3.7.20-3.

The accessibility of each snubber is determined based on radiation level as well as other factors such as temperature, atmosphere, location, etc. The recommendations of Regulatory Guide 8.8, "Information Relevant to Maintaining Occupational Radiation Exposure as Low a Practicable," (Ref. 7) and Regulatory Guide 8. 10, "Operation Philosophy for Maintaining Occupational Radiation Exposure as Low as Practicable," (Ref. 8) are considered in planning and implementing the visual inspection program.Wolf Creek -Unit 1 -TRM B372- eiin3 B 3.7.20-7 Revision 32 1 BASES TECHNICAL TSR 3.7.20.1 (continued)

SURVEILLANCE REQUIREMENTS Since the visual inspections are augmented by a functional testing program, the visual inspection need not be a hands on inspection, but shall require visual scrutiny sufficient to assure that fasteners or mountings for connecting the snubbers to supports or foundations have no visible bolts, pins or fasteners missing, or other visible signs of physical damage such as cracking or loosening.

TSR 3.7.20.2 This TSR is modified by a Note which restricts the performance of thisTSR to during periods of plant shutdown.TSR 3.7.20.2 comprises the functional testing of snubbers.

The testing for these snubbers have been separated into two sample plans as described in Table TR 3.7.20-4.

Sample Plan 1.a (10%) is typically used for the snubbers with small population.

Sample plan 1.b. (Figure TR 3.7.20-1) is typically used for snubbers with large population.

Figure TR 3.7.20-1 was developed using 'Wald's Sequential Probability Ratio Plan" as described in "Quality Control and Industrial Statistics" by Acheson J.Duncan.The sample plan shall be selected prior to the test period and cannot be changed during the test period.Snubber functional testing is performed to the requirements of Table TR 3.7.20-4 and performed prior to the completion of each refueling outage. The 18 month Frequency, in conjunction with the Note, is based on the need to perform this surveillance under the conditions that apply during a unit outage.TSR 3.7.20.3 This TSR addresses the monitoring of the service life of the snubbers in accordance with the Snubber Service Life Program described in TR 5.5.5.The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

Wolf Creek -Unit 1 -TRM B 3.7.20-8 Revision 32 1 BASES REFERENCES

1. ASME Boiler and Pressure Vessel Code,Section III and Xl.2. Regulatory Guide 1.124, "Design Limits and Loading Combinations for Class 1 Linear-Type Component Supports," Revision 1, January 1978.3. Regulatory Guide 1.130, "Design Limits and Loading Combinations for Class 1 Plate-and Shell-Type Component Supports," Revision 1, October 1978.4. "Zion Probabilistic Safety Study", Commonwealth Edison Company, September 1981.5. "Millstone Unit 3 Probabilistic Safety Study," North-East Utilities Company, August 1983.6. NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications.

