L-96-006, Response to RAI Regarding License Amendment Request for Low Pressure Service Water Reactor Building Waterhammer Prevention System Modification to Mitigate Waterhammers Described in Generic Letter 96-06 and Associated Technical..

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Response to RAI Regarding License Amendment Request for Low Pressure Service Water Reactor Building Waterhammer Prevention System Modification to Mitigate Waterhammers Described in Generic Letter 96-06 and Associated Technical..
ML081330241
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 05/07/2008
From: Baxter D
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
GL-96-006
Download: ML081330241 (42)
v • d • e


Text

{{#Wiki_filter:Iuke DAVE BAXTER Vice President Energy Oconee Nuclear Station Duke Energy Corporation ON01 VP17800 Rochester Highway Seneca, SC 29672 

                                                                                 .864-885-4460 864-885-4208 fax dabaxter@dukeenergy.com May 7, 2008 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

Subject:

Duke Energy Carolinas, LLC Oconee Nuclear Site, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 License Amendment Request for Low Pressure Service Water Reactor Building. Waterhammer Prevention System Modification to Mitigate Waterhammers Described in Generic Letter 96-06 and Associated Technical Specifications Request for Additional Information License Amendment Request (LAR) No. 2006-05 In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke) proposes to amend Renewed Facility Operating Licenses Nos. DPR-38, DPR-47, and DPR-55. A LAR was submitted on October 16, 2007 to the Nuclear Regulatory Commission (NRC) seeking review and approval of a plant modification that addresses waterhammer concerns described in Generic Letter (GL) 96-06. The modification will install check valves in the Low Pressure Service Water (LPSW) supply header and automatic pneumatic discharge isolation valves, controllable vacuum breaker valves, and associated circuitry in the LPSW return header to isolate Engineered Safeguards (ES) portions of the LPSW system to mitigate waterhammers. The affected LPSW piping is located inside the containment, the turbine building, and the auxiliary building and provides cooling to the Reactor Building Cooling Units (RBCUs), Reactor Building Auxiliary Coolers (RBACs) and the Reactor Coolant Pump Motor (RCPM) Coolers. This request also proposes Technical Specifications (TS) and associated bases in support of maintaining the ES portions (Containment Heat Removal Function) of the system. Duke met with the NRC on January 24, 2008 to discuss the submittal. In an email dated March 12, 2008, Duke received requests for additional information (RAIs). Enclosure 2 contains Duke's responses to those RAIs. In a phone call on April 9, 2008, Duke was asked to revise a minor error in TS 3.3.27. Attachments 2 and 3 contain that correction. In accordance with Duke administrative procedures and the Quality Assurance Program Topical Report, these proposed changes are still bounded by the review and approval of the Plant 4o72-www. duke-energy.com

Nuclear Regulatory Commission License Amendment Request No. 2006-05 May 7, 2008 Page 2 Operations Review Committee and Nuclear Safety Review Board. Additionally, a copy of this LAR is being sent to the State of South Carolina in accordance with 10 CFR 50.91 requirements. Implementation dates for the Waterhammer Prevention modifications were committed to in a letter to the NRC dated February 14, 2007. To support the commitment dates specified, Duke requests that this amendment be issued by July, 2008, effective upon issuance, with modification implementation to start with Unit 2 startup from the fall 2008 outage and continue through the outages which follow for Units I and 3 in the fall and spring of 2009 respectively. Notes included in the proposed Technical Specifications control the applicability for these Units prior to the modifications being installed and can be removed or modified after the modifications have been completed on all three Oconee Units. Duke will also update applicable sections of the Oconee Updated Final Safety Analysis Report and submit these changes per 10 CFR 50.71(e). There are no new commitments being made as a result of this letter. Inquiries on this proposed amendment request should be directed to Reene' Gambrell of the. Oconee Regulatory Compliance Group at (864) 885-3364. Sincerely, Day Baxter, Vice President Oconee Nuclear Site

Enclosures:

1. Notarized Affidavit
2. Requests For Additional Information Attachments:
1. Reactor Building Auxiliary Coolers LPSW Header Pressure LOOP Accuracy
2. Technical Specifications - Mark Up
3. Technical Specifications - Reprinted Page

Nuclear Regulatory Commission License Amendment Request No. 2006-05 May 7, 2008 Page 3 bc w/enclosures and attachments: Mr. Vic McCree, Regional Administrator (Acting) U. S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. L. N. Olshan, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, D. C. 20555 Mr. Andy Hutto Senior Resident Inspector Oconee Nuclear Site Mrs. Susan E. Jenkins, Manager Infectious and Radioactive Waste Management Section Department of Health & Environmental Control 2600 Bull Street Columbia, SC 29201

Nuclear RegulatoryCommission License Amendment Request No. 2006-05 May 7, 2008 Page 4 bcc w/enclosures and attachments: B. G. Davenport R. V. Gambrell D. J. Williams P. J. Earnhardt H. E. Harling V. B. Bowman T. N. Glenn L. F. Vaughn C. E. Curry C. G. Abellana J. E. Burchfield L. T. Harbinson R. L. Gill - NRI&IA R. D. Hart - CNS K. L. Ashe - MNS NSRB, EC05N ELL, ECO50 File - T.S. Working ONS Document Management

ENCLOSURE 1 NOTARIZED AFFIDAVIT

Enclosure I - Notarized Affidavit License Amendment Request No. 2006-05 May 7, 2008 Page 1 AFFIDAVIT Dave Baxter, being duly sworn, states that he is Vice President, Oconee Nuclear Site, Duke Energy Carolinas, LLC that he is authorized on the part of said Company to sign and file with the U. S. Nuclear Regulatory Commission this revision to the Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55; and that all statements and matters set forth herein are true and correct to the best of his knowledge. Dave Baxter.V ice President Oconee Nuclear Site Subscribed and sworn to before me this ' day of 2008 J-4jý- ,.--I J-)'a4ý11- w Notary Public "My Corommission Expires: Date SEAL

ENCLOSURE 2 REQUESTS FOR ADDITIONAL INFORMATION 

 - Requests For Additional Information License Amendment Request No. 2006-05 May 7, 2008                                                                          Page 1 1.0      REQUESTS FOR ADDITIONAL INFORMATION RAI #1                                    /"

Confirm that the modifications (including those that have already been completed) that are necessary for resolving the Generic Letter 96-06 waterhammer issues as described in letters dated September 30, 2002, March 24, 2003, and October 16, 2007 are in accordance with 10 CFR Part 50, Appendix B requirements. ANSWER: Duke's QA program, as defined in Chapter 3 of Oconee's Updated Final Safety Analysis Report (UFSAR), conforms to 10 CFR Part 50, Appendix B. This is defined as QA Condition 1 for nuclear safety related systems, structures, or components (SSC). The modifications addressed in the letters above were designed and have been or will be installed in accordance with 10 CFR Part 50, Appendix B requirements. All major components were procured in accordance with Duke's Nuclear Procurement Program as QA Condition 1 items. RAI #2 Define the amount of check valve back leakage and other boundary valve leakage that is permitted consistent with not having to consider a waterhammer event. Describe the periodic testing requirements that will be included in the in-service testing program to ensure that the boundary valve leakage conditions are not exceeded. ANSWER: The current analysis allows for up to 25 gpm of boundary valve leakage. Of this 25 gpm, 20 gpm is permitted for the aggregate of boundary valve leakage and 5 gpm reserved for miscellaneous, unspecified leakage (e.g., flange leakage). This allocation of the 25 gpm is subject to change as experience is gained with the system. For a Loss Of Coolant Accident, the heat input from containment is sufficient to expand the water and prevent voiding with this amount of leakage. During events which do not add heat to containment (e.g., Loss Of Offsite Power only), the allowed leakage rate from the LPSW RB Waterhammer Prevention System boundary valves will be very low. In order to allow more leakage, an accumulator is being added that will add water to the system to make-up for valve leakage. To prevent voiding, this accumulator has been designed to supply 25 gpm for one minute. 

