Letter Sequence Approval |
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MONTHYEARML20214J2911979-12-20020 December 1979 Safety Evaluation Re Preliminary Design for safety-grade Anticipatory Reactor Trips on Loss of Main Feedwater &/Or Turbine Trip Project stage: Approval ML20214J2801979-12-20020 December 1979 Safety Evaluation Re Preliminary Design for Upgrading Present control-grade Anticipatory Reactor Trip Sys for Loss of Main Feedwater & Turbine Trip to safety-grade Project stage: Approval ML20213D0761979-12-20020 December 1979 Forwards Safety Evaluation Re Preliminary Design for safety- Grade Anticipatory Reactor Trips on Loss of Main Feedwater &/Or Main Turbine Generator Project stage: Approval ML20213D0721979-12-20020 December 1979 Forwards Safety Evaluation Approving Preliminary Design for safety-grade Anticipatory Reactor Trip on Loss of Feedwater & Turbine Trip.Requests Addl Info for Final Design Approval Project stage: Approval 1979-12-20
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217G0191999-10-15015 October 1999 Safety Evaluation Concluding That Licensee Followed Analytical Methods Provided in GL 90-05.Grants Relief Until Next Refueling Outage,Scheduled to Start on 991001.Temporary non-Code Repair Must Then Be Replaced with Code Repair ML20212L0881999-10-0404 October 1999 SER Accepting Licensee Requests for Relief 98-012 to 98-018 Related to Implementation of Subsections IWE & Iwl of ASME Section XI for Containment Insp for Crystal River Unit 3 ML20212J8631999-10-0101 October 1999 Safety Evaluation Supporting Licensee Proposed Alternatives to Provide Reasonable Assurance of Structural Integrity of Subject Welds & Provide Acceptable Level of Quality & Safety.Relief Granted Per 10CFR50.55a(g)(6)(i) ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 ML20212E6911999-09-21021 September 1999 Safety Evaluation Supporting Proposed EALs Changes for Plant Unit 3.Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety ML20211H7921999-08-13013 August 1999 Safety Evaluation Supporting Amend 126 to License DPR-54 ML20210P1111999-08-0505 August 1999 SER Accepting Evaluation of Third 10-year Interval Inservice Insp Program Requests for Relief for Plant,Unit 3 ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) ML20203A4381999-02-0303 February 1999 Safety Evaluation Supporting EAL Changes for License DPR-72, Per 10CFR50.47(b)(4) & App E to 10CFR50 ML20198M7841998-12-29029 December 1998 SER Accepting Util Proposal to Use ASME Code Case N-578 as Alternative to ASME Code Section Xi,Table IWX-2500 for Arkansas Nuclear One,Unit 2 ML20196F4911998-12-0101 December 1998 SER Accepting Request for Relief ISI2-09 for Waterford Steam Electric Station,Unit 3 & Arkansas Nuclear One,Unit 2 ML20195C4841998-11-0606 November 1998 SER Accepting QA Program Change to Consolidate Four Existing QA Programs for Arkansas Nuclear One,Grand Gulf Nuclear Station,River Bend Station & Waterford 3 Steam Electric Station Into Single QA Program ML20154J2471998-10-0909 October 1998 SER Accepting Inservice Testing Program,Third ten-year Interval for License DPR-51,Arkansas Nuclear One,Unit 1 ML15261A4681998-09-0404 September 1998 Safety Evaluation Supporting Amends 232,232 & 231 to Licenses DPR-38,DPR-47 & DPR-55,respectively ML20236Q4611998-06-30030 June 1998 SER for Crystal River Power Station,Unit 3,individual Plant Exam (Ipe).Concludes That Plant IPE Complete Re Info Requested by GL 88-20 & IPE Results Reasonable Given Plant Design,Operation & History ML20248D7491998-05-28028 May 1998 Safety Evaluation Accepting Licensee Request for Relief from ASME Code Repair Requirements for ASME Code Class 3 Piping ML20217A7211998-04-17017 April 1998 Safety Evaluation Supporting Proposed Alternative for ANO-1 to Implement Code Case N-533 (w/4 H Hold Time at Test Conditions Prior to VT-2 Visual Exam) ML20216G8091998-04-10010 April 1998 Safety Evaluation Accepting Resolution of Crystal River Restart Issues Related to USI A-46 Program