Attachment to letter dated May 1988 from T. E. Murley, NRC to W.S. Wilgus, Chairman the B&W Owners Group.7. Regulatory Guide 8.8, "Information Relevant to Maintaining Occupational Radiation Exposure as Low as Practicable." 8. Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational Radiation Exposure as Low as Practicable." 9. WCAP-1 1618, "MERITS Program-Phase II, Task 5, Criteria Application," including Addendum 1 dated April, 1989, Section 3.7.9.10. NRC letter (D. Terao to R. Muench) dated June 2, 2006, "Wolf Creek Generating Station -Relief Request 13R-03 for the Third 10-Year Interval Inservice Inspection and Examination of Snubbers (TAC NO. MC8571)." Wolf Creek -Unit 1 -TRM B 3.7.20-9 ReviOn 32 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL BASES PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -Table of Contents i 14 DRR 02-1459 2/12/03 ii 25 DRR 05-1996 9/28/05 iii 21 DRR 04-1535 12/1/04 TAB -B 3.0 APPLICABILTY B 3.0-1 3 DRR 99-1581 12/18/99 B 3.0-2 3 DRR 99-1581 12/18/99 B 3.0-3 3 DRR 99-1581 12/18/99 B 3.0-4 20 DRR 04-1410 10/7/04 B 3.0-5 20 DRR 04-1410 10/7/04 B 3.0-6 20 DRR 04-1410 10/7/04 B 3.0-7 20 DRR 04-1410 10/7/04 B 3.0-8 20 DRR 04-1410 10/7/04 B 3.0-9 20 DRR 04-1410 10/7/04 B 3.0-10 20 DRR 04-1410 10/7/04 B 3.0-11 20 DRR 04-1410 10/7/04 B 3.0-12 20 DRR 04-1410 10/7/04 B 3.0-13 20 DRR 04-1410 10/7/04 B 3.0-14 20 DRR 04-1410 10/7/04 TAB -B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7-1 3 DRR 99-1581 12/18/99 B 3.1.7-2 3 DRR 99-1581 12/18/99 B 3.1.7-3 3 DRR 99-1581 12/18/99 B 3.1.8-1 11 DRR 02-0412 4/5/02 B 3.1.8-2 11 DRR 02-0412 4/5/02 B 3.1.9-1 15 DRR 03-0861 7/10/03 B 3.1.9-2 8 DRR 01-0475 5/1/01 B 3.1.9-3 8 DRR 01-0475 5/1/01 B 3.1.9-4 8 DRR 01-0475 5/1/01 B 3.1.9-5 8 DRR 01-0475 5/1/01 B 3.1.9-6 28 DRR 06-1349 7/24/06 B 3.1.9-7 28 DRR 06-1349 7/24/06 B 3.1.9-8 3 DRR 99-1581 12/18/99 B 3.1.10-1 10 DRR 02-0122 2/28/02 B 3.1.10-2 10 DRR 02-0122 2/28/02 B 3.1.10-3 13 DRR 02-1063 9/26/02 B 3.1.10-4 17 DRR 04-0452 5/26/04 TAB -B 3.3 INSTRUMENTATION B 3.3.3-1 2 B 3.3.3-2 2 B 3.3.3-3 2 B 3.3.3-4 2 B 3.3.3-5 2 B 3.3.3-6 2 B 3.3.9-1 B 3.3.9-2 B 3.3.9-3 B 3.3.9-4 B 3.3.10-1 3 3 5 3 8 8 8 DRR 05-0708 DRR 05-1996 DRR 05-0708 DRR 06-1349 DRR 06-1349 DRR 06-1349 DRR 01-0475 DRR 99-1581 DRR 01-0475 DRR 01-0475 DRR 99-1581 4/20/05 9/28/05 4/20/05 7/24/06 7/24/06 7/24/06 5/1/01 12/18/99 5/1/01 5/1/01 12/18/99 Wolf Creek- Unit 1 Revision 33 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL BASES PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.3 INSTRUMENTATION (continued)

B 3.3.10-2 3 B 3.3.10-3 3 B 3.3.11-1 3 B 3.3.11-2 3 B 3.3.11-3 3 B 3.3.11-4 3 B 3.3.11-5 3 B 3.3.12-1 3 B 3.3.12-2 19 B 3.3.12-3 19 B 3.3.13-1 23 B 3.3.13-2 3 B 3.3.13-3 3 B 3.3.13-4 3 B 3.3.14-1 3 B 3.3.14-2 3 B 3.3.14-3 17 B 3.3.14-4 3 B 3.3.15-1 3 B 3.3.15-2 3 B 3.3.15-3 3 B 3.3.15-4 3 B 3.3.16-1 3 B 3.3.16-2 3 B 3.3.16-3 3 B 3.3.16-4 8 B 3.3.16-5 31 B 3.3.17-1 19 B 3.3.17-2 3 B 3.3.17-3 11 B 3.3.17-4 11 B3.3.17-5 3 B 3.3.18-1 14 B 3.3.18-2 14 B 3.3.18-3 26 B 3.3.18-4 26 B 3.3.18-5 26 B 3.3.18-6 26 B 3.3.18-7 26 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 04-1019 DRR 04-1019 DRR 05-0708 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 04-0452 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 99-1581 DRR 01-0475 DRR 07-0657 DRR 04-1019 DRR 99-1581 DRR 02-0412 DRR 02-0412 DRR 99-1581 DRR 02-1459 DRR 02-1459 DRR 06-0050 DRR 06-0050 DRR 06-0050 DRR 06-0050 DRR 06-0050 12/18/99 12/18/99 12/18/99 12/18/99 12/18/99 12/18/99 12/18/99 12/18/99 9/1/04 9/1/04 4/20/05 12/18/99 12/18/99 12/18/99 12/18/99 12/18/99 5/26/04 12/18/99 12/18/99 12/18/99 12/18/99 12/18i99 12/18/99 12/18/99 12/18/99 5/1/01 5/1/07 9/1/04 12/18/99 4/5/02 4/5/02 12/18/99 2/12/03 2/12/03 2/28/06 2/28/06 2/28/06 2/28/06 2/28/06 TAB -B 3.4 REACTOR COOLANT SYSTEM (RCS)B 3.4.3-1 10 B 3.4.3-2 10 B 3.4.3-3 10 B 3.4.3-4 31 B 3.4.10-1 4 B 3.4.10-2 13 B 3.4.16-1 3 B 3.4.16-2 33 B 3.4.16-3 33 B 3.4.17-1 3 DRR 02-0122 DRR 02-0122 DRR 02-0122 DRR 007-0657 DRR 00-0148 DRR 02-1063 DRR 99-1581 DRR 07-1554 DRR 07-1554 DRR 99-1581 2/28/02 2/28/02 2/28/02 5/1/07 4/24/00 9/26/02 12/18/99 9/28/07 9/28/07 12/18/99 Wolf Creek- Unit 1 ii Revision 33 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL BASES PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.4 REACTOR COOLANT SYSTEM (RCS) (continued)

