ML112620402
| ML112620402 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 09/30/2011 |
| From: | Lynnea Wilkins Plant Licensing Branch IV |
| To: | Bannister D Omaha Public Power District |
| Wilkins, L E, NRR/DORL/LPL4, 415-1377 | |
| References | |
| TAC ME4542 | |
| Download: ML112620402 (21) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 30, 2011 Mr. David J. Bannister Vice President and CNO Omaha Public Power District Fort Calhoun Station 444 South 16th St. Mall Omaha, NE 68102~2247
SUBJECT:
FORT CALHOUN STATION, UNIT NO. 1 ~ ISSUANCE OF AMENDMENT RE:
REVISION OF TECHNICAL SPECIFICATIONS TO RELOCATE POWER~
OPERATED RELIEF VALVE/SAFETY VALVE POSITION AND TAIL PIPE TEMPERATURE INSTRUMENTATION (TAC NO. ME4542)
Dear Mr. Bannister:
The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 268 to Renewed Facility Operating License No. DPR-40 for the Fort Calhoun Station, Unit No.1 (FCS).
The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated August 16, 2010, as supplemented by letters dated September 27,2010, April 6, 2011, and June 30, 2011.
The amendment relocates the operating and surveillance requirements for the power~operated relief valve and pressurizer safety valve acoustic position indication and tail pipe temperature from TS 2.15, "Instrumentation and Control Systems," Table 2-5, "Instrumentation Operating Requirements for Other Safety Feature Functions," Items 3, 4, and 5 instrumentation to FCS's Updated Safety Analysis Report. The amendment also revises the surveillance requirement, TS 3.1, "Instrumentation and Control," Table 3-3, "Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls," Items 21,23, and 24.
Additionally, the TS Table 2-5 associated Note 'e' is re-Iettered to Note 'a' and TS Table 2~5 footnote "i' to Note 'e' is deleted.
D. Bannister
- 2 A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.
- ?~/.
Lynnea E. Wilkins, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-285
Enclosures:
- 1. Amendment No. 268 to DPR-40
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 OMAHA PUBLIC POWER DISTRICT DOCKET NO. 50-285 FORT CALHOUN STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. DPR-40
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Omaha Public Power District (the licensee), dated August 16, 2010, as supplemented by letters dated September 27,2010, April 6, 2011, and June 30,2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
- 2
- 2.
Accordingly, Renewed Facility Operating License No. DPR-40 is amended by changes as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-40 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 268, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
- 3.
The license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance. Consistent with the requirements in 10 CFR 50.71(e), implementation shall include revision to the Updated Safety Analysis Report, including Chapter 4.3, to include the effects of all changes made in the facility or procedures and all safety analysis and evaluations performed by the licensee in support of the license amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License No. DPR-40 and Technical Specifications Date of Issuance: September 30, 2011
ATTACHMENT TO LICENSE AMENDMENT NO. 268 RENEWED FACILITY OPERATING LICENSE NO. DPR-40 DOCKET NO. 50-285 Replace the following pages of the Renewed Facility Operating License No. DPR-40 and the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
License Page REMOVE INSERT -3 Technical Specifications REMOVE INSERT
- 2. 15 - Page 14
- 2. 15 - Page 14 3.1-Page 19 3.1 - Page 19
- 3 (4)
Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or when associated with radioactive apparatus or components; (5)
Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.
- 3.
This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is, subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
A.
Maximum Power Level Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not in excess of 1500 megawatts thermal (rate power).
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 268 are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.
C.
Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan," submitted by letter dated May 19, 2006.
OPPD shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The OPPD CSP was approved by License Amendment No. 266.
Renewed Operating License No. DPR-40 Amendment No. 268
TECHNICAL SPECIFICATIONS TABLE 2*5 Instrumentation Operating Reguirements for Other Safety Feature Functions No.