 - Requests For Additional Information License Amendment Request No. 2006-05 May 7, 2008                                                                           Page 2 Periodic testing will be performed to measure the leakage of the boundary valves and ensure that.

the aggregate of that leakage is acceptable. Essentially, the allowed leakage is the sum of the check valves and the air operated valves in series with the greatest leakage. Any changes in these parameters will be evaluated and controlled through normal station processes. RAI #3 Confirm that the previously performed thermal overpressurization resolution remains valid after implementation of the proposed changes for resolving waterhammer issues. ANSWER: The new LPSW RB Waterhammer Prevention System (WPS) affects penetrations (30, 31, 32, 33, 34, 35) that were previously evaluated as being normally open. Since the WPS will momentarily isolate the penetration, thermal relief valves were added to prevent thermal overpressure when the boundary valves are closed. Once the boundary valves reopen following restart of the LPSW pumps, the penetrations will be in the state previously evaluated. No other penetrations are affected by this modification. RAI #4 Provide Setpoint analysis reference in SR 3.3.27.3. Answer: The setpoint analysis is supplied in attachment 1. This calculation is being submitted to aid in the staff's review and approval of this LAR. However, this calculation may be subject to change and will be evaluated and controlled through normal station processes.

ATTACHMENT 1 REACTOR BUILDING AUXILIARY COOLERS LPSW HEADER PRESSURE LOOP ACCURACY

FIGURE 101 3 CERTIFICATION OF ENGINEERING CALCULATION - REVISION LOG ri CERTIFICATION OF ENGINEERING CALCULATION. REVISION LOG Station And Unit Number Oconee Nuclear Title Of Calculation Reactor Buildine Auxiliary Coolers LPSW Header Pressure Loon Accuracy. Type III Calc, Calculation Number OSC-8538 NOTE 1: When approving a Calculation revision with multiple Originators or Checkers, the Approver need sign only one block. r-(15 MAR 2000)

FIGURE 101 ICERTIFICATION OF ENGINEERING CALCULATION CERTIFICATION OF ENGINEERING CALCULATION Station And Unit Number Oconee Nuclear Title Of Calculation Reactor Building Auxiliary Coolers LPSW Header Pressure Loop Accuracy. Type [II Calc. Calculation Number OSC-8538 Total Original Pages "). ii Through 18 Total Supporting Documentation Attachments 5 Total Microfiche Attachments 0 Total Volumes 0 Type I Calculation/Analysis El YES W] NO Microfiche Attachment List ElYESW~No SEEFFORM 101.4 SEE FORM 10 1 4 Microfiche Attachment List These engineering Calculations cover QA Condition 1 Items. In accordance with established procedures, the quality has been assured and I certify that the above Calculation has been Originated, Checked, or Approved as noted below: Originated By Date 12117103 Checked By Date Verification Method: Method 3 D Other Approved By - Date iý1 Issued To Document Management ,1&x24 01Q 1/14Cl( Received By Document Management Comp lete The Spaces Below For Documentation Of Multiple Originators Or Checkers I IPages Through 1Originated By Date lChecked By Date Verification Method: Method IF._ Method ..2 1 F----------- Method_3 . . Other------- LFI-. IPages Through I 1Originated By Date I lChecked By Date Verification Method: Method 1 7. Method 2 ... Method 3 . . Other ELi IPages Through I 1Originated By Date I lChecked By Date I 

    ,Verification Method:                Method 1I*

Method 2 F- Method 3 7I Other El (01 MAR 2003)

Form 101.2 (R3-03) Calculation Number OSC-8538 REVISION DOCUMENTATION SHEET Revision Revision Description Number 0 Original issue for NSM 23107. 1 Revised for NSM 33107. Revised Table of Contents page 1 for page numbers & added item 1.5 to page 1, Revised 5.0 H thru 5.0 L & 5.0 T to add unit 3 document numbers, added NSM 33107 to 6.1.2, Revised 7.2.1 Current Leakage (PTCL) for unit 3 cable, Revised 7.3.1 for new PTCL value, Revised 7.4 & 8.0 for new HELB error. 2 Revised for NSM 13107. Revised Table of Contents page 1,item 7.3 for page number, Revised 5.0 H thru 5.0 L & 5.0 T to add unit 1 document numbers, added NSM 13107 to 6.1.2, Revised 6.1.5 to reflect implementation of transmitter head correction, Added unit 1 data to Figure 7.1-1, Revised 7.2.1 Current Leakage (PTCL) for unit 1 cable, Revised 7.4 to reflect 23 psig setpoint. Revised per NSM-ON-13117, NSM-ON-23117 and NSM-ON-33117 revised Table of Contents Page 1, revised Section 1.2 to include RB LPSW discharge valves control loop, revised section 1.3 to include Reactor Building Isolation circuit, revised Section 2 to include QA condition of added functions of the control loop, revised section 3 to delete single sided reduction factor dialog, revised Section 4.2 to delete Tech Spec Amendment for RBAC and add Tech Spec Amendment for Reactor Building discharge Isolation to be submitted to the NRC, revised Section 5 to add references. Revised Section 6 to reflect the new set points, revised figure 7.1-1 to reflect Zero reference elevation for all affected transmitters on all three Units, revised figure 7.1-2 title description, revised section 7.2.2 to clarify why dead band and environmental allowance was not applicable, revised section 7.2.3 to clarify why process measurement allowance was not applicable, revised section 7.3 and 7.4 to delete single sided reduction factor, revised Section 7.4 to describe Trip Set Point Analysis and Reset Set Point Analysis, revised Section 8 to reflect acceptability of total loop uncertainty of Reactor Building discharge valve control loop. Revised set point to close valves _LPSW-1 121, _LPSW-1 122, _LPSW-1 123, LPSW-1 124 at 18.5 psig decreasing +/- 0.5 psig and open valves _LPSW-1150 and _LPSW-1151 at 18.5 psig decreasing +/- 0.5 psig andto re-open valves _LPSW-1121, _LPSW-1122, _LPSW-1123, LPSW-1 124 at 25 psig increasing +/- I psig and re-close valves _LPSW-1150 and LPSW-1 151 at 25 psig increasing +/- 1 psig per page 23 of Rev.2 of the Mechanical Design Input calculation for NSM-X3117 i.e,, calculation OSC-8144 Rev. 2 I i i (15 MAR 2000)

Attachment A Engineering Manual 4.9 Revision 3 CALCULATION IMPACT ASSESSMENT (CIA) 

"tation / Unit           Oconee     /  1,2,3         Calculation No.       OSC-8538                   Rev. 0         Page      i PIP No. (if applicable)               N/A                         By        B J Shepherd                    Date /;Zf7/ 0 3 Prob. No. (stress & s/r use only)                 I          Checked By                                     Date    0 ,405 Note: A NEDL search is NOT required for NEDL reviewed to identify calculations?                  El    YES       W    NO        calculation originations (i.e. Rev. O's)
       . .. ii C'.

CA rI)"*Q ernially bocksb) Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision. Indiv. Contacted/Date Indiv. Contacted/Date RES [] NGO (Power, I&C, ERRT, (QA Tech. Services (IS1), Reactor) Severe Accident Analysis,Elect. Sys. & Equip., Design & Reactor EZMCE Supp., Civil Structural, Core Mech. & T/H Analysis, Mech. (Primary Systems, Balance of Plant, Rotating Equipment, Sys. & Equip., Nuclear Design Valves & Heat Exchangers, and Safety Analysis, Matls/Metallurgy/Piping) Civil) MOD HE Harling 12-17-03 (Mechanical Engr., Electrical ED Training Engr., Civil Engr.) El Operations - El Local IT OPS Support El Regulatory Compliance Maintenance - Tech. Support E] Chemistry L] Work Control - Program. Supp. D- Radiation Protection 0 Other Group El No Group required to be consulted Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATION PROCEDURES', DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document ownerlgroup and the change required (including any necesssary PIP Corrective Actions).

  • Note: Any design changes, which require changes to Station Procedures, must be transmitted as Design Deliverable Documents.

