ML20059L680
| ML20059L680 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 11/09/1993 |
| From: | Paulitz F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059L650 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 9311170284 | |
| Download: ML20059L680 (10) | |
Text
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C, 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REGULATORY GUIDE 1.97 COMPLIANCE s
COMMONWEALTH EDIS0N COMPANY ZION STATION. UNITS 1 AND 2 DOCKETS NOS. 50-295 AND 50-304
1.0 INTRODUCTION
The operability of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to allow the operator to perform manual actions, and monitor plant status following an accident.
Regulatory Guide (RG) 1.97 specifies the types, design, and qualification criteria that must be met for these instrumcats.
By letter dated April 15, 1991, Commonwealth Edison Company (CECc, the licensee) responded to four deviations identified during an inspection (Reference 16) and provided an update to the Zion RG 1.97, Parameter Summary Table (Reference 17). The update to the table also addressed licensee commitments to comply with RG 1.97 requirements which had not been previously identified. The previous Safety Evaluation (SE) on this subject was sent to the licensee on April 14, 1989 (Reference 5).
2.0 BACKGROUND
As a result of the Three Mile Island accident on March 29, 1979, the Commission issued NUREG-0737, on October 31, 1980, which incorporated all TMI-related items approved for implementation at that time into one document.
On December 17, 1982, the Commission issued Generic Letter (GL) 82-33 to provide aMtional clarification regarding a number of issues including the Safety Parameter Display Systems, and RG 1.97 application to emergency response facilities (Reference 1). RG 1.97 was issued as Revision 2 in December 1980 (Reference 2). The licensee responded to the Commission's request in letters dated August 1, 1986 (Reference 3), and August 24, 1987 (Reference 4).
The staff reviewed this submittal regarding the licensee's commitments to comply with the guidelines of RG 1.97, and provided an SE on August 14, 1989 (Reference 5). This SE concluded that the instrumentation provided by Ceco, for meeting the recommendations of RG 1.97, Revision 2, was acceptable except for the following variables:
1.
accumulator tank level and pressure, 2.
neutron flux, and 3.
containment isolation valve position.
9311170284 931109 DR ADDCK 0500 5
l Between May 23, 1990, and January 31, 1991, the licensee submitted and clarified a proposal to change the Technical Specifications (TS) to incorporate the post-accident monitoring instrumentation into the TS, and provided additional information regarding compliance to RG 1.97 (Reference 6 through Reference 15).
The staff conducted an inspection of the Zion post-accident monitoring instrumentation on March 1, 1991 (Reference 16), and the licensee submitted 7
additional information on April 15,1991 (Reference 17). On November 22, 1991, the licensee again submitted a proposed amendment to the Zion Facility Operating License to change the TS regarding, " Accident-Monitoring Instrumentation" (Reference 18). The Zion Facility Operating License was j
amended on January 27, 1993, to include the " Accident-Monitoring Instrumentation" into the TS (Reference 19).
l 3.0 EVALUATION l
By letter of April 15, 1991, CECO documented the completion of previous commitments to re-evaluate the current reactor coolant system pressure recorder design and provided an update of the Parameter Summary Table for staff review. The update to the table also addressed a concern with the instrument ranges identified in the above inspection report (Reference 17).
The staff evaluated the information submitted by the licensee on April 15, 1991, based on the requirements of RG 1.97, Revision 2, and its referenced Standard ANSI /ANS-4.5-1980 (Reference 2a) which referenced Standard ANSI /IEEE N41.26, Std. 497-1977 (Reference 2b). The staff also evaluated the post-accident instruments to those instruments identified in the licensee's Emergency Operating Procedures (References 20, 21 and 22).
3.1 Inspection Report 50-295/90032 (DRS) and 50-304/90032 (DRS)
The inspection report (Reference 16) identified the following deviations from commitments that the licensee previously made to the Commission:
3.1.1 _ Reactor Coolant System Pressure Recorder Failure of Recorder PR-403 would cause the operator to lose a recorded trend of Reactor Coolant System (RCS) wide range pressure from redundant instrument channels PT-403 and PT-405. The pressure recorder, therefore, does not meet the single failure requirements of RG 1.97.
The licensee stated in Notes 9 and 10 of the Notes and References section of the RG 1.97 Parameter Summary Table in Reference 17, that the two redundant pressure channels, PT-403 and PT-405, are isolated from the recorder and each channel is indicated on the control board; therefore, this pressure information is available to the operator if the recorder failed.