Functional Unit Minimum Operable Channels Minimum Degree of Redundancy Permissible Bypass Condition 1
CEA Position Indication Systems 1(a)
Pressurizer Level 1
None Not Applicable NOTES:
(a)
If one channel of CEA position indication is inoperable for one or more CEAs, requirements of specification 2.15 are modified for item 1 to "Perform TS 3.1, Table 3~3. Item 4 within 15 minutes following any CEA motion in that group." Specifications 2.15(1), (2), and (3) are not applicable.
2.15 ~ Page 14 Amendment No. M,95,110,24Q,295,297 I 268
TECHNICAL SPECIFICATIONS TABLE 3-3 (Continued)
MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TESTING OF MISCELLANEOUS INSTRUMENTATION AND CONTROLS Surveillance Channel Description Function Frequency
- 19. Auxiliary Feedwater Flow
- a.
Check M
- b.
Calibrate R
- 20. Subcooled Margin Monitor
- a.
Check M
- b.
Calibrate R
- 21. PORV Operation
- a.
Test R
- 22. PORV Block Valve Operation
- a.
Check Q
and Position Indication
- b.
Calibrate R
- 23. Not Used
- 24. Not Used Surveillance Method CHANNEL CHECK CHANNEL CALIBRATION CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST Cycle valve. Valve is exempt from testing when it has been closed to comply with LCO Action Statement 2.1.6(5}a.
Check valve stroke against limit switch position.
3.1 - Page 19 Amendment No. 39,54,110,161,182, ~~~
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO, 1 DOCKET NO, 50-285 1,0 INTRODUCTION By letter dated August 16, 2010, as supplemented by letters dated September 27,2010, April 6, 2011, and June 30, 2011 (Agencywide Documents Access and Management System (ADAMS)
Accession Nos, ML102290067, ML102720964, ML110970356, and ML111822610, respectively), Omaha Public Power District (OPPD, the licensee) proposed changes to the Technical Specifications (TSs) for Fort Calhoun Station, Unit No.1 (FCS). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on January 25, 2011 (76 FR 4388).
The proposed changes would relocate the operating and surveillance requirements (SRs) for the power-operated relief valve (PORV) and pressurizer safety valve (PSV) position and tail pipe temperature instrumentation from the TSs to the FCS Updated Safety Analysis Report (USAR). Specifically, the proposed changes would revise:
TS 2.15, "Instrumentation and Control Systems," Table 2-5, "Instrumentation Operating Requirements for Other Safety Feature Functions," and TS 3,1, "Instrumentation and Control," Table 3-3, "Minimum Frequencies for Checks, Calibrations and Testing of Miscellaneous Instrumentation and Controls,"
Functions 3,4,5, and the associated footnotes a, b, c, and d in TS Table 2-5, and Functions 23 and 24 in TS Table 3-3, would be relocated from the FCS TSs and placed in the FCS USAR.
TS Table 3-3 Function 21 would be revised to be consistent with NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants," Revision 3.0. Additionally, the TS Table 2-5 associated Note 'e' would be re-Iettered to Note 'a' and TS Table 2-5 footnote 'i' to Note 'c' would be deleted.
- 2
2.0 REGULATORY EVALUATION
2.1
System Description
The FCS reactor coolant system (RCS) is protected against overpressurization by control and protective circuits, such as the pressurizer pressure high reactor trip, and by the two PORVs and the two PSVs connected to the top of the pressurizer. Upon opening, these valves discharge steam into the pressurizer quench tank, which condenses and collects the valve effluent. When PORVlPSV valves lift or seats leak, two independent monitoring systems (acoustic and temperature) alert the operator to the passage of steam or liquid through those valves. The PORV/PSV acoustic monitor detects downstream acoustic vibrations generated from the steam flowing through the valve and actuates an alarm in the control room. In addition, a temperature sensor upstream of the acoustic sensor detects the temperature increase in the line and actuates a control room alarm when the valve releases steam or liquid. Therefore, both monitoring systems provide the operator with PORV/PSV position information and alert the operator with the indication of flow through those valves.
