Sser Accepting SPDS Contingent Upon Resolution of Concerns Re Maint & Improvement of Placement & Visual Access of Containment Isolation Panel & Minor Human Factors Engineering ConcernsML20214J852 |
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Yankee Rowe |
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11/18/1986 |
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NRC |
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ML20214J423 |
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NUDOCS 8612020004 |
Download: ML20214J852 (8) |
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Category:SAFETY EVALUATION REPORT--LICENSING & RELATED ISSUES
MONTHYEARML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20211J3361999-08-27027 August 1999 Safety Evaluation Supporting Amend 152 to License DPR-3 ML20207F9491999-03-0505 March 1999 Safety Evaluation Supporting Amend 151 to License DPR-3 ML20202H5871999-02-0303 February 1999 Safety Evaluation Supporting Amend 150 to License DPR-3 ML20249A7901998-06-17017 June 1998 Safety Evaluation Supporting Amend 149 to License DPR-3 ML20237F1671993-02-19019 February 1993 Safety Evaluation Supporting Amend 147 to License DPR-3 ML20058F2201990-11-0202 November 1990 Safety Evaluation Accepting Util Response to Generic Ltr 83-28 Re post-trip Review - Data & Info Capability ML20058C4061990-10-22022 October 1990 Safety Evaluation Supporting Amend 137 to License DPR-3 ML20059G2411990-09-0606 September 1990 Safety Evaluation Supporting Amend 135 to License DPR-3 ML20058L6651990-08-0202 August 1990 Safety Evaluation Supporting Amend 134 to License DPR-3 ML20058L0321990-08-0202 August 1990 Safety Evaluation Supporting Amend 133 to License DPR-3 ML20055C8601990-06-18018 June 1990 Safety Evaluation Supporting Amend 132 to License DPR-3 ML20248H7391989-10-0303 October 1989 Safety Evaluation Not Accepting Procedure Generating Program for Plant.Program Should Be Revised to Reflect Items Described in Section 2 of Rept.Revision Need Not Be Submitted to NRC ML20247F1431989-09-0707 September 1989 Safety Evaluation Supporting Amend 124 to License DPR-3 ML20247E6831989-08-31031 August 1989 Safety Evaluation Supporting Amend 123 to License DPR-3 ML20246F2771989-07-11011 July 1989 Safety Evaluation Supporting Mods to ECCS Evaluation Model, Including Changes to FLECHT-based Reflood Heat Transfer Correlation,Steam Cooling Model & post-critical Heat Flux Heat Transfer Model ML20195D6701988-11-0101 November 1988 Safety Evaluation Supporting Amend 120 to License DPR-3 ML20205G1961988-10-25025 October 1988 Safety Evaluation Supporting Amend 119 to License DPR-3 ML20204G4871988-10-17017 October 1988 Safety Evaluation Supporting Amend 118 to License DPR-3 ML20205C4061988-10-14014 October 1988 Safety Evaluation Supporting Amend 117 to License DPR-3 ML20207L7051988-10-12012 October 1988 Safety Evaluation Supporting Amend 116 to License DPR-3 ML20207E8151988-08-0505 August 1988 Safety Evaluation Supporting Amend 115 to License DPR-3 ML20151M4911988-07-29029 July 1988 Safety Evaluation Supporting Amend 114 to License DPR-3 ML20151K3801988-07-25025 July 1988 Safety Evaluation Supporting Amend 113 to License DPR-3 ML20151K8571988-07-19019 July 1988 Safety Evaluation Supporting Amend 112 to License DPR-3 ML20153A8661988-06-29029 June 1988 Safety Evaluation Accepting Util Proposed Reflood Steam Cooling Model ML20196K2741988-06-28028 June 1988 Safety Evaluation Supporting Amend 111 to License DPR-3 ML20195K1501988-06-17017 June 1988 Safety Evaluation Supporting Amend 110 to License DPR-3 ML20195C5851988-06-13013 June 1988 Safety Evaluation Supporting Amend 109 to License DPR-3 ML20155K5141988-06-0909 June 1988 Safety Evaluation Supporting Amend 108 to License DPR-3 ML20154J7661988-05-18018 May 1988 Safety Evaluation Supporting Amend 107 to License DPR-3 ML20216J4081987-06-26026 June 1987 Safety Evaluation Supporting Amend 106 to License DPR-3 ML20216C1111987-06-18018 June 1987 Safety Evaluation Granting Three of Seven Requests Submitted by Util for Relief from Inservice Insp & Testing Requirements.Four Requests Withdrawn,Per 870122,0410 & 0507 Ltrs ML20215C5881987-06-0404 June 1987 Safety Evaluation Supporting Util 860505,870402,& 0506 