DOCUMENT GROUP CHANGE REQUIRED IP/O/A/0250/0010 RES Tech Spec 3.3.29 MOD ONTC-2-124B-0023-001 MOD Page 1 of 1

Attachment A Engineering Manual 4.9 Revision 3 CALCIJLATION IMPACT ASSESSMENT (CIA) Station / Unit Oconee / 1,2,3 Calculation No. OSC-8538 Rev. I Page ii PIP No. (if applicable) N/A By B J Shepherd .JY7W,&-,Date ,2 p4, q' FTrob. No. se-& sir use only) Checked By ,, , Date "11-71o0 Note: A NEDL search is NOT required for NEDL reviewed to identify calculations? El YES W NO calculation originations (i.e. Rev. O's) Idormany in thts 3 Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision. Indiv. Contacted/Date Indiv. Contacted/Date MWJRES El NGO (Power, I&C, ERRT, Warren Bright, 2/12/04 (QA Tech. Services (ISI), Reactor) Severe Accident Analysis,Elect. Sys. & Equip., Design & Reactor MCE Supp., Civil Structural, Core Scott Manning, 2/12/04 (Primary Systems, Balance of Mech. & T/H Analysis, Mech. Plant, Rotating Equipment, Sys. & Equip., Nuclear Design Valves & Heat Exchangers, and Safety Analysis, Civil) Matls/MetallurgyPiping) MOD HE Harling, 2/16/04 (Mechanical Engr., Electrical Ei Training Engr., Civil Engr.) El Operations - Li Local IT OPS Support Li Regulatory Compliance Li Maintenance - Tech. Support Di Chemistry Li Work Control - Program. Supp. Di Radiation Protection E: Other Group Li No Group required to be consulted Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATION PROCEDURES*, DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document ownerlgroup and the change required (including any necesssary PIP Corrective Actions).

  • Note: Any design changes, which require changes to Station Procedures, must be transmitted as Design Deliverable Documents.

DOCUMENT GROUP CHANGE REQUIRED IP/O/A/0250/O010 RES Tech Spec 3.3.29 MOD ONTC-2-124B-0023-001 MOD ONTC-3-124B-0023-001 MOD Page 1 of 1

Attachment A Engineering Manual 4.9 Revision 3 CALCULATION IMPACT ASSESSMENT (CIA) Station / Unit Oconee / 1,2,3 Calculation No. OSC-8538 Rev. 2 Page iii ?IP No. (if applicable) N/A By B J Shepherd D-ffae /l O-Prob. No. (stress & s/r use only) Checked By 0-1, S'*fWC-' Date 0-'-1 O Y Note: A NEDL search is NOT required for NEDL reviewed to identify calculations? YES N calculation originations (i.e. Rev. O's) (formally SAROS) Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision. Indiv. Contacted/Date Indiv. Contacted/Date RES II NGO (Power, I&C, ERRT, Warren Bright, 9/21/04 (QA Tech. Services (ISI), Reactor) Severe Accident Analysis,Elect. Sys. & Equip., Design & Reactor MCE [] Scott Manning,'*- _,O ¢-? Supp., Civil Structural, Core (Primary Systems, Balance of Mech. & T/H Analysis, Mech. Plant, Rotating Equipment, Sys. & Equip., Nuclear Design Valves & Heat Exchangers, and Safety Analysis, Civil) Matls/Metallurgy/Piping) MOD HE Harling, *46M4- %04- '/o-(Mechanical Engr., Electrical D Training Engr., Civil Engr.) D Operations- D Local IT OPS Support D Regulatory Compliance D] Maintenance - Tech. Support D Chemistry Work Control - Program. Supp. D Radiation Protection D] Other Group _______________ Er No Group required to be consulted Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATION PROCEDURES*, DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document owner/group and the change required (including any necesssary PIP Corrective Actions).

  • Note: Any design changes, which require changes to Station Procedures, must be transmitted as Design Deliverable Documents.

DOCUMENT GROUP CHANGE REQUIRED IP/O/A/0250/0010 RES Tech Spec 3.3.29 MOD ONTC- 1-124B-0023-001 MOD ONTC-2-124B-0023-001 MOD ONTC-3-124B-0023-001 MOD Page 1 of 1

Engineering Manual 4.9 CALCULATION IMPACT ASSESSMENT (CIA) Station / Unit Oconee / Calculation No. OSC-8538 Rev. 3 Page iv PIP No. (if applicable) By Fred Custer Date 7-- a C' C Prob. No. (stress & s/r use only) Checked By zTfi b Date 1i'-0(j Note: A NEDL search is NOT required for NEDL reviewed to identify calculations? El YES W] NO calculation originations (i.e. Rev. O's) (formally SAROS) Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision. Indiv. Contacted/Date Indiv. Contacted/Date RES [] NGO (Power, I&C, ERRT, (QA Tech. Services (ISI), Reactor) Severe Accident Analysis,Elect. Sys. & Equip., Design &-Reactor LI MCE Supp., Civil Structural, Core (Primary Systems, Balance of Mech. & T/H Analysis, Mech. Plant, Rotating Equipment, Sys. & Equip., Nuclear Design Valves & Heat Exchangers, and Safety Analysis, Matls/Metallurgy/Pipinng Civil) MOD H.E. Harling 02-06-06 (Mechanical Engr., Electrical El Training Engr., Civil Engr.) El Operations - LI Local IT OPS.Support El Regulatory Compliance Maintenance - Tech. Support [0] Chemistry El Work Control - Program. Supp. LI Radiation Protection LI Other Group LI No Group required to be consulted Listed below are~the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATION PROCEDURES*, DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document owner/group and the change required (including any necesssary PIP Corrective Actions).

  • Note: Any design changes, which require changes to Station Procedures, must be transmitted as Design Deliverable Documents.

DOCUMENT' GROUP CHANGE REQUIRED [P/A/0250/0010 RES Tech Spec 3.3.27 & B 3.3.27 MOD ONTC-1-124B-023-001 MOD ONTC-2-124B-023-001 . MOD ONTC-3-124B-023-001 MOD Page 1 of 1

Engineering Manual 4.9 CALCULATION IMPACT ASSESSMENT (CIA) Station / Unit Oconee / Calculation No. OSC-8538 Rev. 4 Page v PIP No. (if applicable) By Fred Custer Date 5/21/2007 Prob. No. (stress & s/r use only) Checked By ,,-4L*"-- Date 6 1.2 -4 7 Note: A NEDL search is NOT required for NEDL reviewed to identify calculations? El YES W] NO calculation originations (i.e. Rev. O's) (formally SAROS) Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision. Indiv. Contacted/Date Indiv. Contacted/Date El RES 11 NGO (Power, I&C, ERRT, (QA Tech. Services (ISI), Reactor) Severe Accident Analysis,Elect. Sys. & Equip., Design &-Reactor NI MCE Supp., Civil Structural, Core (Primary Systems, Balance of Mech. & T/H Analysis, Mech. Plant, Rotating Equipment, Sys. & Equip., Nuclear Design Valves & Heat Exchangers, and Safety Analysis, Civil) Matls/Metallurgy/Pipinng MOD Henry Harling (Mechanical Engr., Electrical El Training Engr., Civil Engr.) El Operations - El Local IT OPS Support Li Regulatory Compliance Ei Maintenance - Tech. Support Li Chemistry I1 Work Control - Program. Supp. El Radiation Protection Li Other Group Li No Group required to be consulted Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATION PROCEDURES*, DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document owner/group and the change required (including any necesssary PIP Corrective Actions).

  • Note: Any design changes, which require changes to Station Procedures, must be transmitted as Design Deliverable Documents.