The licensee added acoustic monitors to the FCS TS through Amendment No. 54 dated January 19,1981 (Legacy ADAMS Accession No. 8101300614), to meet the requirements of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," issued July 1979 (ADAMS Accession No. ML090060030), and NUREG-0737, "Clarification of TMI Action Plan Requirements," issued November 1980 (ADAMS Accession No. ML051400209). The licensee stated that it uses the temperature sensors installed downstream of the PORVlPSV to identify flow through these valves when acoustic position indication is inoperable. These sensors provide indication and alarm in the control room and indication on the plant computer.
2.2 General Requirements Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TSs as part of the license. The Commission's regulatory requirements related to the content of TSs are contained in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.36, 'Technical specifications." The TS requirements in 10 CFR 50.36 include the following specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. However, the rule does not specify the format and content for TS categories.
IMPROVED STANDARD TECHNICAL SPECIFICATIONS PROGRAM Interim Policy Statement on TS Improvements The NRC and industry have representatives sought to develop guidelines for improving nuclear power plant TS content and quality. On February 6, 1987, the Commission issued an "Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (52 FR 3788). In September 1992, the Commission issued NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants" (STS), which was developed using the guidance
- 3 and criteria contained in the Commission's Interim Policy Statement. The STS are a model for developing improved TS for Combustion Engineering plants. The Interim Policy Statement criteria ensure that improved TS would consistently reflect system configurations and operating characteristics for the Combustion Engineering design. In addition, the generic Bases statements provide the basis for each of the STS requirements.
Final Policy Statement on TS Improvements On July 22, 1993, the Commission issued its Final Policy Statement indicating that satisfying the guidance in the policy statement also satisfies Section 182a of the Atomic Energy Act and 10 CFR SO.36 (S8 FR 39132). The Final Policy Statement described the improved STS safety benefits and encouraged licensees to use the improved STS as the basis for plant-specific TS amendments and for complete conversions to the improved STS. Further, the Final Policy Statement gave guidance for evaluating the required scope of the Improved Technical Specification (ITS) and defined the guidance criteria for determining which of the LCOs and associated surveillances should remain in the ITS. Using this approach, licensees should keep existing LCO requirements that fall within or satisfy any of the Final Policy Statement criteria in the TSs. Those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The Commission codified the four criteria in 10 CFR SO.36 (c)(2)(ii)(60 FR 369S3, July 19, 1995).
Final Policy Statement Criteria The Final Policy Statement criteria are as follows:
Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to fission product barrier integrity.
Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Criterion 4. A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.
The regulations in 10 CFR SO.36(c)(2) state, in part, that When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.
-4 The regulations in 10 CFR 50.36{c){3) state that Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The NRC's guidance for the format and content of licensee TS can be found in NUREG-1432, "Standard Technical Specifications Combustion Engineering Plants," Revision 3.0.
NRC Regulatory Guide (RG) 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions during and following an Accident," Revision 2, issued December 1980 (ADAMS Accession No. ML060750525), lists five types (Types A-E) of variables to help designers select the accident monitoring instrumentation and applicable criteria. Categories 1, 2, and 3 separate the type criteria into groups for a graded approach to requirements, depending on the importance to safety or the measurement of a specific variable.
In a memorandum dated September 18,1992 (ADAMS Accession No. ML003763736), the Commission approved the NRC staffs proposal in SECY-92-223, "Resolution of Deviations Identified During the Systematic Evaluation Program," not to apply 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," to plants with construction permits prior to May 21, 1971. FCS was licensed for construction prior to May 21, 1971, and at that time committed to the draft General Design Criteria (GDC). The draft GDC, which are similar to Appendix A, "General Design Criteria for Nuclear Power Plants," in 10 CFR Part 50, are contained in Appendix G, "Response to 70 Criteria," of the FCS USAR.