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l RG 1.97 only requires the recording of a single channel of a parameter unless direct and immediate trend or transient information is essential for operator j
information or action, in which case the recording should be continuously available on redundant dedicated recorders.
The staff reviewed Emergency Operating Procedures (EOPs) (References 20, 21, and 22), and found no reference that directs the operator to observe RCS pressure trends or transients. Direct and immediate trend or transient information is, therefore, not essential for operator information or action.
Therefore, the staff finds that the single recorder PR-403, recording RCS pressure for redundant instrument channels PT-403 and PT-405, is acceptable.
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3.1.2 Instrument Rangg Category 1, Type A, variables did not meet the RG '.97 instrument ranges.
These deviations were not previously identified by.he licensee as an i
exception to RG 1.97.
l VARIABLE RE0UIRED RANGE ACTUAL RANGE Steam Generator WR Level From tube sheet to bottom 575 inches of Chevron boxes (588 inches)-
Steam Generator NR Level Top of tubes to bottom 150 inches of Chevron boxes (171 inches)
Pressurizer Level Bottom to top-526 inches (634 inches)
Steam Line Pressure 0-1260 psig 0-1200 psig i
a.
Steam Generator Level Note 11 of the RG 1.97 Parameter Summary Table in Reference 17 for steam generator water level instruments states:
Wide range steam generator level indication of "0-100%" corresponds to a measurement of 0-575 inches of water column. The lower tap is located approximately 18 inches above the tube sheet and the upper tap is located slightly above the lower portion of the lower cyclone separator. This range is acceptable because it is the range used in the Final Safety Analysis Report (FSAR) accident analysis and is the range used in the existing E0Ps.
The range of the steam generator level channels is physically constrained by the location of the level taps on the vessels (standard Westinghouse design).
The staff's review of E0P E-1, " Loss of Reactor or Secondary Coolant"; found no reference to the SG 1evel instrument range. The difference between the required range of the steam generator wide range (WR) water level and the i
actual range is 13 inches or about 2% of the required range.
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. RG 1.97, Table 3, states the purpose of this variable is to monitor operation and the RG makes no reference to SG narrow range requirements. The staff finds the difference of 2% between the required range and the actual range acceptable because it will not preclude monitoring operation and, therefore, meets the intent of the RG.
b.
Pressurizer Water Level The licensee's associated Notes 14 and 15 for pressurizer water level state that:
"the vertical distance between the pressurizer level reference leg tap and the upper tap is approximately 520." The lower tap is located approximately 60 inches above the bottom of the surge nozzle and the upper tap is located approximately 53 inches below the top of the spray nozzle. The volume associeted with the distance between the surge nozzle and the bottom tap and from the top tap to the spray nozzle is not used in the pressurizer level measurement nor is it required for proper operation of the pressurizer.
The indicated range is sufficient for normal and emergency operation because this indicated range was used in the FSAR accident analysis and is the range used in the existing E0Ps. The range of the pressurizer level channels are physically constrained by the location of the level taps on the vessels (standard Westinghouse design).
The staff reviewed E0P E-1 and found only one reference to pressurizer level which required operator action. This was to direct the operator to restart the Emergency Core Cooling System (ECCS) pump if the pressurizer level, "cannot be maintained greater than 4% (30% for adverse containment)."
RG 1.97, Table 3, lists the pressurizer level as Type D, Category 1 Variable, with a range from bottom to top to ensure proper operation of the pressurizer.
The volume associated with the distance between the surge nozzle and the bottom tap and from the top tap to the spray nozzle is not used in the pressurizer level measurement nor is it required for proper operation of the pressurizer; therefore, the staff finds the difference between the RG 1.97 instrument required range and the actual instrument range acceptable.
c.
Steam Generator Pressure The RG 1.97 required range for steam generator pressure instruments is from 0 to 120% of the lowest safety valve st.tting (1050 psig). Therefore, the i
required range is 0 to 1260 psig.
1 The actual instrument range is 0-1200 psig with a reference to Note 16. Note 16 states:
Technical Specifications Limiting Condition for Operation (TS LCO) 3.7.1 requires twenty ASME code safety valves (5 per steam generator) be operable for full power operation. The set pressure for these valves range from 1050 psig to 1100 psig. The twenty code safety valves have a combined capacity to relieve the total design steam flow of one unit and the existing steam generator pressure range of 0-1200 psig is sufficient to monitor all expected steam generator pressure conditions.