DOCUMENT GROUP CHANGE REQUIRED OSC-8144 Rev. I Mod Mechanical OSC-8144 Rev.2 required to be issued to account for required setpoint changes Page 1 of 1

I OSC-8538 Rev. 4 Page 1 TABLE OF CONTENTS Section Paize Number Calculatiion Impact Assessment Forms i thru v 1.0 STATEMENT OF PROBLEM/PURPOSE 3 1.1 Purpose 3 1.2 Analyzed Loop Function 3 1.3 Plant Conditions Requiring Operation 3 1.4 Location and Applicable Environment 4 1.5 East Penetration Room HELB Temperature 4 2.0 RELATION TO QA CONDITION/NUCLEAR SAFETY 4 3.0 DESIGN CALCULATION METHOD 4 4.0 FSAR/TECHNICAL SPECIFICATION APPLICABILITY 5

5.0 REFERENCES

5 6.0 ASSUMPTIONS/DESIGN INPUT 9 6.1 Assumptions 9 6.2 Design Input/Bases 9 7.0 CALCULATION 9 7.1 Instrument Block Diagram 10 7.2 Device/Loop Uncertainty Term Determination 11 7.3 Total Loop Uncertainty Determination 15 7.4 Setpoint Analyses and/or Acceptability of Loop Indication Uncertainty 17 7.5 Values to Aid in Determining Loop Past Operability 19

8.0 CONCLUSION

S/RESULTS 20 Attachments Number of Pages 1 - Rosemount 1154 transmitter Data Sheet 00813-0100-4514 dated June 1999 Rev.AA 9 2 - Rochester Instruments ET-1215 Current 

                   & Voltage Alarm module Data Sheet                                4 3 -    Rochester Instruments SC-1300 Voltage/

Current Transmitter 2 4 - Rosemount ATM-00049.15-0401 Power Supply Data Sheet 2

I OSC-8538 Rev. 4 Page 2 Attachments Number of Pages 5 - NRC Docket Letter to Jim Hampton Dated Dec. 5, 1994 2

OSC-8538 Rev. 4 Page 3 1.0 STATEMENT OF PROBLEM/PURPOSE 1.1 Purpose The purpose of this calculation is to determine the uncertainties for the Oconee Units 1, 2 and 3 LPSW Supply to Reactor Building Coolers pressure transmitters and current switches. The applicable tag numbers listed below were taken from Reference 5.1 and 5.M. Units 1, 2, & 3 Transmitters: _LPSPT0010 LPSPT00I 1 _LPSPT0012 _LPSPTOO13 Current switches: _LPSIS0010 _LPSIS0011 _LPSIS0012 _LPSIS0013 The calculation further identifies specific calibration requirements which were used as input for the loop uncertainty determination and determines a loop "as-found" tolerance. 1.2 Analyzed Loop Function The subject pressure transmitters monitor LPSW supply pressure to the Reactor Building Coolers and provide input to current switches which actuate the Reactor Building LPSW Isolation circuit when LPSW supply pressure reaches a decreasing setpoint of 18.5 psig. The function of the Reactor Building LPSW Isolation circuit is to close the LPSW RBAC Supply and Return Isolation Valves (_LPSW-1054, -1055, -1061, and -1062) and Reactor Building LPSW Discharge Valves (_LPSW-1121, -1122, -1123, and -1124) and open LPSW vent valves _LPSW-1150 and _LPSW1151 on decrease of LPSW discharge pressure such as occurs during a LOOP to prevent a water hammer in the LPSW piping and reopen (_LPSW- 1121, -1122, -1123, and -1124) and reclose vent valves _LPSW-1 150 and _LPSW- 1151 when LPSW pressure is restored to normal to allow the LPSW system to perform its Engineered Safeguards Reactor Building (Containment) heat removal function.. The pressure transmitters monitor LPSW to Reactor Building supply pressure over a range from 0 to 100 psig. The Reactor Building LPSW Isolation circuit is not required to mitigate a HELB in the East Penetration Room. However, the instrumentation is designed to operate during and after a HELB and will provide isolation of the Reactor Building during this event if the LPSW pumps trip. 1.3 Plant Conditions Requiring Operation The LPSW supply to the pressure transmitters and current switches are designed to provide input signals to actuate the Reactor Building LPSW Isolation circuit when LPSW supply pressure reaches a decreasing setpoint. The Reactor Building LPSW Isolation circuit is required for "Modes 1,2,3, or 4 of plant operations when LPSW is not shared with another Unit" and "At all times for all Units sharing LPSW when any sharing unit is in Modes 1,2,3, or 4" of plant operation.

. OSC-8538 Rev. 4 Page 4 1.4. Location and Applicable Environment The applicable environmental parameters are specified below based on the respective instrument locations and functional requirements (from Section 1.3). The transmitters are located in the East Penetration Room and the current switches are located in the Cable Room. TABLE 1.4A EAST PENETRATION ROOM Temperature - Aux Building/ 60-130'F (normal)

  • Radiation - Aux Building/ 1 x 106 Rads (40 yr normal dose) 0 Pressure atmospheric CABLE ROOM
  • Temperature - Aux Building/ 60-1 00°F (normal)
  • Radiation - Aux Building/ less than 1 x 102 Rads (40 yr normal dose)
  • Pressure atmospheric 1.5 East Penetration Room HELB Temperature OSC - 8104 Rev. 1, Appendix A, Figures A-48 & A-49 define the HELB temperature profile for the transmitter mounting location and signal cable routing in the East Pen Room. The signal cable is routed at approximately elev 830 ft, Reference Figure A-49.

The transmitters are mounted at the elevations shown in Fig 7.1-1, Reference Figure A-48. 2.0 RELATION TO QA CONDITION/NUCLEAR SAFETY This calculation was designated a QA Condition 1 Calculation as the LPSW supply header Pressure Transmitters are relied upon to initiate the Reactor Building LPSW Isolation circuit when LPSW supply header decreasing pressure reaches the trip set point. The Reactor Building LPSW Isolation circuit is required to mitigate the possibility of a water hammer occurring in the RBAC's and RBCU's LPSW piping due to a LOOP event. The calculation is also designated QA Condition 1 because the LPSW Reactor Building Discharge isolation circuitry will momentarily isolate the LPSW Engineered Safeguards Containment heat removal flow path from the Reactor Building. 3.0 DESIGN CALCULATION METHOD The methodology employed by this calculation is based on Reference 5.B.a, b and c. The methodology accounts for random-independent (x,y), random-dependent (w,u) and non-random/bias (v,t) uncertainty terms (applied only in an additive manner). The magnitude of each term is combined to determine the "Total Loop Uncertainty" (TLU), as follows:

OSC-8538 Rev. 4 Page 5 

                        +TLU = +-{x2 +     y2 + (w + u) 2
                                                          )} + v + t           [

S[EQUATION 3-11 

                        -TLU    =  -- X2 + y2 + (W + u) 2}11/2 - v  - t J

Specific uncertainty terms and terminology are defined as follows: This calculation provides input to determine the setpoint for the pressure transmitters and current switches that input to the Reactor Building LPSW Isolation circuit. Uncertainty Terms: A - Device/rack Accuracy. PMA - Process measurement allowance. CL - Current leakage. PSE - Device power supply effect. CTE - Calibration tolerance. RES - Resolution/Readability. D - Device drift. SA - Seismic allowance. EA - Environmental allowance SPE - Static pressure effects. MTE - Measuring & test equipment TE - Device/rack temperature effect. PEA - Primary element allowance DB - Deadband. CE - Calibration Effects REP - Repeatability (including M&TE and CTE) Terminology/Abbreviations: AL - Analytical Limit. P]L - Process Limit. AV - Allowable value. S]P - Nominal Set Point. OL - Operating Limit. S]L - Safety Limit. 4.0 FSAR/TECHNICAL SPECIFICATION APPLICABILITY 4.1. Units 1, 2 & 3 Oconee FSAR, Section 9.2.2.2.3 (This section will be revised as part of mod process). 4.2. Technical Specifications Sections 3.3.27 & B 3.3.27 (License Amendment Request for Reactor Building LPSW Isolation circuit to be submitted to NRC)

5.0 REFERENCES

A. Design Basis Specification for the Low Pressure Service Water System, Spec. OSS-0254.00-00-1039, Revisions per NSM ON-23107/AL l & ON-33107/BL 1. B. a) EDM-102: Instrument Setpoint/Uncertainty Calculations, Revision 3. b) ISA-$67.04, Part I, Setpoints for Nuclear Safety-Related Instrumentation, Approved September 1994.