B 3.4.17-2 33 DRR 07-1554 9/28/07 B 3.4.17-3 33 DRR 07-1554 9/28/07 B 3.4.17-4 33 DRR 07-1554 9/28/07 B 3.4.17-5 33 DRR 07-1554 9/28/07 B 3.4.17-6 33 DRR 07-1554 9/28/07 B 3.4.17-7 33 DRR 07-1554 9/28/07 B 3.4.17-8 33 DRR 07-1554 9/28/07 B 3.4.17-9 33 DRR 07-1554 9/28/07 B 3.4.18-1 3 DRR 99-1581 12/18/99 B 3.4.18-2 3 DRR 99-1581 12/18/99 B 3.4.18-3 3 DRR 99-1581 12/18/99 TAB -B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)B 3.5.1-1 3 DRR 99-1581 12/18/99 B 3.5.1-2 3 DRR 99-1581 12/18/99 B 3.5.1-3 3 DRR 99-1581 12/18/99 TAB -B 3.6 CONTAINMENT SYSTEMS B 3.6.1-1 26 DRR 06-0050 2/28/06 B 3.6.1-2 17 DRR 04-0452 5/26/04 B 3.6.1-3 17 DRR 04-0452 5/26/04 B 3.6.1-4 17 DRR 04-0452 5/26/04 B 3.6.1-5 17 DRR 04-0452 5/26/04 B 3.6.1-6 17 DRR 04-0452 5/26/04 B 3.6.1-7 26 DRR 06-0050 2/28/06 TAB -B 3.7 PLANT SYSTEMS B 3.7.7-1 B 3.7.7-2 B 3.7.7-3 B 3.7.7-4 B 3.7.8-1 B 3.7.8-2 B 3.7.8-3 B 3.7.13-1 B 3.7.13-2 B 3.7.13-3 B 3.7.17-1 B 3.7.17-2 B 3.7.17-3 B 3.7.19-1 B 3.7.19-2 B 3.7.19-3 B 3.7.20-1 B 3.7.20-2 B 3.7.20-3 B 3.7.20-4 B 3.7.20-5 B 3.7.20-6 B 3.7.20-7 8 8 28 8 25 25 25 15 5 26 26 26 26 5 5 5 3 3 32 32 32 32 32 DRR 01-0475 DRR 01-0475 DRR 06-1349 DRR 01-0475 DRR 05-1996 DRR 05-1996 DRR 05-1996 DRR 03-0861 DRR 00-0958 DRR 06-0050 DRR 06-0050 DRR 06-0050 DRR 06-0050 DRR 00-0958 DRR 00-0958 DRR 00-0958 DRR 99-1581 DRR 99-1581 DRR 07-1058 DRR 07-1058 DRR 07-1058 DRR 07-1058 DRR 07-1058 5/1/01 5/1/01 7/24/06 5/1/01 9/28/05 9/28/05 9/28/05 7/10/03 8/17/00 2/28/06 2/28/06 2/28/06 2/28/06 8/17/00 8/17/00 12/18/99 12/18/99 12/18/99 7/10/07 7/10/07 7/10/07 7/10/07 7/10/07 Wolf Creek -UnIt 1 iii Revision 33 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL BASES PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)TAB -B 3.7 PLANT SYSTEMS (continued)

