ML20202B307
| ML20202B307 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 01/22/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20202B293 | List: |
| References | |
| NUDOCS 9901290049 | |
| Download: ML20202B307 (6) | |
Text
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3 UNITED STATES NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20066 4 001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 104 TO FACILITY OPERATING LICENSE NPF-68 AND AMENDMENT NO. 82 TO FACILITY O'PERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY. INC.. ET AL.
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 DOCKET NOS. 50-424 AND 50-425
1.0 INTRODUCTION
By letter dated September 3,1998, as supplemented by letter dated December 8,1998, Southem Nuclear Operating Company, Inc., et al. (the licensee), proposed license amendments to change the Technical Specifications (TSs) for Vogtle Electric Generating Plant (Vogtle),
Units 1 and 2. The proposed amendments would change the TSs for Vogtle Units 1 and 2 to:
(1) support the replacement of (a) the Nuclear instrumentation System Source Range and Intermediate Range Channels and (b) Post Accident Neutron Flux Monitoring System (NFMS) and (2) delete the requirement for performing response time testing of the source range channels and power range detector plateau voltage determinations. The December 8,1998, supplement provided clarifying information that did not change the scope of the September 3, 1998, application and the initial proposed no significant hazards consideration determination.
2.0 DISCUSSION The overpower protection provided by the out-of-core nuclear instrumentation consists of three discrete but overlapping ranges (source range (SR), intermediate range (IR), and power range (PR)) for monitoring reactor flux. Continuation of reactor startup operation or power increase requires a permissive signal from the higher range instrumentation channels before the lower range level trips can be manually blocked by the operator. The PR low setpoint trip and the IR and SR trips are designed to protect the reactor core against power excursions during reactor startup or low-power operation. The SR and IR trips provide redundant protection during reactor startup or low-power operation.
in order to improve system reliability, the existing SR and IR excore detector system supplied by Westinghouse is being replaced with an equivalent neutron-monitoring system manufactured by Gamma-Metrics. The Westinghouse-supplied post-accident NFMS, which currently provides indication of reactor core flux in post-accident conditions in accordance with the guidance in Regulatory Guide (RG) 1.97, will also be replaced by the new Gamma-Metrics system. In addition, the Gamma-Metrics system will replace the current High Flux at Shutdown Alarm (HFASA) function used to alert the operators in case of an inadvertent boron dilution event.
The licensee has proposed changes to the present TS requirements because of design differences between the old and new reactor flux monitoring systems. In addition to these 9901290049 990122 PDR ADOCK 05000424 P
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2-l changes, the licensee has proposed to delete the requirements for performing response time testing of the source range channels and power range detector plateau voltage determinations.
In a December 8,1998, supplemental letter, the licensee provided additional clarifying information on the proposed TS changes and also provided two additional TS Bases changes that were identified after the original September 3,1998, license amendment request was made.
l 3.0 EVALUATION The proposed design change replaces the existing Class 1E Westinghouse excore SR, IR, and NFMS with an equivalent neutron-monitoring system using the Gamma Metrics Series 300 j
design. This total replacement includes the detectors and the associated processing electronics. The new Gamma-Metrics detector system consists of fission chambers that will pedorm the SR, IR, and post-accident monitoring functions. The licensee stated that the new system meets the safety-related Class 1E design requirements in the current Vogtle licensing j
basis for the existing system and meets RG 1.97, Revision 2, and Branch Technical Position CMEB 9.5-1 guidance for the design of post accident monitoring instrumentation. The new Gamma-Metrics equipment is compatible with the rest of the nuclear instrumentation and reactor protection systems as well as with the Plant Safety Monitoring System (PSMS) for post-accident monitoring, and will perform all the functional requirements of the equipment being replaced; however, the new system design differs in six major aspects from the present Nuclear instrumentation System (NIS) design, which necessitated changes to the TS and the TS Bases. These differences are (1) change in source range detector output, (2) change in intermediate range scale units from amps to percent power, (3) change in source range scale from six to seven decades, (4) change in HFASA setpoint, (5) no need to deenergize the Gamma-Metrics SR detector high voltage, and (6) no r eed to determine detector plateau curve for calibration of Gamma-Metrics equipment.
The following TS changes are proposed:
(1) Delete Note 2 in Surveillance Requirement 3.3.1.11.
(2) For TS Table 3.3.1-1, " Reactor Trip System Instrumentation":
. Increase allowable value of IR neutron flux trip (Function 4).
. Increase allowable value of SR neutron flux trip and delete response time Surveillance Requirement 3.3.1.15 (Function 5).
. Convert P-6 allowable value and trip setpoint from " amp" to "% RTP [ reactor thermal power)" (Function 16.a).