OSC-8538 Rev. 4 Page 6 c) ISA-RP67.04-Part 11-1994, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation, Approved September 1994. C. Oconee Technical Specification 3.3.27, (New'Tech Spec, Amended by NSM ON-X3117. D. Oconee Manual OM-267.0968-001, Instruction Book for Model 1154 Alphaline Pressure Transmitters. E. Oconee Units 1, 2 & 3 Environmental Qualification Criteria Manual, Revision 14. F. Oconee Technical Specification Bases B3.3.27, (New Tech Spec, Amended by NSM ON-X3117. G. Oconee Procedure for Reactor Building LPSW Isolation Circuit Calibration: IP/0/A/0250/0010 (procedure was originated by NSM ON-X3107 and is to be revised per NSM ON-X3117) H. Oconee Isometric Piping & Tubing Layout Drawings: Unit 1 0-439-13107-001 Rev. A 0-439-131,07-003 Rev. A 0-429-H-136-00-1 Rev. A O-429-H-137-00-1 Rev. A O-429-H-138-00-1 Rev. A 0-429-H-139-00-1 Rev. A Unit 2 0-1439-23107-001 Rev. D 0-1439-23107-003 Rev. C 0-1429-H-136-00-1 Rev. A 0-1429-H-137-00-1 Rev. A 0-1429-H-138-00-1 Rev. A 0-1429-H-139-00-1 Rev. A Unit 3 0-2439-33107-001 Rev. B 0-2439-33107-003 Rev. B 0-2429-H-136-00-1 Rev. A 0-2429-H-137-00-1 Rev. A 0-2429-H-138-00-1 Rev. A 0-2429-H-139-00-1 Rev. A I. Oconee Drawing Instrument Detail Drawings:

OSC-8538 Rev. 4 Page 7 Unit 1 0-422H-136, Rev. A 0-422H-137, Rev. A 0-422H-138, Rev. A 0-422H-139, Rev. A Unit 2 0-1422H-136, Rev. A 0-1422H-137, Rev. A 0-1422H-138, Rev. A 0-1422H-139, Rev. A Unit 3 O-2422H-136, Rev. A 0-2422H-137, Rev. A 0-2422H-1138, Rev. A 0-24221-1-139, Rev. A J. Flow Diagrams for the Low Pressure Service Water System: UNIT I OFD- 124B- 1.1, Rev. 44A OFD-124B-1.2, Rev. 22A UNIT 2 OFD-124B-2.1, Rev. 46B UNIT 3 OFD-124B-3.1, Rev. 43A K. Oconee Piping Layout Drawings: UNIT I O-439-A, Rev. 63 0-439-B, Rev. 52 UNIT 2 0-1439-A, Rev. 49 0-1439-B, Rev. 55 UNIT 3 O-2439-A, Rev. 52 0-2439-B, Rev. 35 L. Oconee Instrument Location Drawings:

OSC-8538 Rev. 4 Page 8 UNIT 1 O-422-A-008, Rev.20A UNIT 2 0- 1422-A-008, Rev. 19A UNIT 3 O-2422-A-008, Rev. 16A M. Oconee Equipment Database for Low Pressure Service Water System. N. Oconee Calculation OSC-6085, Revision 2, Measurement & Test Equipment (M&TE) Uncertainties.

0. Justification of the NSD-219 2X OOT Rule for Instrument Loops for which CEN I&C is Responsible for the Uncertainty Analysis, OSC-7165, Revision 3.

P. Oconee Manual OM-267-0969-001, Qualification Report for Pressure Transmitter Model 1154. Q. ASME Steam Tables Sixth Edition. R. Oconee Manual OM-316-0087-001, Qualification Report for Brand Rex Cable S. OSC - 8104 Rev 1, High Energy Line Breaks in the Penetration Room. T. Oconee Elementary Diagrams for Reactor Building LPSW Isolation Circuit Analog Channels 1 thru 4: UNIT 1 OEE-138-55, -56, -57, -58 Rev A UNIT 2

  • OEE-238-55, -56, -57, -58 Rev A UNIT 3 OEE-338-55, -56, -57, -58 Rev A U. Electrical Design Input Caic. OSC-8734 for NSM-ON- 13117,23117, and 33117 I - V. Mechanical Design Input Calc. OSC-8144 rev. 1 for NSM-ON-13117,23117 & 33117

OSC-8538 Rev. 4 Page 9 6.0 ASSUMPTIONS/DESIGN INPUT 6.1 ASSUMPTIONS 6.1.1 Since no drift tolerance is provided by the vendor, drift is assumed to be equal to the Current Switch reference accuracy (repeatability value). 6.1.2 It is assumed that the cable installed by NSM- 13107, NSM-23107 & NSM-33107 (Stock Code 46073, 1SPXJ16G.3) for the LPSW Pressure transmitters will be Brand Rex and that Reference 5.R is valid for this cable. 6.1.3 It is assumed that the IR data for test cable C5120-5B at the 1.6 hour point in the test is valid for the period of To through T 1 .6 HOURS .(i.e. 2.0 x 106 ohms). It is assumed that the cable IR data listed, is a conductor to ground reading. It is also assumed that the cable IR data can be used as a conductor to conductor IR value. 6.1.4 OSC-8104, for a MSLB in the East Pen Room, shows a peak temperature of 527 °F which decreases to 420 °F in 20 seconds and decreases to 390 °F in 100 seconds then reaching an ambient temperature of 120 °F 15 minutes after To. Reference 5.R shows that a temperature of 320 °F is reached 10 seconds after start of testing and that 390 OF is reached 100 seconds after start of testing. The 390 OF is maintained to 15 minutes after start of testing, then decreases to 355 °F for the next 1.1 hours. Since the East Pen Room MSLB temperature drops in 20 seconds to 420 OF and then in another 80 seconds reaches the temperature used in the Reference 5.R testing, it is assumed that the testing performed is valid for the East Pen Room conditions. 6.1.5 The mechanical system calculation OSC-8144 has determined that actuation to close _LPSW-1054, 1055, 1061, 1062, 1121, 1122, 1123, and 1124 between 15 and 24 psig is acceptable but 18.5 psig +/- 0.5 psig was chosen to ensure better system behavior and that actuation to re-open _LPSW- 1121, 1122, 1123, and 1124 between 21 and 31 psig is acceptable and that 25 +/- 1 psig was chosen to prevent toggling of the valve. The zero reference elevation of 815 feet 6 inches and Static head correction for the process tap elevation/transmitter elevation have been included in the transmitter calibration by IP/O/A/0250/010. 6.2 DESIGN INPUT/BASES 6.2.1 The postulation of a LOOP concurrent with a seismic event and a single failure is within the Oconee licensing basis (Attachment 5). 7.0 CALCULATION The calculation is organized as follows: Sections 7.1 contains a block diagram of the LPSW header pressure transmitters; Section 7.2 contains the device uncertainties; Section 7.3 contains the determination of the total loop uncertainty; Section 7.4 contains the setpoint analysis; and Section 7.5 contains values to aid in determining loop past operability.

OSC-8538 Rev. 4 Page 10 7.1 Instrument Block Diagram Figures 7.1-1 & 7.1-2 are block diagrams of th-._ Peactor Building LPSW Isolation circuit. FIGURE 7.1-1 INSTRUMENT BLOCK DIAGRAM AUXILIARY BUILDING EAST PENETRATION ROOM TAP 

                                                              ~Transmitter The tap elevations on all three Units for transmitters PT0010, PT0011, PT0012, and PT0013 are designated between 812' and 814'. The Zero Reference elevation on all three Units is to be designated at 815' 6" for transmitters PT0010, PT0011, PT0012, and PT0013.