B 3.7.20-8 32 DRR 07-1058 7/10/07 B 3.7.20-9 32 DRR 07-1058 7/10/07 B 3.7.21-1 3 DRR 99-1581 12/18/99 B 3.7.21-2 3 DRR 99-1581 12/18/99 B 3.7.21-3 3 DRR 99-1581 12/18/99 B 3.7.21-4 3 DRR 99-1581 12/18/99 B 3.7.22-1 3 DRR 99-1581 12/18/99 B 3.7.22-2 9 DRR 01-1236 9/19/01 B 3.7.22-3 3 DRR 99-1581 12/18/99 B 3.7.22-4 3 DRR 99-1581 12/18/99 B 3.7.23-1 21 DRR 04-1535 12/1/04 B 3.7.23-2 21 DRR 04-1535 12/1/04 B 3.7.23-3 21 DRR 04-1535 12/1/04 B 3.7.23-4 21 DRR 04-1535 12/1/04 B 3.7.23-5 21 DRR 04-1535 12/1/04 B 3.7.24-1 21 DRR 04-1535 12/1/04 B 3.7.24-2 21 DRR 04-1535 12/1/04 B 3.7.24-3 21 DRR 04-1535 12/1/04 B 3.7.24-4 21 DRR 04-1535 12/1/04 B 3.7.24.5 21 DRR 04-1535 12/1/04 B 3.7.24-6 21 DRR 04-1535 12/1/04 TAB -B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.11-1 3 DRR 99-1581 12/18/99 B 3.8.11-2 27 DRR 06-0799 5/18/06 B 3.8.11-3 27 DRR 06-0799 5/18/06 B 3.8.11-4 27 DRR 06-0799 5/18/06 B 3.8.11-5 27 DRR 06-0799 5/18/06 B 3.8.11-6 27 DRR 06-0799 5/18/06 TAB -B 3.9 REFUELING OPERATIONS B 3.9.7-1 3 DRR 99-1581 12/18/99 B 3.9.7-2 3 DRR 99-1581 12/18/99 B 3.9.9-1 3 DRR 99-1581 12/18/99 B 3.9.9-2 3 DRR 99-1581 12/18/99 B 3.9.9-3 17 DRR 04-0452 5/26/04 TAB -B 3.10 EXPLOSIVE GAS AND STORAGE TANK RADIOACTIVITY MONITORING B 3.10.1-1 3 DRR 99-1581 12/18/99 B 3.10.1-2 3 DRR 99-1581 12/18/99 B 3.10.1-3 3 DRR 99-1581 12/18/99 B 3.10.2-1 3 DRR 99-1581 12/18/99 B 3.10.2-2 3 DRR 99-1581 12/18/99 B 3.10.2-3 3 DRR 99-1581 12/18/99 B 3.10.3-1 3 DRR 99-1581 12/18/99 B 3.10.3-2 3 DRR 99-1581 12/18/99 B 3.10.3-3 3 DRR 99-1581 12/18/99 B 3.10.3-4 3 DRR 99-1581 12/18/99 Wolf Creek- UnIt 1 iv Revision 33 LIST OF EFFECTIVE PAGES -TECHNICAL REQUIREMENT MANUAL BASES PAGE (1) REVISION NO. (2) CHANGE DOCUMENT (3) DATE EFFECTIVE/

IMPLEMENTED (4)Note 1 The page number is listed on the center of the bottom of each page.Note 2 The revision number is listed in the lower right hand corner of each page. The Revision number will be page specific.Note 3 The change document will be the document requesting the change. Therefore, the change document should be a DRR number in accordance with AP 26A-002.Note 4 The date effective or implemented is the date the Technical Requirement Bases pages are issued by Document Control.Wolf Creek- UnIt 1 V Revision 33

WM 08-0003, Wolf Creek Transmittal of Revision 21 to Updated Safety Analysis Report

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