By letter dated August 16, 2010, the licensee appropriately identified the following draft GDC as specified in Appendix G to the FCS USAR:
FCS Design Criterion 12 - Instrumentation and Controls Systems Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.
FCS Design Criterion 16 - Monitoring Reactor Coolant Pressure Boundary states:
Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage.
The NRC staff also used the following for its review:
NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations," July 1979.
NUREG-0737, "Clarification of TMI Action Plan Requirements," issued November 1980, Item 11.0.3.
- 5 "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," dated May 9,1988 (ADAMS Legacy Library Accession No. 9012130142).
3.0 TECHNICAL EVALUATION
3.1 Proposed TS Changes
The licensee's proposed changes are as follows:
TS Limiting Condition for Operation (LCO) 2.15, "Instrumentation and Control Systems" Table 2-5, "Instrumentation Operating Requirements for Other Safety Feature Functions" Item No.3, PORV Acoustic Position Indication-Direct, is being deleted.
Item No.4, Safety Valve Acoustic Position Indication, is being deleted.
Item No.5, PORV/Safety Valve Tail Pipe Temperature, is being deleted.
Notes a, b, c, and d are being deleted and Note Ie' is being re-Iettered as Note 'a.'
Footnote (i) associated with Note c is being deleted.
TS 3.1, "Instrumentation and Control" Table 3-3, "Minimum Frequencies for Checks, Calibrations, and Testing of Miscellaneous Instrumentation and Controls" Item 21, PORV Operation and Acoustic Position Indication, is being revised to delete the words "and Acoustic Position Indication."
Item 21.a, the frequency is being changed from "M" to "R" to reflect channel functional testing of PORV operation, which is more aligned with NUREG-1432 surveillance requirements.
Item 21.b, the PORV acoustic position indication channel calibration is being deleted in its entirety.
Item 23, Safety Valve Acoustic Position Indication, is being deleted and replaced with "Not Used."
Item 24, PORV/Safety Valve Operation Tail Pipe Temperature, is being deleted and replaced with "Not Used."
- 6 In conjunction with the proposed TS changes, the licensee will incorporate operability and SRs for the acoustic position indication and tail pipe temperature indication instrumentation into the FCS USAR and associated plant procedures.
The above proposed TS modification would provide relief for the licensee's burden to meet the hot shutdown requirement if an inoperable PORVlPSV acoustic position indication or tail pipe temperature instrument channel is not restored to OPERABLE status before the allowable outage time has expired.
3.2 Precedents In its letter dated August 16, 2010, the licensee stated, in part, that:
Precedent for removing the PORV/PSV acoustic position indication operability and surveillance requirements was found in Amendments No. 179 and 141 for the Limerick Generating Station, Units 1 and 2, respectively [ADAMS Accession No. ML052550369] as requested in Reference 6.5. Amendments No. 141 and 179 issued on September 27,2005, document the relocation of the operability and surveillance requirements for the reactor coolant system safety/relief valve position instrumentation from the TS to the Limerick Generating Station Technical Requirements Manual. Thus, the proposed LAR for deleting the FCS TS operating requirements and associated surveillance requirements for the PORVs and PSVs acoustic position indication and relocating these requirements to licensee-controlled documents is similar to that approved by the NRC for Limerick as found in Amendments No. 179 and No. 141 to the Limerick Generating Station's Operating License for Units No.1 and No.2, respectively
[ADAMS Accession No. ML052550369].
3.3
NRC Staff Evaluation
The proposed modifications do not Involve any phYSIcal alterations of the FCS facility. The NRC staff reviewed the licensee's application for license amendment request (LAR) against the requirements stated in Section 2.2 of this safety evaluation.
3.3.1 TS LCO 2.15, Table 2 Remove TS Limiting Condition for Operation Items 3, 4, and 5 and Associated Notes TS LCO 2.15, Table 2-5, Items 3, 4, and 5 specify operability requirements for PORVlPSV acoustic position indication and tail pipe temperature instrumentation. The associated notes of Items 3, 4, and 5 specify the associated actions for the LCOs of PORV/PSV pOSition indication instrumentation.