ML20217P8281998-04-0707 April 1998 Safety Evaluation Accepting Relief Authorization for Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250 Bolting Exam for Plants,Per 10CFR50.55a(a)(3)(i) ML20216D6111998-03-12012 March 1998 Safety Evaluation Supporting Amend 188 to License NPF-6 ML20199A1441998-01-0909 January 1998 Safety Evaluation Accepting Relief Request for Delayed Implementation of 10CFR50.55a,until 971231 or Plant Restart, Whichever Occurs First ML20199H3711997-11-19019 November 1997 SER Accepting Approving Request Relief from Requirements of Section XI, Rule for Inservice Insp of NPP Components, of ASME for Current or New 10-year Inservice Insp Interval IAW 50.55(a)(3)(i) of 10CFR50 ML20199D0561997-11-14014 November 1997 Safety Evaluation Approving Ampacity Derating Test Results for Crystal River,Unit 3 Related to GL 92-08, Thermo-Lag 330-1 Fire Barriers ML20212C3751997-10-16016 October 1997 SER Accepting Licensee Response to GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20198J7651997-10-15015 October 1997 Safety Evaluation Accepting 10-yr Interval Insp Program Plan Alternatives for Listed Plants Units ML20217D7561997-10-0101 October 1997 Safety Evaluation Concluding That Testing of Ingersoll-Dresser Pump Model 8HN194 at Test Facility Demonstrates That Crystal River Decay Heat Pumps of Same Model Can Operate at Flows of 100 Gpm for 30 Days ML20216E9921997-09-0404 September 1997 Safety Evaluation Accepting 970623 Request for Relief Re Authorization for Use of ASME Code Case N-416-1 & N-532,ISI Program for Listed Plants ML20198G8271997-08-22022 August 1997 Safety Evaluation Supporting Amend 125 to License DPR-54 ML20141H8411997-07-30030 July 1997 Safety Evaluation Accepting Use of Code Case N-508-1 for All Four Plants for Rotation of Serviced Snubbers & Pressure Relief Valves for Purpose of Testing in Lieu of ASME Code ML20148H2501997-06-0505 June 1997 Safety Evaluation Accepting Proposed Restructuring of Util Through Acquisition Of,& Merger W/Panenergy Corp ML20138K0561997-05-0505 May 1997 SER Approving Licensees IPE Process Capable of Identifying Severe Accidents & Severe Accident Vulnerabilities,For Plant,Unit 2 ML20138J0151997-05-0505 May 1997 Safety Evaluation Approving Request for Relief 95-050,Rev 1, for Plant,Unit 3 ML20138E4411997-04-30030 April 1997 Safety Evaluation on ASME Code Case N-509 for Crystal River Nuclear Plant,Unit 3 ML20140F3771997-04-28028 April 1997 Safety Evaluation Supporting Staff Evaluation of Plant, Unit 3 Nuclear Generating Plant IPE ML20134N7121997-02-20020 February 1997 Safety Evaluation Accepting Relief Request 96-04 for Plant ML20134B7091997-01-29029 January 1997 SER Accepting Fire Barrier Sys Relied by Licensee to Meet NRC Fire Protection Requirements for Following Raceway Types & Sizes ML20133N6511997-01-22022 January 1997 Safety Evaluation Accepting Licensee Request to Use Code Case N-524 as Alternative to ASME Code Section XI for Plant ML20149M6801997-01-17017 January 1997 Safety Evaluation Accepting Licensee 960807 Results of Analyses Re Operability Evaluation of Main Steam Sys W/Bent Rod Hangers at Plant ML20133D3471997-01-0606 January 1997 Safety Evaluation Supporting Amend 155 to License DPR-72 ML20149M4221996-12-12012 December 1996 Safety Evaluation Supporting Update Insvc Insp Programs to 1992 & Portions of 1993 ASME Boiler & Pressure Vessel Code, Sect XI for Licenses DPR-51,NPF-6,NPF-38,NPF-29 & NPF-47. Technical Ltr Rept Encl ML20134P8411996-11-25025 November 1996 Safety Evaluation Denying Request for Relief 96-001 Re Second 10-yr Interval ISI Program Plan,Due to Failure to Provide Basis for Impracticality ML20107F5611996-04-17017 April 1996 Safety Evaluation Providing Guidance on Submitting plant- Specific Info W/Respect to IST Program Alternatives Request 1999-09-22