1 (3) Change reference to IR scale (Function 4) in Bases B 3.3.1.
l l
l
--=
l s
7 s
' (4) For Bases B 3.3.1," Applicable Safety Analyses":
Remove reference to deenergizing SR detectors above P-6 (Function 5).
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. Remove reference to deenergizing/ energizing SR detectors (Function 16.a).
Remove reference to deenergizing SR detectors (Function 16.e).
j Remove reference to detector plateaus and correct a typographical error in SR 3.3.1.11. The typographical error referencing 50 percent RTP is being corrected 1
to 75 percent RTP, making it consistent with SR 3.3.1.6.
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Revise Reactor Trip System (RTS) Instrumentation Bases page B 3.3-60 to include this amendment in the list of references (Reference 6 on page B 3.3-60).
Revise HFASA Instrumentation Bases page B 3.3-172 to remove reference to the SR instrumentation being deenergized.
(5) For Bases B 3.3.8, the HFASA setpoint will be changed from "2.3" to "s 2.3."
(6) For Bases B 3.9.3:
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. Replace reference to BF detectors with fission chambers and revise instrument 3
range and accuracy to be consistent with new instrumentation.
Remove reference to SR detector plateau from SR 3.9.3.2.
The tables that follow show the before and after values for neutron flux trip setpoints and allowable values as they relate to the design changes as they are reflected in proposed changes in IR neutron flux, SR neutron flux, and IR neutron flux, P-6.
BEFORE CHANGE Table 3.3.1-1 Function Allowable Value Nominal Trip Setpoint 4.
Intermediate Range s 31.1% RTP 25% RTP*
Neutron Flux 5.
Source Range s 1.4 E5 cps 1.0 E5 cps Neutron Flux 16.a. Intermediate Range 2 6E-11 amp 1E-10 amp Neutron Flux, P-6
- (Current Equivalent) i l
f
,.6 l 1 AFTER CHANGE l
Table 3.3.1-1 Function.
Allowable Value Nominal Trip Setpoint 4.
Intermediate Range s 41.9% RTP" 25% RTP Neutron Flux (Changed)
(No Change) j 5.
Source Range s 1.7 E5 cps" 1.0 E5 cps l
Neutron Flux (Changed)
(No Change) 1 16.a.' Intermediate Range 21.2 E-5% RTP" 2.0 E-5% RTP Neutron Flux, P-6 (Changed)
(Equivalent Conversion)
" Derived value based on setpoint calculation.
As shown in the preceding tables, the P-6 (Function 16.a) value is the only trip setpoint that is changing; the new value is an equivalent conversion based on the approximate linear relationship between the IR current and reactor power. The licenses stated that the relationship between the IR current and reactor power has been verified through review of several cycles of l
plant data from both Vogtle units. The P-6 setpoint value was determined by linear extrapolation of the IR current versus power to obtain the power equivalent of the P-6 setpoint.
The new allowable values were derived on the basis of rack uncertainty values for the new Gamma Metrics instrumentation. The setpoint methodology used at Vogtle is based on Westinghouse report WCAP-11269, which is referenced in TS Bases B 3.3.1. The methodology of statistically combining the uncertainty terms has not changed; however, instead l
of applying the rack uncertainty values to a linear scale, as is presently done in the l
Westinghouse setpoint methodology, the uncertainty values are applied to a logarithmic scale.
l This is a more appropriate method since the inst umentation is operating in the logarithmic mode; therefore, the proposed change to the setpoints and allowable values will implement realistic values based on the design capabilities of the instrumentation.
Additionally, as referenced in the licensee's December 8,1998, letter, the methodology used for the design change calculations, which verify the correct correlation between the SR neutron flux l-trip and the P-6 permissive setpoints, is similar to that used by the Sequoyah Steam Electric i
l Station (SSES) licensee for Amendment No.136 for SSES Unit 1, dated April 27,1990, and Amendments 185 and 177 for SSES Units 1 and 2, respectively, dated July 26,1994. The SSES TS changes were associated with the replacement of the Westinghouse SR and IR l
detectors with Gamma Metric detectors similar to the proposed Vogtie design change. The SR, IR, and P-6 setpoints are also consistent with Westinghouse functional requirements for nuclear l-startup instrument protection. The functional requirements specify a range of settings for the SR, IR, and P-6 setpoints. A comparison of the proposed Vogtle SR, IR, and P-6 setpoints with those of SSES and the Watts Bar Nuclear Plant, which are also Westinghouse four-loop plants, indicates that the values are comparable.
The reliability of the RTS has not been decreased by the proposed TS changes because the i
SR allowable value continues to be well below the power range setpoint of 25 percent RTP.