The proposed changes would remove PORV/PSV acoustic position indication and tail pipe temperature from FCS TS Table 2-5, Items 3, 4, and 5, and associated Notes a, b, c, and d.
This LAR also proposes to delete Footnote i associated with Note c, which was added to FCS TS Table 2-5 for an emergency amendment request subsequently approved by the NRC.
- 7 3.3.2 TS 3.1, Table 3 Revise SR Items 21, 23, and 24 TS 3.1, Table 3-3, Items 21, 23, and 24 specify the SRs for PORV/PSV acoustic position indication and tail pipe temperature indication instrumentation. A monthly channel functional test (or monthly channel check for tail pipe temperature) and refueling interval channel calibration are required.
The proposed change would modify TS 3.1, Table 3-3, Items 21, 23, and 24, to relocate the PORV/PSV position indication and tail pipe temperature indication instrumentation operability and SRs from the TS to the FCS USAR and associated plant procedures. Per the licensee's letter dated August 16, 2010, future changes to PORV/PSV position instrumentation requirements will be subject to the controls of 10 CFR 50.59, "Changes, tests, and experiments." Besides deleting the acoustic position indication portion of SR Item 21 in TS Table 3-3, the licensee proposed to maintain the PORV operation SR in Item 21.
Specifically, the licensee would revise Item 21.a to reflect the performance of the PORV operation channel functional test on its existing refueling (R) frequency and would delete the monthly (M) frequency denoted in the TS for the acoustic position indication.
3.3.3 Evaluation The licensee stated in Section 3.0, the technical evaluation section of the application, that the PORV/PSV acoustic and tail pipe temperature position indication instrumentation are collectively referred to as safety relief valve (SRV) position indication instrumentation or SRV position instrumentation.
In the application, the licensee analyzed the safety basis for SRVs using the Final Policy Statement criterion of 10 CFR 50.36. The SRVs are a part of the primary success path in the USAR accident analysis in that they can actuate to mitigate a design basis accident (DBA) and, therefore, meet Criterion 3. Accordingly, their operability is required by TS 2.1.6, "Pressurizer and Main Steam Safety Valves." However, the SRV position indication does not detect or indicate a Significant abnormal degradation of the reactor coolant pressure boundary (RCPB) considered by Criterion 1.
SRV position indication is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis considered in Criterion 2.
The FCS USAR Section 4.3, "Reactor Coolant System - Component and System DeSign and Operations," states, in part, that the PORV and PSVs have acoustic monitors and tail pipe temperature indicators in the control room to provide flow indication. While the function of SRVs is part of the primary success path and the SRVs actuate to mitigate a DBA or transient, SRV position indication does not form a part of the primary success path since USAR accident analysis assumes the SRVs function as deSigned (i.e., the USAR analysis assumes no operator action based on SRV position for the valves to perform their primary success path function considered in Criterion 3).
- 8 Furthermore, the licensee stated that the loss of this instrumentation has no effect on the probabilistic safety assessment, and has not been shown to be significant to public health and safety as considered in Criterion 4.
Finally, the licensee stated that the PSVs discharge into the pressurizer quench tank. The temperature, pressure, and liquid level of this tank are indicated and alarmed in the control room. A change in these parameters would alarm and alert the operator of a PSV discharge condition. Abnormal operating procedures and emergency operating procedures contain instructions noting that RCS leakage to the pressurizer quench tank is indicated by a rise in tank pressure, temperature, or level and rising or elevated pressure relief line temperatures or flow indication from the relief line acoustic monitors. Thus, failure of PSV position indication would not pose a significant challenge to the ability of the operations staff to respond to a DBA or plant transient.