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The staff finds the maximum pressure of 1200 psig for the steam generator steam pressure instrument is greater than the highest setting of the safety relief valve pressure of 1150 psig and the TS requires twenty ASME code safety valves (5 per steam generator) be operable for full power operation.
Therefore, the staff finds the difference between the RG 1.97 required instrument range and the actual instrument range acceptable.
3.2 Additional Licensee Identified Deviations From RG 1.97 The licensee's submittal dated April 15, 1993, included the Zion RG 1.97, Parameter Summary Tables 1 and 2, and Notes and References (Reference 17).
The staff's review of these tables and notes and references identified the following deviations from RG 1.97 requirements:
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3.2.1 Control Board Indicators and Recorders Notes 1, 2, and 3 of Notes and References (Reference 17), state that control board indicators and recorders are not qualified to RG 1.89, " Qualification of Class IE Equipment for Nuclear Power Plants," and RG 1.100, " Seismic Qualification of Electric Equipment for Nuclear Power Plants," and RG 1.75,
" Physical Independence of Electrical Systems for Nuclear Power Plants,"
separation requirements.
RG 1.97 requires the following:
Accident monitoring information display channels shall be environmentally qualified in accordance with RG 1.89.
Equipment qualification applies to the complete instrumentation channel from sensor to display where the display is a direct-indicating meter or recording device.
All Type A and those Type B information display channels used during Phase I shall be qualified in accordance with RG 1.100 to continue to function within the required accuracy following, but not necessarily during, a safe shutdown earthquake. Type B information display channels used during Phase II and all Type C information display channels are not required to be seismically.
Redundant or diverse channels should be electrically independer.t and physically separated from each other and from equipment not classified important to safety in accordance with RG 1.75.
The licensee states that:
indicators are electrically isolated'from the safety-related portion of the channels and are observed daily by the operators.
Further, the TS require a monthly channel check and a channel calibration each refueling outage; the seismic qualification of Zion's control board instrumentation which is required for safe shutdown will be evaluated through the Seismic Qualification Utility Group (SQUG) program.
Instrumentation required for safe shutdown will be upgraded as deemed necessary based upon the recommendations of the SQUG evaluation; Zion, Unit 1, i
began commercial operation in December 1973 which predates the issuance of the Regulatory Guides.
Further, indicators and recorders are located in the r-e
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control room which are protected from high energy pipes, rotating equipment missiles and adverse accident environments.
The separation requirements of RG 1.75 were establish to preclude failures of redundant channels associated with maintenance or surveillance activities within the control boards. The probability of this type of failure is reduced by adequate training and procedures for these activities.
The staff agrees with the deviation from the requirements of RG 1.97, for control board indicators and recorders in reference to the requirements of RG 1.89, RG 1.100, RG 1.75, and RG 1.97 discussed above.
3.2.2 RCS Hot and Cold Lea Water Temperature All RCS hot leg water temperature (Wide Range) instrumentation, Variables I and 19, have a common power supply and the resistance temperature detectors (RTDs) pass though a common containment electrical penetration. All RCS cold leg water temperature (Wide Range) instrumentation, Variables 2 and 20, have a common power supply and the RTDs pass through a common containment electrical penetration. The variable is Category I that requires redundancy subject to the single failure criteria. The RCS hot and cold leg temperature instrumentation power supplies and electric containment penetrations are separate from each other; however, both the RCS hot leg and cold leg water temperature measurements are subject to failure of their respective power supply or containment electrical penetration which does not meet the single failure criterion for Category 1 channels.
The licensee states that: the second element, of the duel element RTDs, provides two wide range hot and cold leg channels with indication on the remote shut down panel. These channels are electrically independent of the channels displayed in the control room.
Variable No. 13, core exit Thermocouple Temperature displayed on the main control board, may be used as diverse indication to the RCS hot leg RTDs and Variable No. 7, Steam Generator Pressure, may be used to determine the approximate value of RCS cold leg temperature (i.e., saturation temperature for steam generator pressure).
Although the RCS hot and cold leg temperature variable does not meet the single failure criteria, the staff finds the design acceptable because independent measurements can be monitored at the remote shutdown panel and diverse channels are available.