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217L8931999-10-31031 October 1999 Rev 1 to BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20217G0191999-10-15015 October 1999 Safety Evaluation Concluding That Licensee Followed Analytical Methods Provided in GL 90-05.Grants Relief Until Next Refueling Outage,Scheduled to Start on 991001.Temporary non-Code Repair Must Then Be Replaced with Code Repair 3F1099-19, Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated1999-10-13013 October 1999 Part 21 Rept Re Damage on safety-grade Cable Provided to FPC by Bicc Brand-Rex Co.Damage Was Created During Cabling Process While Combining Three Conducters.Corrective Action Program Precursor Card PC99-2868 Was Initiated ML20217B0931999-10-0606 October 1999 Part 21 Rept Re Damaged Safety Grade Electrical Cabling Found in Supply on 990831.Damage Created During Cabling Process While Combining Three Conductors Just Prior to Closing.Vendor Notified of Reporting of Issue ML20212L0881999-10-0404 October 1999 SER Accepting Licensee Requests for Relief 98-012 to 98-018 Related to Implementation of Subsections IWE & Iwl of ASME Section XI for Containment Insp for Crystal River Unit 3 ML20212J8631999-10-0101 October 1999 Safety Evaluation Supporting Licensee Proposed Alternatives to Provide Reasonable Assurance of Structural Integrity of Subject Welds & Provide Acceptable Level of Quality & Safety.Relief Granted Per 10CFR50.55a(g)(6)(i) ML20212L1141999-10-0101 October 1999 Safety Evaluation Granting Request for Exemption from Technical Requirements of 10CFR50,App R,Section III.G.2.c 0CAN109902, Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With1999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20212E9031999-09-30030 September 1999 FPC Crystal River Unit 3 Plant Reference Simulator Four Year Simulator Certification Rept Sept 1995-Sept 1999 3F1099-02, Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With1999-09-30030 September 1999 Monthly Operating Rept for Sept 1999 for Crystal River,Unit 3.With ML20216J6271999-09-27027 September 1999 Rev 0 to CALC-98-R-1020-04, ANO-1 Cycle 16 Colr ML20212F5261999-09-22022 September 1999 SER Approving Request Reliefs 1-98-001 & 1-98-200,parts 1,2 & 3 for Second 10-year ISI Interval at Arkansas Nuclear One, Unit 1 ML20212E6911999-09-21021 September 1999 Safety Evaluation Supporting Proposed EALs Changes for Plant Unit 3.Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 ML20212C1501999-08-31031 August 1999 Non-proprietary Version of Rev 0 to Crystal River Unit 3 Enhanced Spent Fuel Storage Engineering Input to LAR Number 239 ML20211L1321999-08-31031 August 1999 EAL Basis Document 3F0999-02, Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With1999-08-31031 August 1999 Monthly Operating Rept for Aug 1999 for Crystal River,Unit 3.With 0CAN099907, Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with1999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Ano,Units 1 & 2. with ML20211F4281999-08-25025 August 1999 Safety Evaluation Concluding That Licensee Provided Acceptable Alternative to Requirements of ASME Code Section XI & That Authorization of Proposed Alternative Would Provide Acceptable Level of Quality & Safety ML20211B7291999-08-16016 August 1999 Rev 2 to Cycle 11 Colr ML20211H7921999-08-13013 August 1999 Safety Evaluation Supporting Amend 126 to License DPR-54 ML20210P1111999-08-0505 August 1999 SER Accepting Evaluation of Third 10-year Interval Inservice Insp Program Requests for Relief for Plant,Unit 3 ML20210U5341999-07-31031 July 1999 Monthly Operating Rept for July 1999 for Crystal River,Unit 3 ML20209F5601999-07-31031 July 1999 EAL Basis Document, for Jul 1999 0CAN089904, Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with1999-07-31031 July 1999 Monthly Operating Repts for July 1999 for Ano,Units 1 & 2. with ML20210K8831999-07-29029 July 1999 Non-proprietary Addendum B to BAW-2346P,Rev 0 Re ANO-1 Specific MSLB Leak Rates 0CAN079903, Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with1999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Ano,Units 1 & 2. with ML20210U5411999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for