ML20151M445
| ML20151M445 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 07/27/1988 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151M442 | List: |
| References | |
| NUDOCS 8808050167 | |
| Download: ML20151M445 (4) | |
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NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR PEfD RESULATION SUPPORTING AMENDMENT NO 21 TO FACILITY OPERATIiv.
' JEj NO. NPF-37 AND NPF-66 BYRON STATION, UNITS I AND 2 20CKET'N05.50-454AND50-455 AND SUPPORTING AMENDMENT NO.10 TO FACILITY OPERATING LICENSES NO. NPF-72 AND NPF-77 BRAIDWOOD STATION UNITS 1 AND I DOCKET NOS. 50-456 AND 50-457
1.0 INTRODUCTION
By letter dated Jcnuary 18, 1988, Comonwealth Edison (Ceco), the licensee, submitted a proposed amendment to Facility Operating Licenses No. NPF-37, NPF-66, NPF-72, and NPF-77, for the Byron and braidwood Stations, Units 1 and 2.
The proposed amendment requests seven miscellaneous Technical Specification (TS) changes which are further discussed in Section 2.0.
A Notice of Consideration of Issuance of Amendaient to Facility Operating Licence and Proposed No Significant Hazards Consideration Detennination and Opportunity for Hearing related to the requested action was published in the Federal Register on April 6, 1988 (53 FR 11367). No requests fo'-
hearing and no public coceents were received.
2.0 DISCUSSION AND EVALUATION The following are descriptions and evaluations of each of the seven TS chaiges for Byron and Braidwood Units 1 and 2.
None of the changes involve physical modifications to the facilities.
It should be noted that some of the changes are specific to Syron Station.
Description o~ Change; Technical Specification Pages 3/4 4-27, 3/4 4-28, B3/4 4-5, 83/4 4-6, and 6-18 The proposed change deletes the requirement for a Special Report to the Cnnission if reactor coolant specific activity exceeds 1 microcurie per ga m dose equivalent I-131 for greater than 500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> in any consecutive six month period.
It also deletes the requirement to shutdown a plant if reactor coolant iodine activity limits arc exceeded for 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> in a 12-month period.
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. Evaluation This change is based on Generic Letter 85-19, dated September 27, 1985.
The letter stated that reporting requirements for iodine spiking can be reduced from a short term report to an item included in the Annual Report.
The letter further states that the requirements to shut down a plant after 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> with iodine activity above the limit can be eliminated due to the fact that nuclear fuel has been greatly improved in recent years, with the result that normal coolant iodine is well below the limit.
Appropriate actions would he initiated long before accumulating 800 hours0.00926 days <br />0.222 hours <br />0.00132 weeks <br />3.044e-4 months <br /> above the iodine activity limit.
The proposed TS changes are consistent with the nodel TS recomended in the Generic Letter, and are therefore considered acceptable.
Description of Change; Technical Specification Page 3/4 4-40 The proposed change revises Figure 3.4-4, "Nominal PORY Pressure Relief Setpoint Versus RCS Temperature for the Cold Overpressure Protection System Applicable up to 10 EFPY." The changes reflect a larger uncertainty in the wide range temperature instrumentation and prevent the need for additional stress analyses following :n overpressure event.
Evaluation The requested change to Cold Overpressure Mitigation System (COMS) setpoints is based on a letter from Westinghouse dated November 16, 1985 for two reasons:
(1) A larger uncertainty in the wide range temperature instrumentation is assumed; and (2) The updated COMS setpoints eliminate the need for a detailed stress evaluation of the PORY inlet and discharge piping and steam generator tube sheet following a single overpressure event.
The revised setpoints' meet 10 CFR Part 50, Appendix G criteria and are consistent with the Byron /Braidwood FSAR. The margin of safety has not been reduced because the change is in the conservative direction and is therefore bounded by previous analyses. The proposed change is considered acceptable.
Description of Change; Technical Specification Page 3/4 5-1 The proposed change revises TS Surveillance 4.5.1.1, which deals with accumulator operability. The proposed change deletes the words "by the absence of alarms" from the phrase:
"Verifying, by the absence of alarms, the contained borated water level and nitrogen cover-pressure in the tanks."
Evaluation The licensee requested the change because the current wording could be 1
interpreted that the unit must be shut dcwn if an annunciator failed.
Deleting the words "by the absence of alarms" permits the operators to verify the required accumulator borated water level and nitrogen cover pressure by using other instruments.
Duplicate level channels and
t pressure channels provide signals to two sets of safety-related instruments in the control room which can be used to read accumulator water 'evel and nitrogen cover pressure. The proposed TS change still requires the verification of accumulator parameters, but will allow the operators flexibility in how the parameters are verified.
The proposed change is consistent with Section 6.3 of the FSAR which describes the accumulators and the associated instrumentation.
The proposed TS change has no effect on safety and is considered acceptable.
Descripti.on of Change; Technical Specification Page 3/4 6-23 The proposed change corrects a typographical error for one Safety Injection Yalve number on Table 3.6-1, from "SI 8805D" to SI 89050,"
for Byron Station TS only, Evaluation The corrected valve number, "SI 8905D," is consistent with those listed in Braidwood TS Table 3.6-1 and Byron /Braidwood FSAR Table 6.2-58.
The change is administrative in nature, has no effect on safety, and is considered acceptable.
Description of Change; Technical Specification Page 3/4 7-14 The proposed change corrects a typographical error in the value of the ultimate heat sink (VHS) cooling tower basin water level from 873.5 feet to 873.75 feet, for Byron Station TS only.
Evalustion The corrected water level of 877.75 feet above mean sea level is consistent with other portions of Byron TS 3/4 7.5 which reference a minimum VHS cooling tower basin water 1, vel. The change is administrative in nature, has no effect on safety, and is considered acceptable.
Description of Change; Technical Specification o' ' 5-6 The proposed change revises TS Table 5.7-1, "Component Cyclic or Transient Limits," so that it is consistent with the design limits contained in Section 3.9 of Byron /Braidwood FSAR.
Evaluation The proposed change raises the limit for reactor coolant system (RCS) leak tests from 50 to 200, the limit for RCS hydrostatic pressure tests from 5 to 10, and the limit for secondary coolant system hydrostatic tests from 5 to 10.
It also raises the limits for primary and secondary pressures during hydrostatic testing to 1.25 times the design pressures as required by the ASME Code. The changes are consistent with Section 3.9 of the FSAR and Section XI of the ASME Code. Although these changes raise the number of transients the plants are pennitted to withstand, the changes are consistent with the FSAR.
Therefore, the proposed changes are acceptable.
s, t Description of Change; Technical Specification Pages 6-7, 6-8, and 6-13 The proposed changes are being made to update some Commonwealth Edison management titles and 'arify the functional authority of Quality Assurance personnel.
ihe change requested for Page 6-7 has previously been corrected in the Braidwood I'S.
The proposed changes are administrative in nature. Because there are no significant changes in duties, the changes have no adverse effect on safety, and are considered acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
This amendment involves primarily changes in the installation or use of facility components located within the restricted area defined in 10 CFR Part 20, and changes in reporting and surveillance requirements. The changes are primarily administrative in nature and do not involve any physical modifications to the facility. The amendment involves no significant increase in the amounts and no significant change in the tyoes of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Comission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public coment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(0) and (c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.
4.0 CONCLUSION
The staff has further 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) sucn activities will be conducted in compliance with the Comission's regulations, and the issuance of this amendment will not be inimical to the comon defense and security or the health and nfety of the public, l
5.0 ACKNOWLEDGEMENT This evaluation was prepared by B. A. Azab.
Dated:
July 27, 1988
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