3.2.3 Containment Water level The Containment Water Level (Wide Range) Variable 27, Type B, and Variable 37, Type C, environmental qualification (EQ) test report for the transmitters i
states that the transmitters exhibit a high level of inaccuracy due to thermal non-repeatability when exposed to Design Basis Accident (DBA) temperatures.
Containment Recirculation Sump Level (CRSL) has replaced Containment Water level (Wide Range) as a Type A variable because CRSL is the indication used in the station procedures to verify that adequate net positive suction head is
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available to the residual heat removal (RHR) pumps during cold leg recirculation. However, CRSL does not meet the EQ requirements of a Type A variable, and the required modifications have turned out to be cost prohibitive. Therefore, the licensee will be submitting a TS amendment request to use the refueling water storage tank (RWST) water level, which
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already meets the requirements of a Type A variable, as the primary indicator 4
of the appropriate time at which to switch over to the recirculation mode of core cooling. As confirmatory indication, the licensee intends to use r
containment water level, which is a Type B variable. The containment water level transmitters will be environmentally qualified in refueling outages in 1995. Use of the RWST level is an acceptable method of determining the point at which to switch over to the recirculation mode at similar facilities.
For the purposes of Regulatory Guide 1.97 requirements, this issue is closed, and it will be followed through the licensee's TS amendment.
3.2.4 Neutron Flux Modifications will be completed during refueling outages in 1995 that replace the source and the intermediate / power range neutron flux monitoring instrumentation with types that meet Regulatory Guide 1.97, Category 1, requirements.
3.2.5 Refuelina Water Storaae Tank level The RWST level instrument loops have been modified to meet the Regulatory Guide 1.97, Category 1, requirements.
3.2.6 Condensate Storaae Tark level Two new Condensate Storage Tank (CST) level indication loops have been installed that meet Regulatory Guide 1.97, Category 1, instrumentation requirements.
3.2.7 Pressurizer Safety Valve Position The Pressurizer Safety Valve Positions (Primary Indications), Variable 55c, have been modified to upgrade the discharge temperature indicators to Category 2 requirements.
4.0 CONCLUSION
Based on our review, we find that the licensee either conforms to RG 1.97 or the staff finds the following deviations acceptable:
a.
A single recorder PR-403 recording RCS pressure for redundant instrument channels PT-403 and PT-405.
b.
The difference between the required range and the actual range for the steam generator wide range water level, Variable 4 and 61, instrument.
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- I The difference between the required range and the actual range for i
c.
Pressurizer Level, Variable 6 and 56.
d.
The difference between the required range and the actual range for the steam generator pressure, Variable 7 and 62.
e.
The staff agrees with the deviation from RG 1.89, RG 1.100, RG 1.75, and RG 1.97 for indicators and recorders in the control room.
f.
The RCS hot and cold leg temperature variables do not meet the single i
failure criteria; however, the staff finds the design acceptable because independent measurements can be monitored at the remote shutdown panel and diverse channels are available.
The staff finds the following resolutions of deviations from RG 1.97 acceptable:
l i
a.
The licensee will submit a TS amendment to allow use of RWST level as the primary indicator that core cooling should be switched to the recirculation mode.
b.
The Neutron Flux source and intermediate range neutron flux monitoring instrumentation is not environmentally qualified. The licensee is proceeding with a modification to meet Category I requirements that will be completed during refueling outages in 1995.
c.
The RWST level instrument loops have been modified to be in compliance with RG 1.97 Category I requirements.
d.
The CST Water Level indicator loops have been modified to be in compliance with RG 1.97 Category I requirements.
e.
The pressurizer safety valve positions (primary indication) have been modified to be in compliance with RG 1.97 Category 2 requirements.
Principal Contributor:
Fred Paulitz Date:
November 9, 1993 i
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5.0 REFERENCE 1.
Generic Letter, D.G. Eisenhut (NRC) to All Licensees of Operating Reactors, Applicants for Operating Licenses, and Holders of Construction Permits, " Supplement No. I to NUREG-0737--Requirements for Emergency l
Response Capability (GL No. 82-33)," December 17, 1982.
2.
Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Condition i
During and Following an Accident," December 1980.
2a.
American National Standard /American Nuclear Society, ANSI \\ANS-4.4-1980,
" Criteria for Accident Monitoring Functions in Light-Water-Cooled Reactors."
2b.