Submittals Re Seismic Reevaluation of Plant.Concludes That Foundation Soils Under Reactor & Under Vapor Container Have Adequate Strength to Support Seismic Load from Earthquake ML20213G9161987-05-13013 May 1987 Safety Evaluation Supporting Amend 105 to License DPR-3 NUREG-0825, Safety Evaluation Supporting Util 840709,1231 & 851024 Repts Re Evaluation of Plant for Wind & Tornado Events as Requested in Integrated Plant Safety Assessment Rept, Sections 4.5 & 4.8.Risk from Wind/Tornado Events Assessed1987-05-13013 May 1987 Safety Evaluation Supporting Util 840709,1231 & 851024 Repts Re Evaluation of Plant for Wind & Tornado Events as Requested in Integrated Plant Safety Assessment Rept, Sections 4.5 & 4.8.Risk from Wind/Tornado Events Assessed ML20213D9671987-05-0707 May 1987 Safety Evaluation Supporting Amend 104 to License DPR-3 ML20207S6231987-03-10010 March 1987 Safety Evaluation Supporting Util 860122,0812,1028 & 870204 Submittals Re Fracture Toughness Requirements for Protection Against PTS Events ML20211N5881987-02-19019 February 1987 Safety Evaluation Re First Level Undervoltage Protection Testing.Testing Unnecessary ML20211L3951987-02-17017 February 1987 Safety Evaluation Supporting Amend 103 to License DPR-3 Re Max Nominal Enrichment of Fuel ML20207N8811987-01-0707 January 1987 Safety Evaluation Supporting Amend 102 to License DPR-3 ML20207N4261987-01-0606 January 1987 Safety Evaluation Supporting Amend 101 to License DPR-3 ML20207J9451986-12-30030 December 1986 SER Accepting Util 831105 & 850709 Responses to Generic Ltr 83-28,Item 2.1 (Part 2), Vendor Interface Program - Reactor Trip Sys Components ML20215E1201986-12-0909 December 1986 Safety Evaluation Supporting Util 830419 & 0830,840119, 851022 & 860930 Responses Re Conformance to Reg Guide 1.97. Plant Design Acceptable W/Exception of Neutron Flux Variable ML20214X3391986-12-0101 December 1986 Safety Evaluation Supporting Amend 100 to License DPR-3 ML20214J8521986-11-18018 November 1986 Sser Accepting SPDS Contingent Upon Resolution of Concerns Re Maint & Improvement of Placement & Visual Access of Containment Isolation Panel & Minor Human Factors Engineering Concerns ML20215E6471986-10-0202 October 1986 Safety Evaluation Supporting Util Requests for Exemption from Specific Requirements in App R to 10CFR50.Existing Fire Protection Provides Level of Protection Equivalent to Technical Requirements of App R ML20210S1791986-09-23023 September 1986 Safety Evaluation Supporting Amend 99 to License DPR-3 ML20212Q1151986-08-27027 August 1986 Safety Evaluation Supporting Util 830412 Proposal to Provide Integrated Safe Shutdown Sys Which Could Be Used for Safe Shutdown in Event of Fire at Facility ML20212N0161986-08-20020 August 1986 Safety Evaluation Supporting Amend 98 to License DPR-3 1999-08-27
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217C1311999-10-0808 October 1999 Safety Evaluation Supporting Amend 153 to License DPR-3 ML20211J5111999-08-31031 August 1999 Rev 29 to Yankee Decommissioning QA Program ML20211J3361999-08-27027 August 1999 Safety Evaluation Supporting Amend 152 to License DPR-3 ML20209D5391999-06-22022 June 1999 Rev 29 to Yaec Decommissioning QA Program ML20207F9491999-03-0505 March 1999 Safety Evaluation Supporting Amend 151 to License DPR-3 ML20202H5871999-02-0303 February 1999 Safety Evaluation Supporting Amend 150 to License DPR-3 ML20154P9691998-10-16016 October 1998 Rev 28 to Yankee Atomic Electric Co Decommissioning QA Program ML20249A7901998-06-17017 June 1998 Safety Evaluation Supporting Amend 149 to License DPR-3 ML20216C4581998-02-27027 February 1998 Response to NRC Demand for Info (NRC OI Rept 1-95-050) ML20203L1931998-02-25025 February 1998 Duke Energy Corp,Duke Engineering & Svcs,Inc,Yankee Atomic Small Break LOCA Technical Review Rept ML20203L2451998-02-23023 February 1998 Assessment Rept of Engineering & Technical Work Process Utilized at De&S Bolton Ofc ML20203L1621998-02-18018 February 1998 Rept of Root Cause Assessment Review ML20203L2691998-02-16016 February 1998 Duke Engineering & Svcs Assessment Process Review Rept ML20199B4601998-01-20020 January 1998 Special Rept:On 980105,meteorological Monnitoring Instrumentation for Air Temp Delta T Inoperable for