Accident monitoring instrumentation is provided to monitor variables and systems over their anticipated ranges for accident conditions, as appropriate, to ensure adequate safety during and following accidents. These variables are used by the control room operating personnel to perform their role in the emergency plan in the evaluation, assessment, and monitoring of events, and execution of control room functions. American National Standards Institute/American Nuclear Society N4.5-1980, "Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors," delineates criteria for determining the variables to be monitored in the control room by the operator. Instrumentation deSignated as Types A through D aid the designer in selecting the accident monitoring instrumentation and applicable criteria.
Categories 1, 2, and 3 separate the Type criteria into groups for a graded approach to requirements depending on the importance to safety or the measurement of a specific variable.
RG 1.97 Type A instruments monitor primary information required to permit the control room operator to take specific manually-controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for DBA events.
RG 1.97 Category 1 instruments are designed for full qualification, redundancy, continuous real-time display, and onsite (standby) power. Accordingly, the NRC staff position to include all plant-specific RG 1.97 Type A instruments specified in the plant's SER, and all RG 1.97 Category 1 instruments to be included in TSs can be applied to FCS.
RG 1.97 describes PORV and safety valve position as a Category 2, Type D variable, which is defined as follows:
Category 2 provides for qualification, but it is less stringent because it does not include seismic qualification, redundancy, or continuous display and because it requires only a high-reliability power source (not necessarily standby power).
Type D instruments provide information to indicate the operation of individual safety systems and other systems important to safety.
In its letter dated August 16, 2010, the licensee stated that PORV and safety valve flow is a RG 1.97 Type D, Category 2, instrumentation. Type D instruments provide information to indicate the operation of individual safety systems and other systems important to safety.
- 9 Category 2 instruments are designed to less stringent qualifications that do not require seismic qualification, redundancy, or continuous display, and require only a high reliability power source (not necessarily standby power).
Because FCS identifies its PORV/PSV position instrumentation as a Category 2, Type 0 variable, its relocation of the PORV/PSV position indicator SR from the TS to a licensee controlled document conforms with the NRC's position on the application of 10 CFR 50.36 screening criteria to post-accident monitoring instrumentation.
The NRC staff reviewed the licensee's technical analysis of the proposed changes to the accident monitoring instrumentation section of the TSs against the requirements in 10 CFR 50.36. Based on the discussion above, the NRC staff determined that the PORV/safety valve position indication accident monitoring instrumentation does not meet the Type A or Category 1 instrumentation designations. In addition, the NRC staff also determined that the proposed changes are consistent with the requirements in the STSs. Therefore, the NRC staff concludes that relocating the proposed PORV SRV position indication accident monitoring instrumentation requirements from the FCS TSs to a licensee-controlled document is acceptable.
In its letter dated August 16, 2010, the licensee analyzed the safety basis for PORV/PSV position indications using the statement criteria in 10 CFR 50.36(c)(2)(ii). On the basis of Criterion 1, the licensee stated that the PORV/PSV acoustic position indication and tail pipe temperature indication does not detect or indicate a significant abnormal degradation of the RCPB. During the review process, the NRC staff noted that NUREG-0737, Item 11.0.3 requires that the direct indication of PORV/PSV valve position to be listed in the TSs, and was concerned with licensee's evaluation regarding the Criterion 1. The NRC staff noted that the PORV/PSV valve position indications are allowed to be relocated to the licensee-controlled document per NRC's document, "NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy statement Criteria to Standard Technical Specifications," dated May 9, 1988, and the Commission's Final Policy Statement dated July 22, 1993 (58 FR 39132).
In addition, the NRC staff recognized that PORV/PSV position indication is not the only means used to detect the opening of a PORV/PSV valve as described in GOC 16. The FCS evaluation for the compliance with draft GOC 16 in the FCS USAR shows that various means are available for the detection of the opening of a PORV/PSV valve. Those indications would include pressurizer quench tank pressure, temperature, or level; containment building radiation level; condenser off-gas radiation level; steam generator blowdown radiation level; containment humidity and temperature; containment sump level; and volume control tank level.