Crystal River,Unit 3 3F0799-01, Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With1999-06-30030 June 1999 Monthly Operating Rept for June 1999 for Crystal River,Unit 3.With ML20207E7231999-06-0202 June 1999 Safety Evaluation Authorizing Proposed Alternative Exam Methods Proposed in Alternative Exam 99-0-002 to Perform General Visual Exam of Accessible Areas & Detailed Visual Exam of Areas Determined to Be Suspect ML20196A6251999-05-31031 May 1999 Non-proprietary Rev 0 to TR BAW-10235, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs ML20196A0191999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Arkansas Nuclear One,Units 1 & 2.With ML20210U5601999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for Crystal River,Unit 3 3F0699-07, Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With1999-05-31031 May 1999 Monthly Operating Rept for May 1999 for Crystal River,Unit 3.With ML20195C6271999-05-28028 May 1999 Non-proprietary Rev 0 to Addendum to Topical Rept BAW-2346P, CR-3 Plant Specific MSLB Leak Rates ML20195D1991999-05-28028 May 1999 Probabilistic Operational Assessment of ANO-2 SG Tubing for Cycle 14 ML20196L2031999-05-19019 May 1999 Non-proprietary Rev 0 to BAW-2346NP, Alternate Repair Criteria for Tube End Cracking in Tube-to-Tubesheet Roll Joint of Once-Through Sgs ML20206M7711999-05-11011 May 1999 SER Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 ML20195D1901999-05-0606 May 1999 Annual Rept ML20210U5631999-04-30030 April 1999 Revised Monthly Operating Rept for Apr 1999 for Crystal River,Unit 3 3F0599-04, Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With1999-04-30030 April 1999 Monthly Operating Rept for Apr 1999 for Crystal River Unit 3.With 0CAN059903, Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with1999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Ano,Units 1 & 2. with ML20206F0691999-04-29029 April 1999 Safety Evaluation Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205M6941999-04-12012 April 1999 Safety Evaluation Granting Relief for Second 10-yr Inservice Inspection Interval for Plant,Unit 1 ML20205R6351999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Ano,Units 1 & 2. with ML20205D6061999-03-31031 March 1999 Safety Evaluation Supporting Licensee Proposed Approach Acceptable to Perform Future Structural Integrity & Operability Assessments of Carbon Steel ML20204D9661999-03-31031 March 1999 Non-proprietary Rev 1,Addendum a to BAW-2342, OTSG Repair Roll Qualification Rept 3F0499-04, Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With1999-03-31031 March 1999 Monthly Operating Rept for Mar 1999 for Crystal River Unit 3.With ML20205D4711999-03-26026 March 1999 SER Accepting Util Proposed Alternative to Employ Alternative Welding Matls of Code Cases 2142-1 & 2143-1 for Reactor Coolant System to Facilitate Replacement of Steam Generators at Arkansas Nuclear One,Unit 2 ML20204B1861999-03-15015 March 1999 Safety Evaluation Authorizing Licensee Request for Alternative to Augmented Exam of Certain Reactor Vessel Shell Welds,Per Provisions of 10CFR50.55a(g)(6)(ii)(A)(5) 3F0399-04, Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease1999-03-10010 March 1999 Special Rept 99-01:on 990310,discovered Containment Tendons That Required Grease Addition in Excess of Prescribed Limits During Recent Insp Activites.Six Tendons Were Refilled with Appropriate Amount of Grease 1999-09-30
[Table view] |
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- 'o,, UNITED STATES 8' '
- c. NUCLEAR REGULATORY COMMISSION
'f E WASHINGTON, D. C. 20655
\,...../
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR RFACTOR REGULATION OF PRELIMINARY DESIGN FOR SAFETY-GRADE ANTICIPATORY REACTOR TRIPS (ARTS) ON LOSS OF MAIN FEEDWATER AND/OR TURBINE TRIP FOR _,
DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNITS NOS. 1, 2 AND 3 ,
DOCKETS NOS. 50-269, 270 AND 287 )
SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SEC0 NUCLEAR GENERATING STATION .- )
DOCKET NO. 50-312.