Draft American National Standard \\The Institute of Electrical and Electronics Engineers N41.26 Std 497-1977, "IEEE Trial Standard Criteria for Post Accident Monitoring Instrumentation for Nuclear Power Generating Stations."
3.
Letter from P.C. Leblond (Ceco), to H.R. Denton, (NRC) August 1,1986.
4.
Letter from P.C. Leblond (Ceco), to USNRC, August 24, 1987.
t 5.
Letter from C.P. Patel (NRC), to H.E. Bliss (Ceco), "Conformance to Regulatory Guide 1.97, Revision 2, Zion Station, Units 1 and 2, Safety Evaluation Report" (TAC Nos. M51467 and M51368), April 14,1989.
6.
Letter from R.A. Chrzanowski (Ceco), to T.E. Murley (NRC), " Zion Station, Units 1 and 2, Application for Amendment to Facility Operating Licenses DPR-39 and DPR-48, NRC Docket Nos. 50-295 and 50-304," May 23, 1990.
7.
Letter from R.F. Dudley (NRC), to T.J. Kovach (CECO), " Additional Information Required for Commonwealth Edison Company Licensing Reviews,"
June 15, 1990.
8.
Teleconference between Commonwealth Edison Company and C. Patel of the NRC, July 26, 1990.
9.
Letter from R.A. Chrzanowski (Ceco), to T.E. Murley (NRC), " Zion Station, Units I and 2, Regulatory Guide 1.97, NRC Docket Nos. 50-295 and 50-304," July 30,1990.
i 10.
Letter from R.A. Chrzanowski (Ceco), to NRC, " Zion Station, Units I and 2, Supplemental Information to Support an Application for Amendment to Facility Operating Licenses DPR-39 and DPR-48, NRC Docket Nos. 50-295 and 50-304," August 26, 1990.
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11.
Letter from S.F. Stimac (Ceco), to NRC, " Zion Power Station, Units I and 2, Facility License Nos. DPR-39 and DPR-48, RG 1.97 Containment Isolation Valve Position Indication Supplemental Response, NRC Docket Nos. 50-295 and 50-304," October 31, 1990.
t 12.
Teleconference between G. Hausman (NRC), and S. Stimac (Ceco),
December 7, 1990.
i 13.
Letter from S.F. i'imac (CECO), to A.B. Davis (NRC), "2lan Station, Units 1 and 2, Ri i 97/EQ, NRC Inspection Additional Information i
Request, NRC Doc.at !ss. 50-295 and 50-304," December 14, 1990.
14.
Letter from S.F. Stimac (CECO), to T.E. Murley (NRC), " Zion Station, Units 1 and 2, RG 1.97, Status of Open Commitments, NRC Docket Nos.
50-295 and 50-304," December 27, 1990.
15.
Letter from S.F. Stimac (Ceco), to T. Murley (NRC), " Zion Nuclear Power Station, Units 1 and 2, Facility Licenses DPR-39 and DPR-48, Regulatory Guide 1.97, Containment Isolation Valve Position Indication Supplemental Response, NRC Docket Nos. 50-295 and 50-304," January 31, 1991.
t 16.
Letter from M.A. Ring (NRC), to C. Reed (CECO), " Inspection Report 50-295/90032 (DRS) and 50-304/90032 (DRS)," March 1, 1991.
17.
Letter from S.F. Stimac (Ceco), to T.E. Murley (NRC), " Zion Station, Units 1 and 2, Regulatory Guide 1.97, Compliance, NRC Docket Nos. 50-295 t
and 50-304," April 15, 1991.
t 18.
Letter from S.F. Stimac (Ceco), to T.E. Murley (NRC), " Zion Station, Units 1 and 2, Application for Amendment to Facility Operating Licenses DPR-39 and DPR-48 Appendix A, Technical Specifications, NRC Docket Nos.
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50-295 and 50-304," November 22, 1991.
t 19.
Letter from J. E. Dyer (NRC), to Ceco, " Zion Unit 1, Amendment No. 141 to License No. DRP-39 and Zion, Unit 2, Amendment No. 130 to License No.
DPR-48," January 27, 1993.
20.
Emergency Operating Procedure E-0, " Reactor Trip or Safety Injection" February 5, 1993.
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21.
Emergency Operating Procedure E-1, " Loss of Reactor or Secondary Coolant" March 12, 1993.
22.
Emergency Operating Procedure E-3, " Steam Generator Tube Rupture" February 6, 1993.
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