Loop configuration and tag numbers and location were taken from References 5.H and 5.1 FIGURE 7.1-2 INSTRUMENT BLOCK DIAGRAM Reactor Building LPSW Isolation Circuit Instrument Loop LPS PT0010 (CHANNEL 1) I = Power Supply 2 = Transmitter 3 = Current Switch 4 = Signal Isolator (output to OAC) 2 4 3 Same for PT001 1(C12), PT0012(CH3), PT0013(CH4) Loop configuration and tag numbers and location were taken from Reference 5.T

OSC-8538 Rev. 4 Page 11 7.2 Device/Loop Uncertainty Term Identification The Reactor Building LPSW Low Pressure Isolation circuit consists of Rosemount 1154GP6RBN0037 pressure transmitters feeding Rochester XET-1215-T2-T1O-20012 current switches and Rochester XSC-1302-20012 voltage/current transmitter. The XSC-1302 module provides output to the OAC computer and is not part of the low pressure trip circuit, it does not affect the loop uncertainty errors and will not be addressed in the TLU calculations. 7.2.1 Rosemount Pressure Transmitter All uncertainties given in this section are for the Rosemount 1154GP6RBN0037 pressure transmitter (PT) and are taken from Attachment I unless otherwise specified. From Attachment 1, the Upper Range Limit of the Rosemount 1154GP6RBN0037 pressure transmitters is 0 - 100 psig. The Calibrated span is 0 - 100 psig. Accuracy (PTA) (random-independent uncertainty term) Accuracy is +0.25% of calibrated span. (Ref. Assumption 6.1.5) Drift (PTD) (random-independent uncertainty term) Drift is +/-0.2% of Upper Range Limit for 30 months. Power Supply Effect (PTPSE) (random-independent uncertainty term) Power supply effect is +0.005% of span per volt. The Rosemount 1154 is powered by the Rosemount ATM-49-15-0401 25 v power supply (attachment 4). This power supply has an output voltage tolerance of +/-0.5%, or +1.25vdc. For conservatism a voltage tolerance of +/-2.Ovdc will be used. Therefore: PTPSE = +0.005% * +/-2.Ovdc = +/-0.01% of span Static Pressure Effect (PTSPE) N/A for this application of the Rosemount Gauge Pressure Transmitter. Temperature Effect (PTTE) (random-independent uncertainty term) Temperature effect is +/-0.75% of Upper Range Limit per 100F plus +0.5% of span per 100°F. Per Reference 5.0.E, Table EP-2; the normal environmental temperature range for the transmitter mounting location is 60 to 130'F or a 70'F span. The maximum normal range temperature variation from the calibration point is therefore +70'F. PTTE = [+/-0.75% * (100psigURL / 100psigSPAN) + +/--.5%]

  • 70'F / 100lF = +/-0.875% of span

OSC-8538 Rev. 4 Page 12 Environmental Allowance (PTEA) (random-independent uncertainty term) Accuracy within +2.5% URL plus +/-0.5% of span during and after exposure to steam at the following temperatures: 420'F & 50psig for 3 minutes 350°F & 10Opsig for 7 minutes 320'F & 75psig for 8 hours 2650 F & 24psig for 56 hours PTEA = +2.5% * (100psig UJRL / 100psig span) + +0.5% of span = +3% of span. Seismic Allowance (PTSA) (random-independent uncertainty term) Accuracy within +0.5% of upper range limit. PTSA = +/-0.5% * (100psigURL / 100psigSPAN) =+0.5% of span. Current Leakage (PTCL) (bias uncertainty term) Current leakage will be determined using the methodology and formulas from Reference 5.B.c and based upon assumptions 6.1.2, 6.1.3, & 6.1.4. cable resistance From Reference 5.R, Cable IR is 2.0.x 106 ohms for a 15 foot test cable. UNIT I Installed cable in East Pen Room For lLPS PT0010 is 135 feet. 1LPS PTOO 11,12 & 13 mounting location for cable length and environment will be enveloped by ILPS PT0010 UNIT 2 Installed cable in East Pen Room For 2LPS PTOO10 is 125 feet. 2LPS PTOO 11,12 & 13 mounting location for cable length and environment will be enveloped by 2LPS PTOO 10. UNIT 3 Installed cable in East Pen Room For 3LPS PTOO10 is 160 feet. 3LPS PTOO11,12 & 13 mounting location for cable length and environment will be enveloped by 3LPS PT0010. Cable leakage will be calculated based on the cable length for 3LPS PT0010. This will envelope units 1 & 2. RCABLE = 15ft (2.0 x 106 Ohms) 160ft RCABLE = 187,500 ohms 1 /REQ = 1 / REQI + 1/(REQ2 + REQ 3 ) 

         / REQ = I / 187,500 + I / (187,500 + 187,500)

REQ = 125,000 ohms Current leakage

OSC-8538 Rev. 4 Page 13 VS = Loop Power Supply voltage = 25 v IS = Pressure Transmitter output current RL = Resistance of current loop loads = 500 ohms PTCL = [VS - IS(RL)] / (REQ + RL) PTCL = [25v - 0.004A(500 ohms)] / (125,000 ohms + 500 ohms) PTCL = 0.000183267A x 1O0OmA/A PTCL= 0.183mA current leakage (% of span) PTCL(%) = {PTCL / mA span} x 100% PTCL(%) = {0.183mA / l6mA} x. 100% PTCL(%) = +1.145% of span Radiation (R) N/A the LPSW Reactor Building Isolation Circuit is not required to operate under significant radiation exposure scenarios. 7.2.2 Rochester Instruments Current & Voltage Alarm module All uncertainties given in this section are for the Rochester Instruments Current & Voltage Alarm module (CS) (commonly referred to as "current switch") Model XET-1215-T2-T10-20012 and are taken from Attachment 2 unless otherwise specified. From Attachment 2, the current switch has an input signal range 4 to 20mA and an operating power input requirement of 11 5vac +/-20%, 60Hz. Accuracy (CSA) (random-independent uncertainty term) Accuracy is +0.1% of calibrated span (repeatability value) Drift (CSD) (random-independent uncertainty term) Drift is +0.1% (Assumption 6.1.1) Deadband (CSDB) Deadband is N/A. Per pages 2, 3 and 4 of Attachment 2 the current switch reset point is supplied with the deadband set at 0.5% of span and requires the RIO and R26 resistors to be replaced to adjust it to a setting other than 0.5%. This 0.5% deadband value is negligible for this application. Power Supply Effect (CSPSE) (random-independent uncertainty term) Power supply effect is +/-0.15% of span per +/-20% change in power supply voltage. Note: The device tolerance envelopes the EDM - 102, Appendix B, Para 102B.2 standard assumption of+/- 10% variation in nominal supply voltage. CSPSE = +/-0.15% of span

OSC-8538 Rev. 4 Page 14 Temperature Effect (CSTE) (random-independent uncertainty term) Temperature effect (Trip Point Stability) is +0.01% of span per degree F. Per Reference 5.0.E, Table EP-3; the normal environmental temperature range for the current switch mounting location is 60 to 100°F or a 40'F span. The maximum normal range temperature variation from the calibration point is therefore +40'F. CSTE = +0.01%

  • 40°F = +/-0.4% of span.

Environmental Allowance (CSEA) N/A because the current switch is in a mild environment and any temperature effects are bounded by the CSTE term. Radiation (R) N/A the LPSW Reactor Building Isolation Circuit is not required to operate under significant radiation exposure scenarios. Seismic Allowance (CSSA) (random-independent uncertainty term) Solid state electronics are not normally affected by vibration. However, for conservatism a seismic allowance of = +/-0.10% of span will be included in the TLU calculation. 7.2.3 Calibration Effects (CE) Primary Element Allowance (PEA) N/A because there is no primary element. Process Measurement Allowance (PMA) N/A because this application does not meet any of the four requirements necessary for considering PMA in a header pressure measurement listed in section 102.7.3 of EDM-102 Ref. B. Calibration (setting) Tolerance Effects (PTCTE) (random-independent uncertainty term) PTCTE is equal to the device accuracy since the IP calibrates it to within these limits. CSCTE is equal to the device accuracy since the IP calibrates it to within these limits. Measurement & Test Equipment (M & TE) (random-independent uncertainty term) PIM&TE = Heise HQS-1 & 2 Pressure Module 0.1% = +/-0.29psi **/ 100 = +/-0.29 % of span MM&TE = Fluke 45 DMM = +/-0.039mA** / 16mA = +/-0.24 % of span CSM&TE = Transmation 1040 Calibrator (current source) = +/-0.052mA**/ l6mAspan = 