l ARKANSAS POWER & LIGHT COMPANY ARKANSAS NUCLEAR OllE, UNIT 1 DOCKET NO. 50-313 FLORIDA POWER CORPORATION l
CRYSTAL RIVER NUCLEAR GENERATING STATION UNIT NO. 3 DOCKET NO. 50-302 I. BACKGROUND Following the accident at Three Mile Island Unit 2, an assessment of feedwater transients in the Babcock and Wilcox (B&ti) designed pressurized water reactors was performed. The results of that review were reported in NUREG-0560 This i report highlighted a concern regarding the challenges to the power-operated l relief valves (PORV) in the B&W design. In response to I&E Bulletin 79-05B, i
the licensees lowered the existing setpoint for the high pressure reactor trip and raised the setpoint of the PORV. By inverting these setpoints the challenge rate to the PORY and thus the chance of it not reseating following actuation was reduced.
To provide additional margin to the automatic opening setpoint, the licensees pmposed design provisions for direct reactor trip on loss of main feedwater or turbine trip. This design modification was approved and incorporated as part of the required actions of the Commission's Confirmatory Shutdown Orders issued in May 1979. In order to achieve a timely implementation it was deter-mined that a " control-grade" design was sufficient for the short-term. In the short-term the licensees implemented hardwired, control-grade trips inde- ,
pendent of the reactor protection system (RPS). For the long-term, the trips were to be upgraded to safety-grade and become part of the RPS.
As part of the long-term requirements, each licensee submitted their proposed designs for safety-grade reactor trips to be incorporated in the existing RPS.
8705270746 791220 ~
PDR P ADOCK 05000269 ppg,
These submittals are listed as References 1 through 8 of Attachment 2 to this evaluation. The following evaluation is applicable to Oconee Units 1, 2 and 3; Arkansas Nuclear One, Unit 1, Rancho Seco and Crystal River Unit 3.
II. EXISTING RPS The existing plant RPS includes four redundant and independent channels.
Each channel has its own independent input sensors that are physically and electrically separated from the sensors of the other channels. The present trip conditions that are monitored by these sensors and channels include:
- 1. Nuclear power / flux (high)
- 2. Nuclear power based on flow (high)
- 3. Nuclear power based on reactor coolant
- 4. Reactor coolant system pressure (high) pump status (high)*
- 5. Reactor coolant system pressure (low)
- 6. Reactor coolant system pressure based on temperature (low)
- 7. Reactor coolant temperature (high)
- 8. Reactor building (containment) pressure (high)
Within the RPS cabinets each of the four channels contain a logic string of the above inputs. Any individual actuation will cause the logic string to trip and actuate a trip relay. The trip relays of the four channels form a two-out-of-four coincident logic to open the reactor trip breakers.
III. DESCRIPTION OF PROPOSED DESIGN The licensees have proposed ARTS which will actuate on turbine trip and/nr main feedwater pump trip. These anticipatory trips provide additional protection and conservatism beyond that provided by the existing RPS.
No credit is taken for these trips in the FSAR Chapter 15 analyses. Pre-viously existing and diverse parameters will cause a reactor trip should these proposed trips fail to function.
The proposed trips are to be incorporated into the existing RPS. They each contain four redundant and independent inputs to interface with the four i RPS channels.
l The turbine trip is to be sensed by four independent pressure switches.
The feedwater pump trip is similarly sensed (for each pump) by four inde-pendent pressure switches. The logic is arranged such that both main feedwater pumps must be tripped to cause reactor trip.