For the IR, the increase in allowable value does not impact overall RTS reliability because the
4 aD o 1
5-uncontrolled rod withdrawal analysis in Section 15.4.1 of the Vogtle Updated Final Safety Analysis Report (UFSAR) indicates that the rise in power is so rapid that the effect of an error in the trip setpoint on the actual time at which the rods release is negligible. The reactor trip actuation at 35 percent RTP (as assumed in the accident analysis) by the PR or 41 percent RTP by the IR (allowable value) will result in essentially the same accident response and, therefore, will maintain the overall RTS reliability provided by the IR. In the accident analyses, no credit is taken for the automatic protective actuation of the SR or IR trips. The current I
surveillance requirements for the IR trip do not include a requirement for performing a response time test; therefore, it is appropriate to delete the response time test for the SR trip in TS Table 3.3.1-1 by deleting the reference to Surveillance Requirement 3.3.1.15.
The proposed design change will also result in changes to the HFASA TS Bases 3.3.8. The l
primary purpose of the HFASA alarm is to monitor inadvertent boron dilution accidents while the reactor is shut down. The HFASA design submitted by Gamma-Metrics does not provide an alarm setpoint of 2.3 times background as assumed in the safety analyses. Rather,it provides hardware-selectable alarm setpoints of 1.25,1.5,2.0,2.5,3.0, and 4.0 times background. The value of 2.0 times background will be used (that is conservative relative to the value assumed in the safety analyses). The change of the setpoint in the conservative directbn will warn the operators of an unplanned boron dilution event in sufficient time (in excess of 15 minutes preceding loss of shutdown margin) to allow manual action to terminate the event. The main control room and containment alarms are not being modified. The HFASA will continue to be derived from the SR neutron detectors. The TSs are being revised to specify a setpoint of less j
than or equal to 2.3 to be consistent with the safety analyses.
l The proposed design change will also remove the need to deenergize the Gamma-Metrics SR detector high voltage. In the existing Westinghouse system, the SR indication is disabled by deenergizing high voltage to the SR detectors when the SR trip is blocked upon receipt of the
(
P-6 permissive. This is done in order to prevent damage to the BF detectors due to operation 3
I at flux levels beyond their design limits. The need to remove high voltage from the Gamma-Metrics fission chamber detectors no longer exists because the system is not subject to damage at high flux levels. The detectors will remain energized through alllevels of power operation. TS Bases 3.3.1 was revised to remove the requirement to deenergize the high voltage to the SR detectors.
The proposed design change will also eliminate the need to determine the detector plateau l
curve for calibration of the Gamma Metrics equipment. The Gamma Metrics fission chambers do not require detector plateau curves to be obtained as part of the channel calibration. The fission chambers operate in the ionization chamber region of the detector ionization curve at all l
flux levels. The pulse output of the detectors is not dependent on the applied voltage over a l
wide range of voltage, as the fission chambers are operated at a fixed high voltage.
Accordingly, it is appropriate to revise TS 3.3.1.11, by deleting " NOTE 2," to remove the requirement to determine plateau curves for the SR, IR, and PR detectors.
The staff determined that the changes to the SR and IR instrumentation and setpoints, as well as the deletion of SR response time testing, do not affect the Vogtle safety analyses conclusions because the SR and IR trips are not explicitly credited in any design-basis accident.
Only the power range low setpoint trip of 25 percent RTP is assumed to actuate to mitigate the j
uncontrolled rod cluster control assembly withdrawal accident as described in Sections 7.2 and 15.4 of the Vogtle UFSAR. The staff also determined that the HFASA function during a
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i l i boron dilution event will continue to be provided by the new SR detector system. No changes have been made to the HFASA setpoint value assumed in the Vogtle safety analyses. The new j
detector system is qualified as safety-related Class 1E equipment in accordance with the i
guidance of RG 1.97 for post-accident monitoring instrumentation. Finally, the staff determined that the functional and operability requirements for the PR channels are not affected by deleting the requirement for determining detector voltage plateaus; thus, the staff finds that the l
proposed TS changes reflect the design and operational characteristics of the new Gamma-Metrics equipment and do not adversely affect the overall operation or ability of the equipment to perform its intended function. The new TS setpoint values are functionally equivalent to the existing values and are consistent with the plant safety analyses.
On the basis of the NRC staff's review associated with the September 3,1998, application, as supplemented by letter dated December 8,1998, the NRC staff concludes that the licensee's proposed replacement of the Westinghouse NIS, including the SR,IR and post-accident NFMS, with an equivalent system supplied by Gamma-Metrics is acceptable. In addition, the proposed associated TS changes are consistent with the guidance in RG 1.97 for post-accident monitoring equipment, the current licensing basis arid safety analyses, and approved setpoint methodology, and are, therefore, acceptable.
4.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Georgia State official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change surveillance requirements. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (63 FR 53957, October 7,1998).
Accordingly, the amendments meet 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 amendments.
6.0 CONCLUSION
The Commission 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: M. Gareri D. Jaffe Date:
January 22, 1999