ML20248J825
| ML20248J825 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/03/1989 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20248J822 | List: |
| References | |
| NUDOCS 8904170092 | |
| Download: ML20248J825 (4) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY'THE OFFICE OF NUCLEAR REACTOR REGULATION j
i RELATED TO AMENDMENTS NOS. 87 AND 80 TO FACILITY OPERATING LICENSES N05. DPR-42 AND DPR-60 NORTHERN STATES POWER COMPANY
' PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS NOS. 1 AND 2 DOCKETS NOS. 50-282 AND 50-306 l
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INTRODUCTION I
I By letter dated July 18, 1988, as supplemented by letters dated September 15, i
1988 and March 10, 1989 Northern States Power Company (the licensee) requested l
amendments to the Technical Specifications (TSs) appended to Facility Operating j
Licenses Nos. DPR-42 and DPR-60 for the Prairie Island Nuclear Generating
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Plant, Units.Nos. I and 2.
The proposed amendments would change the Technical 1
Specifications by eliminating the reactor trip device associated with steam /
feedwater mismatch flow and low feedwater flow. Specifically, the proposed changes would impact the technical specification in the following areas I
1.
Specification 2.3.A.3(c) dealing with the reactor trip setpoints of
" low steam generator water level - 1 15% of the narrow range instrumegt in coincidence with stealii/feedwater mismatch flow -
g1.0x10 lbs/hr" would be deleted.
2.
Specification Table TS.3.5-2, item 18 dealing with low feedwater flow reactor trip, would be deleted.
3.
Specification Table TS 4.1-1, item 12, Steam Generator Flow Mismatch, would be modified so that surveillance would be performed on steam flow channels only since feedwater flow channels would no longer be used in J
the protection circuit.
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The associated TS bases would also be changed to reflect the removal of the low feedwater flow reactor trip described above.
The proposed changes would become effective after installing the digital feedwater control system incorporating the median signal selector function i
for each unit.
1 In support of the amendments requested, the licensee submitted by letter dated September 15, 1988, a report prepared by Westinghouse Electric Corporation (WCAP-11931 and non-proprietary version WCAP-11932) that describes the advanced l
digital feedwater control system containing the median signal selector.
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2.0 EVALUATION The Prairie Island Nuclear Generating Plant Unit Nos. I and 2 has two Westinghouse designed reactors that trip if a low-low water level is reacned in any one of the steam generators.
The reactor trip derived from the low-low water level in the steam generator protects the reactor from loss of the heat sink in the event of sustained steam /feedwater mismatch flow or low feedwater flow resulting from the loss of normal feedwater caused by a system pipe break inside or outside of containment.
In the event of loss of feedwater for any reason, the reactor would trip when the water level in the steam generator falls to the low-low level setpoint in the reactor trip circuitry.
Therefore, the low-low steam generator water level reactor trip circuit is provided for each steam generator which bounds the reactor trips initiated by the steam /
feedwater mismatch flow and the low feedwater flow.
The low-low steam generator water level reactor trip also ensures that sufficient initial heat removal capabilities (water inventory) is available in the steam' generator to protect the core at the start of the transient.
A review of the Prairie Island updated safety analysis report shows that no credit is taken for the reactor trip initiated by steam /feedwater mismatch flow or low feedwater flow in mitigating the consequences of any of the analyzed accidents.
In cases such as loss of main feedwater, steam or feedwater pipe break inside or outside of containment, or loss of offsite power credit is taken for the low-low steam generator water level reactor trip to ensure safe shutdown.
The steam /foedwater mismatch flow and the low feedwater flow reactor trip were installed to satisfy the requirements of the Institute of Electric and Electronics Engineers Standard 279, 1971 (IEEE-279) paragraph 4.7.3 which is endorsed by the Code of Federal Regulations, 10 CFR Part 50.55a.
