ML20040B373
| ML20040B373 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 12/30/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20040B370 | List: |
| References | |
| NUDOCS 8201250480 | |
| Download: ML20040B373 (5) | |
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- f sec oqjo UNITED STATES g
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NUCLEAR REGULATORY COMMISSION
- *E WASHINGTON. D. C. 20555 o
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' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N05. 5 3 AND 4 7 TO FACILITY LICENSE NOS. DPR-42 AND 0,PR-60 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT N05. 1 AND 2 DOCKET NOS. 50-282 AND 50-306 Introduction By letter dated August 27, 1976 Northern States Power Company (NSP), the licensee, requested amendments to Facility License Nos. DPR-42 and DPR-60 for the PrairieIsland Nuclear Generating Plant Units Nos. I and 2 (PINGP).
The requested amendments proposed changes in the Technical Specifications (TSs) in the following areas.
1.
TS 2.3.A.2h Power Range Flux Rates i
2.
TS 3.3.A.lh(3) RHR Valve Position Indication 3.
TS 3.3.D.2 Cooling Water Header Operability 4.
TS 3.4.A.8 Steam Exclusion Dampers 5.
TSs 3.6, 4.4 and Table TS 4.1-1 Containment and Annulus temperatures 6.
TS 4.5.B.2 Containment Fan Motor Operability tests 7.
TS 5.5.D Environmental Monitoring 8.
Table TS 6.7-1 Special tests Items 1, 2, 3, 6, 7 and the first part of Item 8 have been addressed by
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Amendment Nos. 21 and 15 issued by our letter dated June 6, 1977.
This Safety Evaluation report addresses the remaining open areas that NSP requested.
They are item.4, "TS 3.4.A.8 Steam Exclusion Dampers", item 5, "TS 3.6, 4.4 and Table TS 4.1-1 Containment and Annulus Temperatures" and the remaining parts of Item 8 " Table 6.7-1 Special tests" that have not been covered by Amendments 21 and 15.
Discussion and Evaluation _
I.
Item 4 "TS 3.4.A.8 Steam Exclusion Dampers" (Presently existing under TS 3.4.A.3)
The licensee has proposed a change in the wording of the TS concerned with the steam exclusion dampers. Currently the TS raads as follows:
"Both isolation dampers in each ventilation duct that penetrates rooms containing equipment required for a high energy line rupture outside of containment shall be operable or at least one damper in each duct shall be closed."
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2-The licensee proposes to change this TS to read:
"Both isolation dampers in eacn ventilation duct that. penetrates' rooms containing equipment required for a high energy line rupture outside of containment shall be operable or one damper in each duct with an inoperable isolation damper shall be closed."
As currently phrased in the TS, the specification can be interpreted to require closing at least one damper in all ducts if one duct has an inoperable isolation damper.
This interpretation is certainly not the intent of the TS. We agree with the licensee that the proposed change does clarify the intent of the existing TS 1.e., if a duct contains an inoperable damper, then the operable damper in the affected duct should be closed and in no way has any adverse safety implications. On this basis we have concluded that the proposed change to the TS is acceptable.
II.
Item 5 "TSs 3.6, 4.4 and Table TS 4.1-1 Containment and Annulus Temperatures" The licensee's proposed changes under this item consist.of:the following:
(A) Specifications 3.6.B. 3.6.C and 3.6.D now require that during power operation limits exist on the internal pressure and average air temperature of the containment vessel and;the containment shell temperature.
The proposed change would require these limits to be met whenever containment integrity is required instead of only during power operation. We find the licensee's proposed change is more conservative in that the proposed
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change wuuld envelope other operational modes as well as power operation, since containment integrity is required for all operating modes except when the reactor is in cold shutdown, the reactor is in a refueling ' mode or the fuel inside containment has never been irradiated. Therefore, we concur
..itn the licensee that if limits of the internal pressure and the average air temperature in the containment vessel and the containment shell tempera-ture are exceeded for all modes of operations including power operation then conditions shall be corrected within eight hours or procedures shall be implemented immediately to establish reactor conditions for which containment is not needed.
l On this basis we conclude that the proposed change to the TS is acceptable.
(B) The licensee has proposed to delete the wording in TS 3.6.C and 3.6.0 which refer to the special test program conducted during the first fuel cycle of Unit No.1 and associated surveillance requirenent in Table TS 4.1-1.
During the first fuel cycle of Unit No. 1, a special test program was conducted to determine the average air temperatures from air temperature measurements at a number of locations in the shield I
building and the containment building. Measurements were made monthly to verify that the containment-annulus differential temperature and the shell temperature were within acceptable limits.
Evaluation of data collected during the first fuel cycle has shown that the measured temperatures in the containment, annulus and contain-ment shell are within the limits established in the TS.
The objective m.
. of the special test program had been accomplished, and this requirement no longer exists. Therefore, on this basis the licensee proposed to delete the 'special test program as stated in TS 3.6.C and 3.6,0, and to delete items 28 and 29 from Table TS-4.1-1 which deals with the inspection of the l
instrument used in the special program. We have reviewed the information and the data taken during the first fuel cycle provided by the i
licensee and find that the program objectives have been accomplished.
On this basis, we conclude that the change that deletes the special test program as stated in TS 3.6.C and 3.6.0 and the deletion of itens 28 and 29 from Table TS-4.1-1 are acceptable.