- The trip system in Crystal River 3 does not monitor for this condition.
j ! I
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A reactor flux level premissive (bypass) is provided to facilitate startup and shutdown of the plant. This pemissive automatically blocks the reactor trip upon turbine trip or main feedwater pump trip when reactor power is decreased below 20% power. During power, escalation, as reactor power increases to above 20%, the bypass is automatically removed and the reactor trips are reinstated. This flux signal is part of the existing RPS and therefore, is-implemented in the RPS four channel arrangement.
The pressure switch inputs are routed to the RPS cabinets for interface in the present logic trip strings through added RPS modules. The flux bypass is also implemented through a new RPS module. The RPS modules will contain contact buffers, bistables, and auxiliary relays as required. All additional equip-ment will be designed in accordance with the design bases of the existing RPS and will conform with the acceptance criteria and design requirements of the RPS.
The licensees have stated that the cabinet mounted equipment to be supplied by B&W will be fully testable from the RPS cabinets. The equipment will have l provisions for simulating input signals and verifying the proper response of the RPS channel. This testing will be similar to that presently performed on the RPS and will be integrated into the periodic testing of the cabinets.
With respect to environmental qualification, all equipment associated with the ARTS is located outside containment. In addition, the licensees have i stated that the new RPS modules to be used have been qualified for use in i B&W safety systems. Also, the pressure switch sensors will be equivalent to switches presently used in other plant safety applications. 3 i
With respect to seismic qualification, the licensees have stated that all equipment will be seismically qualified (with the exception of some equipment located in non-seismic Category I areas - See Section IV).
Existing RPS power supplies, flux signals, interlock circuits, and indicators l will be used as required by the added equipment.
IV. EVALUATION In perfoming our evaluation of the licensees' proposed designs, we utilized information provided by the licensees listed as References 1 though 8 of this evaluation.
The infomation presented in these submittals addressed only the preliminary design for the safety-grade ARTS. Included in the information is a brief description of the system and simplified logic and schematic diagrams. i We have concluded that the licensees have identified the design bases and criteria for this additional equipment, as well as provided a preliminary design description; however, the design details are not sufficiently complete to make a determination that the design satisfies the identified criteria. -
Therefore, as additional design details are developed and prior to operation $
of the new equipment, we will require that the licensees submit the final
design for our review and approval. The final design shall include final logic diagrams, electrical schematic diagrams, piping and instrumentation diagrams and location layout drawings.
In our evaluation of the ARTS, we concentrated on the adequacy of the design approach as it pertains to the existing RPS. That is, we detennined whether the trip would meet the requirements of a safety-grade system and whether its addition would in any way degrade the existing RPS.
The " Design Basis" and " Requirements" of the ARTS as required by IEEE 279-1971 are to be equivalent to those of the existing RPS with the exception that the input sensors will not conform to seismic requirements. We conclude that this is acceptable based on the anticipatory nature of these trips and that other fully qualified trips serve as back up protection. There is a related concern with the location of these inputs, which needs to be addressed in more detail by the licensees. Specifically, for those sensors located in non-seismic areas which have previously not contained RPS inputs, we will require that their installation (including circuit routing) be analyzed to demonstrate that the effects of credible faults (i.e., grounding, shorting, application of high voltage, or electromagnetic interference) or failures in these areas will not be propagated back to the RPS and degrade the RPS performance or oper-ability. This will require that specific provisions (e.g., conduit) be -
utilized to keep the circuits sufficiently separated. Therefore, we will require that the licensees submit such an analysis, along with the final design, prior to the operation of the ARTS.
With respect to equipment qualification, the licensees have supplied " Seismic and Environmental Qualification Sumary Reports" for the equipment to be supplied by B&W. The balance of the qualification information is not yet available. Therefore, we require that the information for the remaining equipment be submitted when available. In addition, as part of the final design package, we require information which demonstrates that the envion-mental test conditions bound the actual worst case accident conditions expected at the installed locations. The detailed test procedures and test data will be examined as part of the review of the final design.