     +/-0.33% of span

(** = Data from Reference 5.0.N, Table 7.0-i) PTCE =[PTCTE2 + PIM&TE2 + MM&TE 2]/2 =[+/-0.25%2 + +/-0.29%2 + +/-0.24%2]'/2 = 

     +/-0.45% of span

I OSC-8538 Rev. 4 Page 15 CSCE = [CSCTE 2 + CSM&TE 2]f2 = [+/-0.1%2 + +/- 0 .3 3 %2]'.2 = +0.34% of span 7.3 Total Loop Uncertainty Determination The Total Loop Uncertainty (TLU) for the Reactor Building LPSW Isolation Circuit will be determined for the applicable conditions using Equation 3-1 and the device uncertainties contained in Section 7.2. The determined Reactor Building LPSW Isolation Circuit TLU will be valid for the entire 0 to 100 psig range of the LPSW header pressure measurement. 7.3.1 TLU The total loop uncertainties for the'Reactor Building LPSW Isolation Circuit consist of errors contributed by the LPSW header pressure transmitter, current switch and the calibration effect (CE). UNCERTAINTY CALCULATION PTA = +0.25% of span PTD = +/-0.20% of span PTPSE= +0.01% of span PTTE = +/-0.875% of span PTCE +/-= 0.45% of span PTEA = -3.0%'of span PTSA = +/-0.50% of span PTCL = +1.145% of span CSA = +/-0.10% of span CSD = +/-0.10% of span CSPSE= +/-0.15% of span CSTE = +/-0.40% of span CSCE = +/-0.34% of span CSSA = +0.10% of span 2 2 2 2 

 " TLUNORM =    - [PTA 2 + PTD2 + PTPSE2 + PTTE2 + PTCE + CSA + CSD + CSPSE + CSTE
               + CSCE 2 ]112 (% span)
              =+/- [0.252 + 0.202 +0.012 + 0.8752 + 0.452 + 0.102 + 0.102 +0.152 ++/-0.402+ 0.342 ]1/2

(% span) 

             =+/- 1.18% span=+/- 1.18 psi
 " TLUsEIs = [PTA2 + PTD2 + PTPSE2       +   PTTE2 + PTCE         + PTSA + CSA2 + CSD2 + CSPSE2 +

CSTE2 + CSCE + CSSA 2 ]1/ 2 (% span)

OSC-8538 Rev. 4 Page 16 

            = + [0.252 + 0.202 +0.012 + 0.8752 + 0.452 +0.502 + 0.102 +      0.102 +0.152 + 0.402 +

0.342 +0.102]12 (% span) 

            =+ 1.28% span= + 1.28 psi 2        2        2       2          2
             + [PTA 2 + PTD 2 + PTPSE 2 + PTTE + PTCE + PTEA + CSA + CSD + CSPSE 2

+ TLUHELB = 

              + CSTE2 + CSCE ]11               (% span) + PTCL (% span)
               + [0.252 + 0.202 +0.012 + 0.8752 + 0.452 +/- 3.02    + 0.102 + 0.102  +/-0.152 + 0.402 +

0.342 ]12 (% span) + 1.145 (% span) 

              ++ 4.37% of span      +4.37 psi

-TLUHELB = - [PTA2 + PTD2 + PTPSE2 + PTTE2 + PTCE2 + PTEA2 + CSA2 + CSD2 + CSPSE2 + CSTE2 + CSCE 2]12 (% span) + PTCL (% span) 

             - _ [0.252 + 0.202 +0.012 + 0.8752 + 0.452 + 3.02    + 0.102 + 0.102  +/-0.152 + 0.402  +

0.34 ]1/2 (% span) + (+1.145) (% span) 

            =-2.08 % of span = -2.08 psi

+ TLUHELB + SEIS = + [PTA2 + PTD 2 + PTPSE2 + PTTE2 + PTCE2 + PTEA2 + PTSA 2 + CSA 2 + CSD 2 + CSPSE2 + CSTE2 + CSCE 2 + CSSA 2 ]l/2 *0.84 (% span) + PTCL 

                  = +   [0.252 + 0.202 +0.012 + 0.8752 + 0.45 2+ 3.02 +0.502  + 0.102 + 0.102 +0.152
                    +0.402 +0.342+ 0.102]1/2          (% span) + 1.145 (% span)
                    + 4.41% of span = +4.41 psi

- TLUHELB + SEIS = - [PTA2 + PTD 2 + PTPSE 2 + PTTE2 + PTCE 2 + PTEA2 + PTSA2 + CSA2 + CSD2 + CSPSE2 + CSTE 2 + CSCE 2 + CSSA 2 ]1/2 (% span) + PTCL 

                  = -  [0.252 + 0.202 +0.012 + 0.8752 + 0.45 2+ 3.02 + 0.502 + 0.102  + 0.102 +0.152
                    + 0.402 + 0.34 2+ 0.102]1/2       (% span) + (+1.145) (% span)
               -2.12 % of span = -2.12 psi

.I OSC-8538 Rev. 4 Page 17 7.4 Setpoint Analysis and/or Acceptability of Loop Uncertainty The calculated LPSW Reactor Building Isolation circuit uncertainty applies to the conditions outlined in Section 12. The LPSW pressure transmitters monitor LPSW header pressure and the associated current switches will have a setpoint to actuate the LPSW Reactor Building Isolation circuit when LPSW decreases to 18.5 psig. The LPSW Reactor Building Isolation Circuit has the following uncertainties during normal environmental conditions. TLUNORM 

        +          =  +/- [PTA 2 + PTD2 + PTPSE 2 + PTTE 2 + PTCE 2 + CSA 2 + CSD 2 + CSPSE 2 +

CSTE 2 + CSCE2 ]1/ 2 *0-84 (% span) 

                   =  +/- [0.252 + 0.202 +0.012  + 0.8752 + 0.45'+ 0.102 + 0.102 +0.152 + 0.402      +

0.342 ]1/2 (% span) 

                      = 1.18% span=+ 1.18 psi Trip Set point analysis The LPSW Reactor Building Isolation Circuit uncertainty is +/- 1.18 psi (under normal conditions). Therefore, for Reactor Building LPSW isolation to occur at a decreasing pressure of 18.5 psig, the Reactor Building Isolation Circuit should be set at a value of 18.5 psig.

Calibrated Setpoint = 18.5 psig Process Actuation Point = Calibrated Setpoint - TLUNORM = 18.5 psig - 1.18 psig Process Actuation Point = 18.5 psig - 1.18 psig = 17.32 psig This is within the mechanical system calculation OSC-8144 acceptable process analytical range of 15 to 24 psig. Closure of the Reactor Building LPSW Isolation valves is required when the LPSW Pumps trip due to a LOOP. Low LPSW header pressure is the process parameter being monitored to determine that the LPSW pumps have tripped. Closure of the Reactor Building LPSW Isolation valves is required to preclude occurrence of a water hammer when the LPSW pumps restart. The loop errors for normal conditions and for normal conditions + seismic are essentially the same i.e., 1 psig and therefore this error will be used in establishing the setpoint for closure of the Reactor Building LPSW Isolation valves. Closure of the Reactor Building LPSW Isolation valves will still occur during a HELB + SEIS condition because the worst case HELB + SEIS error as shown above is +4.41 psig. Therefore, when the LPSW Pumps trip and the true process pressure goes to 0 psig, during the HELB +SEIS environment the Reactor Building LPSW Instrumentation would see a current signal equivalent of 4.41 psig. This is 14.09 psig below the 18.5 psig set point therefore the current switch in each channel will trip and command closure of the valves. Since the