Paragraph 4.7.3 of IEEE-279 states in part that a single random failure in a control system shall not prevent or block the proper action of a protection system from occurring.
Specifically, the criteria of Paragraph 4.7.3 are not met when a single failure in the main feedwater control system prevents the low-low steam generator water level channels from tripping the reactor.
Therefore, the low feedwater flow and the steam /feedwater mismatch flow reactor trip was installed in order to achieve an adequate substitute for meeting the separation criteria of Paragraph 4.7.3 of IEEE-279.
The staff has reviewed the report supporting the amendments requested prepared by Westinghouse Electric Corporation titled " Advanced Digital Feedwater Control System Median Signal Selector for Northern States Power Prairie Island Units 1 and 2" (WCAP-11931) describing the details of the enhanced feedwater control system.
The feedwater control system is enhanced by the installation of the median signal selector which effectively eliminates the concern regarding a single random failure causing a control system action that results in a I
condition requiring protective action and preventing proper operation of a protection system channel designed to protect against this condition. Therefore, the mechanism for providing acceptable control and protection system interactions
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is achieved between the steam generator low-low water level protective function
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and the feedwater control system in accordance with the requirements of IEEE Standard 279(1971).
The isolation devices providing protection for the steam generator low-low water level reactor protective function are Foxboro Model 66 i
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. BC-0 style C current repeater that were subjected to an extensive-testing program by Westinghouse Electric Corporation. The results of this testing program has been reported in the Westinghouse Electric Corporation Topical Report WCAP-7508-L (December 1970) and the staff found the test results acceptable as discussed in our letter dated June 6, 1973. By letter dated March 10, 1989 the licensee provided supplemental infomation confiming that
- the isolation devices are covered within the scope of the testing program for the maximum credible faults. As reported in the Westinghouse Topical Report WCAP-7685-A, (May 1975), the isolation amplifiers provide an effective electrical barrier.for the input (protection. side) signal when the output (control side)'
signal was subjected to faulted conditions.
In addition, the licensee verified that the maximum and minimum signal. levels allowed to pass through the isolation devices will in no way degrade the median signal selector such that damage would occur to the main feedwater flow control system. The.verificationand validation program perfomed at Westinghouse does include an actual testing program supporting.the claim'of adequate operability of the median signal selector upon the. receipt of the maximum and minimum signal. levels that normally passes through the isolators. While. the staff has not yet perfomed an audit of the verification and validation program, our review is complete to a point where we find the application of the advanced digital feedwater control system containing the median signal selector acceptable. This acceptability is contingent upon our finding an adequate verification and validation program during the forthcoming audit at the Westinghouse offices.
On this basis, the reactor trip initiated by steam /feedwater mismatch and low feedwater flow are no longer necessary or required.
In conclusion, the safety analysis shows that no credit is taken for the reactor trip initiated by the steam /feedwater mismatch flow and low feedwater flow in mitigating the consequences of any of the analyzed design bases accidents.
The initial installation of this trip was for the purpose of satisfying the single random failure requirement specified in IEEE 279 (1971) paragraph 4.7.3 for control and protective system interactions. The advanced feedwater control system provides an acceptable method of resolving the interaction concern between the feedwater control and steam generator low-low water level protective function. The control and protection system interaction meets the requirements specified in paragraph 4.7.3 of IEEE 279. The acceptability of the protective function meeting the separation requirement of IEEE 279 (1971) is predicted upon an acceptable audit of the manufacturer's verification and validation program associated with the software for the median signal selector. On this basis, the staff finds the proposed changes involving the i
elimination of the steam /feedwater mismatch flow and the low feedwater flow reactor trip acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on April 3, 1989 (54 FR 13445).
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. :., Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendment will not have a significant effect on the quality of the human environment.
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4.0 CONCLUSION
We have 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, and (2) such l
activities will be conducted in compliance with the Commission's regulations, and 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 Contributor: Dominic C. Dilanni Date: April 3,1989 1
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