The licensee stated the proposed amendment concerning relocation of PORV/PSV acoustic position indication does not affect the means of detecting RCS leakage described above for compliance with GOC 16. The methods of detecting RCS leakage are still available.
Finally, the failure of PORV/PSV acoustic position indication and tail pipe temperature indication would not significantly challenge the operating staff's ability to respond to a DBA or transient because the abnormal operating procedures and emergency operating procedures provide symptom-based instruction to the staff for mitigating an upset condition of the plant (Le.,
- 10 decreasing pressurizer pressure, increasing quench tank pressure, temperature, or level and rising or elevated pressure relief line temperatures or flow indication from the relief line acoustic monitors). In addition, the NRC staff noted that NUREG-1432, does not list operability and SRs for the PORV/PSV position indication instrumentation.
The NRC staff noted that the proposed SR changes are consistent with the requirements in the STS except the staff raised a concern regarding the proposed change to revise the surveillance frequency of Table 3-3, Item 21, "PORV Operation" CHANNEL FUNCTIONAL TEST from monthly to refueling frequency. In the response to the NRC staff's request for additional information dated May 26,2011 (ADAMS Accession No. ML111430724), the licensee clarified in its letter dated June 30,2011, that this SR frequency (18 months) change is aligned with STS SR 3.4.11.2, "Perform a complete cycle of each PORV." The licensee also indicated that the STS SR 3.4.12.6, "Perform CHANNEL FUNCTIONAL TEST on each required PORV, excluding actuation," on a frequency of 31 days remains unchanged per FCS TS Table 3-3, Item 18, "Low-Temperature Setpoint Power-Operated Relief Valves." Therefore, the NRC staff concludes that the proposed SR changes are consistent with evaluations made during development of the STS and are, therefore, acceptable.
The NRC staff's evaluation of the licensing design basis finds that relocating the PORV/SRV position indication instrumentation TSs is consistent with the Commission's Final Policy Statement criteria in that Criterion 1 was intended to assure that TSs control those instruments specifically installed to detect RCS leakage which present a threat of Significant compromise to the RCPB. Leakage detection systems inside containment are designed with the capability of detecting leakage less than the established leakage rate limits and providing appropriate alarm of excess leakage in the control room. These actions provide adequate response before a significant break in the RCPB can occur. Thus, identifying the source of RCS leakage does not include reliance on the PORV/SRV acoustic monitors since this instrumentation is not used to identify leaks that present a threat of significant compromise to the RCPB.
The NRC staff reviewed the licensee's technical analysis of the proposed TS changes against the requirements in 10 CFR 50.36. Based on the above, the NRC staff concludes that the removal of the PORV/PSV acoustic position indication and tail pipe temperature indication from FCS TS for accident monitoring instrumentation is acceptable. In addition, the staff concludes that the proposed changes are consistent with the requirements in NUREG-1432 and the guidelines in RG 1.97. Therefore, the NRC staff concludes that deletion of the proposed PORV/PSV acoustic position indication and tail pipe temperature indication in the FCS TS and the revision of the related SR is acceptable.
The NRC staff reviewed the licensee's technical analysis of the proposed changes to the PORV/SRV position indication instrumentation for TS LCOs and SRs, including the associated footnotes, against the requirements in 10 CFR 50.36. Based on the above, the NRC staff concludes that the PORV/SRV acoustic position indication instruments do not satisfy any of the criteria in 10 CFR 50.36 for items required to be maintained in the TSs. In addition, the NRC staff concludes that the proposed changes are consistent with the requirements in the STSs.
Therefore, the NRC staff concludes that relocating the proposed PORV/SRV position indication instrumentation TS requirements and associated footnotes from the FCS TSs to a licensee-controlled document is acceptable.