More than 7 Days.Caused by Breakdown in Wiring Between Junction Box at 199 Foot Level.Wiring Replaced ML20203J3001997-12-31031 December 1997 Ynps 1997 Annual Rept ML20217N0981997-08-21021 August 1997 LER 97-S02-00:on 970725,discovered Uncontrolled Safeguards Documents.Caused by Personnel Error.Matls Retrieved & Stored in Safeguards Repositories ML20210H0991997-08-0707 August 1997 LER 97-S01-00:on 970709,potential Compromise of Safeguards Info Occurred.Caused by Human error.Stand-alone Personal Computer & Printer Not Connected to Network,Have Been Located within Text Graphics Svc Dept ML20149K7781997-07-24024 July 1997 Special Rept:On 970520 & 0714,air Temp Delta T Channel Indicated Temp Difference Between Top & Bottom of Meteorological Tower.Caused by Reversed Input Wiring to Channel.Restored Air Temp Delta T Channel Operability ML20141E4671997-05-30030 May 1997 Rev 28 to Operational QA Program ML20135C8461996-12-31031 December 1996 Yankee Nuclear Power Station 1996 Annual Rept ML20132G6771996-12-20020 December 1996 Rev 27 to YOQAP-I-A, Operational QA Program ML20058N4771993-12-20020 December 1993 Rev 0.0 to Yankee Nuclear Power Station Decommissioning Plan ML20059K8491993-12-15015 December 1993 Clarifications to Pages 2,41,43 & 44 of 44 in Section I, Organization of YOQAP-I-A,Rev 24, Operational QA Program ML20059C5011993-10-29029 October 1993 Special Rept:On 931019,meteorological Instrumentation Channel for Delta T Declared Inoperable.Caused by Ceased Aspirator Motor Located at Top of Tower.Motor Replaced ML20056H1741993-06-10010 June 1993 Preliminary Assessment of Potential Human Exposures to Routine Tritium Emissions from Yankee Atomic Electric Co Nuclear Power Facility Located Near Rowe,Ma ML20237F1671993-02-19019 February 1993 Safety Evaluation Supporting Amend 147 to License DPR-3 ML20125C7161992-12-0707 December 1992 Part 21 Rept Re Possibility for Malfunction of Declutching Mechanisms in SMB/SB-000 & SMB/SB/SBD-00 Actuators. Malfunction Only Occurs During Seismic Event.Balanced Levers May Be Purchased from Vendor.List of Affected Utils Encl ML20198D2541992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Recertification Program ML20198D2481992-05-13013 May 1992 Yankee Nuclear Power Station Certified Fuel Handler Initial Certification Program ML20062H1981990-11-30030 November 1990 Plant Specific Fast Neutron Exposure Evaluations for First 20 Operating Fuel Cycles of Yankee Rowe Reactor ML20058H2841990-11-0303 November 1990 Special Rept:On 901101,control Rod 24 Found Disconnected from Drive Shaft.Drive Shaft Latching Will Be Initiated ML20058F2201990-11-0202 November 1990 Safety Evaluation Accepting Util Response to Generic Ltr 83-28 Re post-trip Review - Data & Info Capability ML20062E8331990-10-31031 October 1990 Monthly Operating Rept for Oct 1990 for Yankee Atomic Power Station ML20058G1471990-10-31031 October 1990 Vol 2 to Star Methodology Application for PWRs Control Rod Ejection Main Steam Line Break ML20058C4061990-10-22022 October 1990 Safety Evaluation Supporting Amend 137 to License DPR-3 ML20062B6751990-09-30030 September 1990 Monthly Operating Rept for Yankee Atomic Power Station for Sept 1990 ML20059G2411990-09-0606 September 1990 Safety Evaluation Supporting Amend 135 to License DPR-3 ML20059E3071990-08-31031 August 1990 Safety Assessment of Yaec 1735, Reactor Pressure Vessel Evaluation Rept for Yankee Nuclear Power Station. Detailed Plan of Action W/Listed Elements Requested within 60 Days After Restart to Demonstrate Ability to Operate Longer ML20059E8001990-08-31031 August 1990 Monthly Operating Rept for Aug 1990 for Yankee Atomic Power Station ML20058P7841990-08-14014 August 1990 Part 21 Rept Re Misapplication of Fluorolube FS-5 Oil in Main Steam Line Pressure Gauges.All Four Indicators Replaced W/Spare Gauges Which Utilize High Temp Silicone Oil ML20058N6581990-08-13013 August 1990 Special Rept Re Diesel Fire Pump & Tank Inoperable for Greater than Seven Days for Draining,Cleaning & Insp.During Period Redundant Pumping Capacity Available Via Two Remaining Electric Driven Fire Pumps ML20058L0321990-08-0202 August 1990 Safety Evaluation Supporting Amend 133 to License DPR-3 ML20058L6651990-08-0202 August 1990 Safety Evaluation