The ARTS testability, particularly with respect to the sensors (paragraph 4.9 of IEEE 279-1971), is not sufficiently addressed for us to conclude that adequate provisions are being incorporated to accomplish the RPS channel tests.
Therefore, we will require that the licensees include provisions to perform channel functional tests at power on a periodic basis (i.e., during RPS monthly surveillance tests).
As part of the final design submittal, we will require that the licensees provide the RPS check-out procedure which will demonstrate both the opera-bility of the new trips and the continued operability of the previous RPS.
V. CONCLUSION The licensees have identified the design bases and design requirements for the
" Anticipatory Reactor Trips". They have also provided a preliminary design description. We have concluded that this identification along with the preliminary design description provides sufficient bases for approval of the preliminary design.
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4 In order to approve the final design, as soon as possible, we shall require
' the licensee to submit the above identified information as soon as it is available. .In addition, a site visit may be required and would be coordinate with our Office of Inspection and Enforcement.
Attachments:
- 1. Summary.of Infonnation Needed for Final Design Approval
- 2. References o
i l
i 4
i I
i i
py,A Su#'^ Attachment 1 ASOL
, SUMARY OF INFORMATION NEEDED FOR FINAL DESIGN APPROVAL i
Page nunber in SER Requiremebt 5- 7
- 3/ 4. The final design submittal should include the final logic i diagrams, electrical schematic diagrams, piping and instru-mentation diagrams and location layout drawings.
- 4. For sensors located in non-seismic areas which have not
! previously contained RPS inputs, perfom and submit an analysis which shows that the installation (including circuit routing) is designed such that the effects of 1 credible faults (i.e., grounding, shorting, ,
application of high voltage, or electromagnetic inter-ference) or failures in these areas could not be propa-gated back to the RPS and degrade the RPS performance or operability, t 4. Submit " Seismic and Environmental Qualification Summary Reports" for the equipment which has not been previously y' submitted. In addition, we require that you demonstrate that the environmental test conditions bound the actual +
. worst case accident conditions expected at the installed
! locations.
- 4. Assure that the ARTS testability includes provisions to
- i. perfom channel functional tests at power. Testing of this
' circuitry is to be included in the RPS monthly surveillance tests. ;
V
- 5. Include in the final design submittal the RPS check-out j procedure which will demonstrate both the operability of >
the new trip circuitry and the continued operability of 5
the previous RPS.
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a-~ c Attachment 2 REFERENCES - ANTICIPATORY REACTOR TRIP t
OCONEE Units 1, 2 and 3 (NRC), dated May 21, 1979
- 1. Letter from W. O. Parker (DUKE) to H. R. Denton
Subject:
Response to IE Bulletin 79-05B.
- 2. Letter from W. O. Parker (DUKE) to H. R. Denton (NRC), dated October 5,1979
Subject:
Response to NRC letter dat.ed September 7,1979 (Request for Additional 2 Information).
ARKANSAS NUCLEAR ONE, Unit 1 2
- 3. Letter from D. C. Trimble (AP&L) to K. V. Seyfrit (NRC), dated May 21,1979
Subject:
Response to IE Bulletin 79-05B.
- 4. Letter from D. C. Trimble (AP&L) to R. W. Reid (NRC), dated October 8,1979
Subject:
Response to NRC letter dated September 7,1979 (Request for Additional Information).
RANCHO SECO
- 5. Letter from W. C. Walbridge (SMUD) to R. H. Engelken (NRC), dated May 21, 1979
Subject:
Response to IE Bulletin 79-05B.
- 6. Letter from J. J. Mattimoe (SMUD) to R. W. Reid (NRC), dated October 5,1979
Subject:
Response to NRC letter dated September 7,1979 (Request for Additional
, Information).
CRYSTAL RIVER, Unit 3
- 7. Letter from W. P. Stewart (FPC) to J. P. O'Reilly (NRC), dated May 21,1979
Subject:
Response to IE Bulletin 79-05B.
- 8. Letter from W. P. Stewart (FPC) to R. W. Reid (NRC), dated October 2, 1979
Subject:
Response to NRC letter dated September 7,1979 (Request for Additional
. In formation) .
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