OSC-8538 Rev. 4 Page 18 Reactor Building LPSW Low Pressure Instrumentation is not required to mitigate a HELB in the Pen Room it is not necessary to utilize the HELB string error value in establishing the trip set point. Reset Set Point Analysis The LPSW Reactor Building Isolation Circuit uncertainty is +/- 1.18 psi (under normal conditions). Therefore, the reset for Reactor Building LPSW isolation to occur at an increasing pressure of 25 psig, the Reactor Building Isolation Circuit should be set at a value of 25 psig. Calibrated Setpoint = 25 psig Process Actuation Point = Calibrated Setpoint + TLUNoRM = 25 psig + 1.18 psig Process Actuation Point = 25 psig + 1.18 psig =26.18 psig This is within the mechanical system calculation OSC-8144 acceptable process analytical range of 21 to 31 increasing. Opening of the Reactor Building LPSW discharge valves is required when the LPSW Pumps are restarted following a LOOP or loss of normal LPSW header pressure. Low LPSW header pressure is the process parameter being monitored to determine that the LPSW pumps have been restored. Opening of the Reactor Building LPSW discharge valves is required to mitigate an ES event. The loop errors for normal conditions and for normal conditions + seismic are essentially the same i.e., 1 psig and therefore this error will be used in establishing the setpoint for opening of the Reactor Building LPSW discharge valves. Opening of the Reactor Building LPSW discharge valves will still occur during a HELB + SEIS condition because the worst case positive HELB + SEIS error as shown above is +4.41 psig. Therefore, when the LPSW header pressure is restored and the true process pressure goes to 25 psig, during the HELB + SEIS environment the Reactor Building LPSW Instrumentation would see a current signal equivalent of 29.41 psig. This is 4.41 psig above the 25 psig set point therefore the current switch in each channel will reset and command the opening of the Reactor Building discharge valves. Opening of the Reactor Building LPSW discharge valves will also occur during a HELB + SEIS condition if the worst case negative HELB +SEIS should occur because the worst case HELB + SEIS negative shown above is -2.12 psig. Therefore, when the LPSW header pressure is restored and the true process pressure goes to 25 psig, during the HELB environment the Reactor Building LPSW Instrumentation would see a current signal equivalent of 22.88 psig. This is 2.12 psig below the 25 psig set point but still within the 21 to 31 psig acceptable analytical range. Therefore the current switch in each channel will reset and command the opening of the Reactor Building discharge valves. Since the Reactor Building LPSW Water Hammer Protection Instrumentation is not required to mitigate a HELB in the Pen Room it is not necessary to utilize the HELB +SEIS string error value in establishing the reset set point.

OSC-8538 Rev. 4 Page 19 7.5 Values to Aid in Determining Loop Past Operability The purpose of this section is to determine how much uncertainty, which may appear during the "as-found" portion of the loop calibration, has been accounted for in the total loop uncertainties calculated in Section 7.3. Values recorded during loop as-found calibration which are less than those documented in this section would clearly indicate the loop to be past operable. Values recorded during loop as-found calibration which exceed those documented in this section would require a more detailed review to determine the effects of the increased uncertainty on the past operability of the loop. All values documented in this section are to aid in determining loop past operability only. The limits for loop as-found values associated with engineering notification requirements should be based on the requirements specified in the calibration procedure (Reference 5.G). The following uncertainty terms are applicable when determining the loop past operability values, reference accuracy (A), drift (D), setting tolerance (CTE), measurement and test equipment (MTE). All data and equations used in this section were taken from previous sections of this calculation. Initial As-Found String Check This is the initial as-found calibration, which places the appropriate pressure (current) values corresponding to the required actuation setpoint on the current switches and checks for current switch actuation. The uncertainties due to the Reactor Building LPSW Isolation circuit will be determined for the total acceptable limit. As a separate CE allowance was determined in Section 7.2.3, which encompassed CTE and MTE, the CE allowance will be utilized in the acceptable limit determination. The acceptable limit is determined as follows: LPSW header switches ALAS-FOUND = +/- (PTA2 + PTD 2+ PTCE 2 + CSA 2 + CSD 2+ CSCE 2)1"2 % span 

                           =  +/-(0.252 +0.202 +0.452 + 0.102 + 0.102 + 0.342)1/2
                           = +/- 0.66% span (+/- 0.66 psig)

CTE +/- (PTCTE2 + CSCTE 2)1/2 % span 

                           =_(0.252   + 0.10)1/2 0.27% span ALAS-FOUND/CTE        +/- 0.66% span/0.27% span > 2 times.

The ALAS-FOUND/CTE ratio supports the NSD-219 2X OOT Limit. Therefore, revision of OSC-7165 (Reference 5.0) will not be required.

OSC-8538 Rev. 4 Page 20 No as found data for this instrument loop (new installation) presently exist. Data should be collected and this calculation evaluated to verify that the calculation supports the as found data. This calculation is a Type III Calc. 8.0 Conclusion/Results The purpose of this calculation is to determine the uncertainties for the Oconee Units 1, 2 and 3 LPSW Supply to Reactor Building Isolation Circuitry pressure transmitters and current switches. This calculation, was designated a QA Condition 1 Calculation as the LPSW Reactor Building Isolation Circuitry pressure transmitters and current switches are relied upon to automatically close LPSW-1054,-1055,-1061,-1062,-l121,-122,-l123,- 1124 and open _LPSW- 1150 and _LPSW-1 151 on decreasing LPSW header pressure and automatically reopen LPSW- 1121,- 122,-i123, and -1 124 and reclose _LPSW- 1150 and LPSW- 115 1on increasing LPSW header pressure. The methodology used in this calculation is based on EDM-102 and the ISA Setpoint Standard and Recommended Practice (References 5.B.a, 5.B.b and 5.B.c, respectively). The assumptions used in this calculation, as well as certain design bases and inputs, are given in Section 6.0. The Total Loop Uncertainties from Section 7.3 are: 

            +/- TLUNORM = +/- 1.18% span= +/- 1.18 psi
            +/- TLUSEIS = +/- 1.28% span= +/- 1.28 psi
            + TLU1IELB  =   + 4.37% of span = +4.37 psi
            -TLUHELB=    -.  -2.08 % of span = -2.08 psi
            + TLUHELB +SEIS= + 4.41% of span = +4.41 psi
            - TLUHELB + SEIS  =  -2.12 % of span = -2.12 psi The total loop uncertainties calculated are acceptable and show that the closure of the Reactor Building LPSW Isolation Valves _LPSW-1054, 1055, 1061, 1062, 1121,1122,1123,1124 and the opening of the vent valves _LPSW-1 150 and _LPSW-1 151 will occur on loss of LPSW header pressure. The total loop uncertainties also show that the Reactor Building LPSW discharge valves LPSW-1121,1122,1123 and 1124 will automatically re-open and the vent valves

_LPSW1150 and _LPSW-1151 will reclose to restore the ES success path for LPSW upon restoration of normal LPSW header pressure.

ATTACHMENT 2 TECHNICAL SPECIFICATIONS - MARK UP

LPSW RB Waterhammer Prevention Reset Circuitry 3.3.27 ACTIONS (continued) _ CONDITION REQUIRED ACTION COMPLETION TIME C. Two or more required C.1 Open two LPSW RB Immediately LPSW RB Waterhammer Waterhammer Prevention Pneumatic Prevention analog reset Discharge Isolation channels inoperable, valves in the same header. OR Two required LPSW RB AND Waterhammer Prevention digitallogic C.2 T actions to resto'e Immediately required LPSW RBed l reset channels C inoperable. r WB e L OR -revention analog reset Oor digital logic reset Required Actions and channels to OPERABLE status. associated Completion Times of Condition A or B not met. OCONEE UNITS 1, 2, & 3 3.3.27-2 Amendment Nos. I

ATTACHMENT 3 TECHNICAL SPECIFICATIONS - REPRINTED PAGE

LPSW RB Waterhammer Prevention Reset Circuitry 3.3.27 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Two or more required C.1 Open two LPSW RB Immediately LPSW RB Waterhammer Waterhammer Prevention Pneumatic Prevention analog reset Discharge Isolation channels inoperable, valves in the same header. OR Two required LPSW RB AND Waterhammer Prevention digital logic C.2 Initiate actions to restore Immediately reset channels required LPSW RB inoperable. Waterhammer OR Prevention analog reset or digital logic reset Required Actions and channels to OPERABLE status. associated Completion Times of Condition A or B not met. OCONEE UNITS 1,2, & 3 3.3.27-2 Amendment Nos.}}

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