Supporting Amend 134 to License DPR-3 ML20056A1961990-08-0101 August 1990 Special Rept:Two Fire Pumps Inoperable at Same Time.Caused by Necessity to Accomplish Surveillance to Verify Capability to Start Pump on Emergency Diesel Generator 3 & Planned 18-month Insp of Diesel Per Tech Specs ML20055G6801990-07-31031 July 1990 Yankee Plant Small Break LOCA Analysis ML20055G7011990-07-31031 July 1990 Yankee Nuclear Power Station Core 21 Performance Analysis ML20055E1591990-07-31031 July 1990 Reactor Pressure Vessel Evaluation Rept ML20055J3221990-07-25025 July 1990 Decommissioning Funding Assurance Rept & Certification ML20055G7051990-07-19019 July 1990 Rev 0 to Yankee Cycle 21 Core Operating Limits Rept ML20055F6751990-06-30030 June 1990 Monthly Operating Rept for June 1990 for Yankee Atomic Power Station 1999-08-31
[Table view] |
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ENCLOSURE 1 9 I SUPPLEMENTAL SAFETY EVALUATION REPORT INPUT FOR YANKEE R0WE NUCLEKR POWEP STATION SAFETY PARAMETER DISPLAY SYSTEM I. INTRODUCTION All holders of operating licenses issued by the Nuclear Regulatory Commission (licensees) and applicants for an operating license must provide a Safety Para-neter Display System (SPDS) in the control room of their plant. The Commission
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approved requirements for the SPDS are defined in Supplenent I to NUREG-0737 II. BACKGROUND The staff's Safety Evaluation Report (SER) on Yankee Rowe's SPDS was transmitted to the licensee December 17, 1984 (Ref. 1). The SER was based on the staff's re-view of the Yankee Atomic Electric Company (YAEC) Safety Analysis Report. That review identified several questions and concerns but concluded that it was accept-able for the licensee to continue implementing the SPDS progran. YAEC responded to the questions in letters dated April 8 and September 3,1985 (Refs. 2 and 3).
III. EVALUATION The staff, assisted by consultants from Science Applications International Corpora-tion (SAIC),conductedanonsiteauditoftheinstalledSPDS, April 22-23, 1986.
The purpose of the audit was to confirm the closure of the open itens from the earlier review and to ascertain that the SPDS had been installed in accordance with the licensee's plan and was functioning properly. The results of the evalu-ation are summarized below and the contractor's Technical Evaluation Report (TER) l is enclosed to provide further detail. The staff agrees with the technical posi-tions and conclusions contained in the TER.
8612O20004 861118 j DR ADOCK 05000 29
-2 A. PARAMETER SELECTION Based on the submittal by YAEC of April 8,1985 and the results of the onsite audit, the staff concludes; (1) that the basis for SPDS parameter selection was appropriate, (2) that the Yankee Rowe critical safety functions provide information equivalent to the NUREG-0737, Supplement I critical safety functions, and (3) that the para-meters selected for SPDS, including containment isolation, provide a complete set of critical safety function information. However, since the containment isolation status display is separate from the SPDS monitor, but is viewable from the location of the SPDS monitor, the staff needs a commitment from the licensee that the relative position, orientation and visual access of the containment isolation status panel with respect to the SPDS station will be maintained or improved. With this con-mitment, the licensee will meet the NUREG-0737, Supplement 1 SPDS requirement for parameter selection.
B. SYSTEM DESIGN The Yankee Rowe SPDS has been defined as the ten emergency operations CRT formats displayed on either, or both, of two CRT monitors. Selection of each display format is accomplished directly by dedicated single action pushbuttons at each monitor.
Essential and non-essential containment isolation status is displayed on a separate panel.
! Display Configuration Staff evaluation of the technical content of the SPDS displays indicated that the safety functions and associated parameters are correctly configured. Evaluation
of the containment isolation display indicated that full or partial isolation can be satisfactorily determined by pattern recognition from the position of either SPDS monitor.