- 11 3.4 Evaluation of Administrative Changes In letters dated May 31,2010, and June 1,2010 (ADAMS Accession Nos. ML101520198 and ML101530319, respectively), OPPD requested an emergency amendment to modify TS 2.15, Table 2-5 Note c to allow a one-time extension of the 7-day allowed outage time for the inoperability of Function 4 regarding safety valve acoustic position indication for RC-142 to allow repair prior to the next entry into Operating Mode 3 (Hot Shutdown) from Operating Mode 4 (Cold Shutdown). The one-time extension was applied through footnote 'i' associated with Note c. This change allowed FCS to continue power operations with inoperable safety valve acoustic position indication on safety valve RC-142. The NRC staff approved the LAR on July 12, 2010, in Amendment No. 265 (ADAMS Accession No. ML101520296). The licensee proposes to delete footnote 'i' associated with Note 'c.'
In its letter dated April 6, 2011, the licensee stated, in part, that The safety valve RC-142 flow indicator, FI-142, failed and was declared inoperable on May 26,2010. Operations personnel entered the 7-day TS LCO 2.15, Table 2-5, Item 4. On June 2, 2010, Operations personnel commenced shutdown of FCS in accordance with TS LCO 2.15, Table 2-5, Item 4, Note c. On June 2, 2010, OPPD received TS Amendment No. 265, and the TS 2.15, Table 2-5, Item 4, Note c, Hot Shutdown statement was exited. The flow indicator FI-142 remained inoperable until all post-maintenance testing was completed satisfactorily. The flow indicator, FI-142 was declared operable on June 4,2010. OPPD used the one-time extension of the 7-day allowed outage time for the inoperability of the safety valve acoustic position indication as permitted by TS Amendment No. 265; therefore, this note is no longer applicable.
The one-time extension for PSV RC-142 instrumentation, functional unit 4, is no longer applicable. The instrumentation for PSV RC-142 is required to be operable in accordance with TS 2.15, Table 2-5. Therefore, the NRC staff concludes that deletion of the footnote 'i' to Note
'c' in TS 2.15, Table 2-5 is administrative in nature and is, therefore, acceptable.
In its letter dated July 12, 2010, OPPD submitted an LAR request to add Note 'e' to TS 2.15, Table 2-5, function 1, "CEA Position Indication Systems." The licensee is requesting to re-Ietter Note 'e' to Note 'a', upon approval of both LARs.
The proposed re-Iettering of Note 'e' to Note 'a' is an administrative change in that the re-Iettering does not change any requirements in TS 2.15. The NRC staff concludes that upon approval of both LARs, this change is acceptable.
3.
3.4 NRC Staff Conclusion
The proposed changes relocate operability and SRs for the SRV position instrumentation, which consists of the PORV/SRV acoustic position indication and the PORVlsafety valve tail pipe temperature instrumentation from the Instrumentation and Control System section, TS 2.15, Table 2-5, Functions 3, 4, and 5, including the associated footnotes and the SRs TS 3.1, Table 3-3, Functions 21, 23, and 24, to the FCS USAR Based on the discussion in Section 3.3 of this safety evaluation, the NRC staff concludes that the proposed changes are acceptable.
- 12 Accordingly, future changes to the PORV/SRV acoustic position indication and the PORV/safety valve tail pipe temperature instrumentation operability and SRs will be performed under the licensee's control program and changes pursuant to the requirements of 10 CFR 50.59.
The NRC staff has determined that the deletion of footnote 'i' to Note 'c' and the re-Iettering of Note 'e' to Note 'a' in TS 2.15, Table 2-5 are administrative in nature and are, therefore, acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Nebraska State official was notified of the proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on January 25, 2011 (76 FR 4388). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The NRC staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: K. Bucholtz P. Chung Date: September 30, 2011
- SE memo dated
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NRR/LPL4/PM 22/11 OFFICE NRR/DSS/SRXB/BC OGC NRR/LPL4/BC NRR/LPL4/PM RHarper MMarkley NAME AUlses LWilkins 9/28/11 9/30/11 9/30/11 DATE 9/27/11