(C) The licensee has proposed additional surveillance requirements to the TS as a result of his review of data from the special test program conducted during the first fuel cycle of Unit No. 1.
Under TS 4.4 the licensee proposes to add the following:
Containment and Shield Building Air Temperature Prior to establishing reactor conditions requiring containment integrity, the average air temperature difference between the' containment and its associated Shield Building shall be verified to be within acceptable limits.
1 Containment Shell Temoerature Prior to establishing reactor conditions requiring containment integrity, the ter:erature of tha containment vessel wall shall be verified to be within acceptable limits.
Evaluation of data collected during the first fuel cycle of Unit No. I showed that the proposed limiting temperature differential of 440F between the average containment and annulus air temperatures and the existing limiting containmsnt shell temperature of 300F for plant operation can be approached only when the plant is in cold shutdown. Accordingly, the licensee proposed the above additional surveillances to monitor containment air and shell temperatures and annulus air temperature prior.to plant heatup from cold shutdown. The objective of these conservative surveillances are to provide assurance that the above cited parameters are within acceptable limits prior to establishing conditior.s requiring containment integrity.
We have reviewed the information and datilaken by the licensee during
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the first fuel cycle. We concur with the licensee that the proposed limiting temperat'ure differential of 44cF between the average containment and annulus air temperatures and the existing limiting contain-ment shell temperature of 30 F for plant operation can be approached only 0
when the plant is in cold shutdown.
Therefore, we conclude that the above proposed surveillance requirement added to the Technical Specifications is acceptable.
9 III.
Item 8 Table TS 6.7-1 Special Tests The licensee has proposed to delete Table 6.7-1 Special Reports.
Currently, this table requires that the licensee submit three special reports to the flRC of studies that were to be conducted during the first fuel cycle of Unit 1.
The special reports covering these studies consist of 1.
Containment System Special Analysis 2.
Fuel Surveillance 3.
Leakage Detection Analysis.
The reports covering these studies resulted from requirements established in the Safety Evaluation Report for Prairie Island dated September 28, 1972 at the time of issuance of the Prairie Island Unit No. 1 operating license.
The licensee has satisfied these requirements by letteis dated March 31, April 9, and May 26, 1976 transmitting the special reports titled: " Coolant Leakage Detection System Performance at the Prairie Island fluclear Generating Plant", " Prairie Island Containment System Special Analysis" and " Unit 1 -
End of Cycle 1 Spent Fuel Inspection".
The report titled " Coolant Leakage Detection System Performance at the Prairie Island Nuclear Generating Plant" shows that the leakage detection systems at PINGP No. I and 2 meet the General Design Criterion 30 of 10 CFR 50 Appendix A.
We agree with the licensee that instruments and techniques-used in the surveillance program provide an effective means of detecting abnormal reactor coolant leakage. Based on our review cf the licensee submittal we find that the leakage detection analysis performed by the licensee is acceptable. We thus conclude that Item 4 " Leakage Detection Analysis" in Table TS 6,7-1 can be deleted.
The report titled " Prairie Island Containment System Soe:ial Analysis" I
shows that temperatures measured in the containnent, in the ar.nulus and at the containment shell are within the limits established in the TS.
Based on our review of this report, we find that the objectives of the special program have been met. We thus conclude that Item 2, Containment System Special Analysis in Table TS 6.7-1 can be deleted.
The report titled " Unit 1 - End of Cycle 1 Spent Fuel Inspection" shows that surveillance of highest power density fuel assemblies during the first refueling cycle of Unit 1 showed no evidence of fuel damage or densifica-tion.
In addition, fuel surveillance personnel from the fuel vendor Westinghouse Corp. have reviewed the results of the inspection and concur that no safety significant items were observed. We have reviewed the licensee's report and find that his conclusions are acceptable. We thus conclude that the licensee has satisfied the requirements of Item 3 Fuel Surveillance in Table TS 6.7-1 and therefcre it can be deleted.
In conclusion, the above Safety Evaluation addresses all of the remaining areas existing in Table TS 6.7-1.
Thus, based on these evaluations Table TS 6.7-1 can be deleted.
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5-Environmental Consideration We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an i
environmental impact statement or negative declaration an'd environ-mental impact appraisal need not be prepared in connection with the issuance of these amendments.
Conclusion We have concluded, based on the considerations d'iscussed above, that:
(1) because~ the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2).
there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date: 0EC 3 01981 9
O O
l 7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION-DOCKET NOS. 50-282 AND 50-306 NORTHERN STATES POWER COMPANY NOTICE OF ISSUANCE OF AMENDMENTS TO FAC1LITY OPERATING LICENSES The U. S. Nuclear Regulatory Commission (the Commission) has issued Amendment No. 53 to Facility Operating License No. DPR-42, and Amendment No. 47 to Facility Operating License No. DPR-60 issued to Northern States Power Company (the licensee), which revised Tech-nical Specifications for operation of Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2 (the facilities) located in Goodhue County, Minnesota. The amendments are effective as of the date of issuance.
The amendments revise the Appendix A Technical Specifications concerned with the operation of the steam exclusion dampers, the applicability of the containment pressure and temperature limits and the deletion of special tests which have been completed.
The application for the amendments complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulatic,s in 10 CFR Chapter I, which are set forth in the license amendments. Prior public notice of the amendments was not required since the amendments do not involve a significant hazards consideration.
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