Display Validity ,-
Most SPDS parameters are fed by redundant signals. The algorithms used to support data validity for these redundant inputs are appropriate.
When used alone for single inputs, signal range checks can provide misleadino information. Consideration should be given to developino single input validity algorithms using related variables.
Invalid data are indicated nn the CRT by parameter values displayed in white.
Since a white parameter value can result from several causes (e.g., input failure, offscale, channel value differences), confusion could result unless the operator is trained to always seek other sources of information anytime a white data value appears.
A check of the data validity in the control room during the audit confirmed that the process worked correctly with one exception. The licensee has committed to find the error in the source range indication and correct the problem.
Maintenance and Configur.. tion Control The audit team evaluated the software maintenance and configuration control pro-cess and documentation, and detemined that the current procedures are appropriate to permit satisfactory maintenance and modifications.
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i Security System access is adequately secured by use of passwords to open remote ports after approval of the shift technical advisor.
Isolation Devices The isolation devices protecting the safety related systems from SPDS were approved previously (Ref. 4). The onsite audit confirmed their proper installation in a mild environment.
C. SYSTEM VERIFICATION AND VALIDATION The verification and validation plan for the Yankee Rowe SPDS is documented in NSAC-61. The audit team review of the plan, and verification and validation re-sults, indicate that the process was comprehensively performed.
D. HUMAN FACTORS ENGINEERING With three minor exceptions, the Yankee Rowe SPDS displays good human factors engineering. The three exceptions are as follows: (1) the cyan color has a low contrast ratio with white; (2) the red indicator lights above each of the SPDS
( controls serve no apparent function; (3) there is no audible alarm signal associ-ated with the SPDS alerts, i
E. USE OF SPDS IF OPERATION Durina the onsite audit, licensed operators and other operations and trainino per-sonnel were interviewed, and the operator training program was reviewed. The SPDS operation was demonstrated during normal plant operation. Display call-up response times were observed and system availability calculations were reviewed.
In summary, the Yankee Rowe SPDS is reliable, available, and useful to the opera-tors in both normal and emergency operations. The SPDS has been designed to function as an integrated tool with the upgraded Emergency Operating Procedures and the training program provides for operation both with and without the SPDS available.
IV. CONCLUSIONS The staff conclusions regarding each of the eight SPDS requirements of Supplement 1 to NUREG-0737, based on review of the reference documentation and the attached l
i TER, and on the onsite audit of April 22-23, 1986, are presented below:
- 1. The SPDS, along with the containment isolation status panel provide a concise display of the critical plant variables.
- 2. The SPDS is conveniently located as long as YAEC commits to main-taining or improving the relative position, orientation, and visual access of the containment isolation status display with respect to i
- 3. The SPDS is continuously displayed in the control room.
- 4. The SPDS is reliable. However, the licensee should evaluate the need to develop more sophisticated data validity algorithms for single input parameters.
- 5. SPDS electrical and electronic isolation devices are acceptable for interfacing SPDS with safety systems.
- 6. The SPDS adequately incorporates human factors engineering princi-ples, with the following minor exceptions, which the licensee should address:
o Source range power during power operations indicates a valid reading; o Cyan color used for parameter values has low contrast to the white used for invalid data coding; o Red indicator lights on SPDS control panels serve no function; o There is no audible alarm signal associated with SPDS alerts;
6 o The licensee has not provided NRC with a written response to the DCRDR HEDs related to SPDS;
- 7. The SpDS and containment isolation status panel provide the minimum information needed to determine plant safety s'tatus with respect to:
o Reactivity control o Reactor core cooling and heat removal from the primary systen o Reactor coolant system inventory o Radioactivity control o Containment conditions.
- 8. SPDS procedures and operator training, addressing actions with and without the SPDS, have been implemented.
In sunnary, the staff has determined that the SPDS will meet the requirements of NUREG-0737, Supplement 1, when the human factors engineering concerns identified above are resolved, and a commitment is established regarding the relative location of SPDS with respect to the containment isolation status l panel.
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REFERENCES
- 1. Letter from NRC to YAEC, Revised SER for Yankee Rowe Safety Parameter Display System, December 17, 1984
- 2. Letter from J. A. Kay, YAEC to J. A. Zwolinski, NRC, Safety Parameter Display Systen, April 8,1985.
- 3. Letter from J. A. Kay, YAEC to J. A. Zwolinski, NRC, Safety Parameter Display System, September 3,1985.
- 4. Letter from NRC to YAEC, Safety Evaluation Report - Safety Parameter Display System, March 17, 1986.
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