ML20070F246
| ML20070F246 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/01/1991 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20070F233 | List: |
| References | |
| 1900, NUDOCS 9103080201 | |
| Download: ML20070F246 (6) | |
Text
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t ' Docket Numbar-50-346
..c.
- License Numbar NPP-3 4 Sarial.Numbar 1900 E,nclosure Page 1 p-APPLICATION FOR AMENDMENT is TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 LAttached is a. requested change to the Davis-Besse Nuclear Power Station.
Unit Number 1 Facility Operating License Number NPP-3 Appendix A, Technical Specifications. Also included is the Safety Assessment _and Significant flazards Consideratlon.-
The proposed change (submitted under cover letter Serial Number 1900) concerns-Technical Specification 3/4.6.1.2, Containment Systems - Containment Leakage.
-Technical Specification Bases-3/4.6.1.2 Containment Systems -
> -Containment Leakage .
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By: ( 'N D. C. Shelton,- Vice President - Nuclear Sworn and- subscribed- before me this lat day of March :1991.
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Docket Numbar 50-346
..- ' uicense Number NPF-3
. Serial'. Number 1900-Enclosure Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station, Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications, Technical Specification 3/4.6.1.2.
A. Time required to implement: -This change is to be implemented within 45 days after NRC issuance of the License Amendment by the NRC.
B. Reason for change (License Amendment Request Number 90-0044): This
-change vill increase the margin between the acceptable secondary containment bypass leakage rate and measured bypass leakage rate. The increased margin vill reduce the potential for othervise unnecessary ecmponent rework and the associated radiation exposure.
C. Salety Assessment and Significant Hazards Consideration: See Attachment.
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Docket Number 50-346
. License Number NPF-3
. Serial. Number-1900 Attachment Pagt 1 of 9 SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR LICENSE AMENDHENT REQUEST NUMBER 90-0044 TITLE A proposed. change to the Davis-Besse Nuclear Powcr Station, Unit No. 1 Operating License, Appendix A, Technical Specification 3/4.6.1.2 containment Leakage and associated Bases.
DESCRIPTION The purpose of this Safety Assessment and Significant Hazards Consideration is to review the proposed change to Davis-Besse Nuclear Power Station Unit No. 1 Technical Sne:ification (TS) 3/4.6.1.2 to ensure that the change does not constitute a significant hazards consideration. The proposed TS change is to increase the allovelle containment bypass leakage rate from 0.015 La to 0.03 La. Due to the namber and type of penetrations in the bypass leakage program, significant component rework can be required to satisfy
-this TS. -The amount of rework is exacerbated as the plant ages and components degiade. Increaaing the allowed bypass leakage rate vill
(;r lste some af the evork borden and any associated radiation exposures.
Ad cluded acs Jministrative changes.
SYS13H5, COMPONENTS AND ACTIVITIES AFFECTED The activity affected is the allowed secondary containment bypass leakage rate. There are no hardware modifications involved.
SAFETY FUNCTIONS OF THE AFFECTED SYSTEMS, COMPONENTS, AND ACTIVITIES The safety function of the TS limit.on the allowed secondary containment bypass leakage rate is to limit the amount of unfiltered leakage from containment so as to keep the resultant doses within 10 CFR Part 100 guideline values. This change would increase the proportion of the containment leakage rate which could be bypass leakage. The overall integrated containment leakage rate is not affected by this change.
EF :CIS ON SAFETY The.allovable containment leakage is defined in terms-of La where La is the overall' integrated leakage rate. La is defined as a leakage rate of 0.50%
(by weight) of the containment air when the containment is pressurized to 38 psig (Pa) per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.- Pa is the peak safety analysis accident pressure.: The containment leakage is enmorised of two components: filtered air and unfiltered air-or bypass leakage. urrently the bypass leakage is limited by TS 3.6.1.2c to 0.015 La. This value, as discussed under Atomic l Energy Commission question 6.2.23 in the Final Safety Analysis Report, was
- an estimation of the leakage at that time based on line size only. This TS change proposes increasing the bypass leakage to 0.03 La. The overall integrated leakage rate, La, is unaffected by this change. Therefore, the consequence of-the proposed change is to increase the fraction of the
Docket Numbar 50 346
. License Number NPF-3
. Serial Number 1900 Attachment Page 2 of 9 containment volume releasei vhich is unfiltered while effectively reducing the fraction released vhich is filtered resulting in a slight increase in the radiological consequer ces. However, the current TS for bypass leakage and the resultant dose cor. sequences are conservative when compared to the confirmatory analysis performed by the NRC during the original licensing process.
The bounding accident for radiological dose consequences is the maximum hypothetical accident (MHA). This accident, as presented in the Updated Safety Analysis Report (USAR), uses the TS value of 0.015 La for the rate of bypass leakege. The confirmatory analysis performed by the NRC, as presented in the " Safety Evaluation Report (SER) Related to Operation of Davis-Besse Nuclear Power Station Unit No. 1" NUREG-0136 Supplement 1, dated April 1977, uses a value of 0.03 La for the rate of bypass leakage. The radiological dose consequences in the SER are accordingly higher than the USAR values.
Due to differences between the SER and USAR analyses and not having a complete NRC input listing, the dose evaluations were reanalyzed using the assumptions given in USAR Section 15.4.6.4. The existing USAR analysis was benchmarked. The dose consequences for the increased bypass leakage vere evaluated using the same assumpticas except for the increased bypass leakage. Consistent with the assumptions it is assumed that the annular region between the Containment Vessel and the Shield Building is at atmospheric pressure upon accident initiation; 13 minutes are required to obtain a negative pressure in that region. It is assumed that all activity escaping the containment Vessel during that time is released directly to the atmosphere without benefit of filtration or mixing. After the negative pressure has been obtained, 3 percent of the leakage is direct to the environment and the other 97 percent of the leakage is collected by the Emergency Ventilation System and exhausted through 95 percent efficient charcoal filters. The following summarizes the results for the thyroid doses.
USAR DOSE (REH) INCREASED BYPASS (REH) SER DOSE (REM)
C0KfROL R00H 13.7 16.44 < GDC 19 (30)
SITE BOUNDARY 210 222.6 279.4 LOV POPULATION ZONE 20 22.8 26.52 Since the increased bypass leakage rate does not affect noble gan releases, the whole body gamma dose due to noble gases are not affected by this change. However, the whole body gamma doses vere reevaluated to determine the impact of increased iodine release or, the whole body gamma dose. These evaluations show that the increase in whole body gamma dose is negligible
- and the gamma dose consequences currently in the USAR are unchanged.
, Dockot Number 50-346
' . License Number NPF-3
. Serial Number 1900 Attachment Page 3 of 9 The above discussion substantiates the confirmatory enalysis previously performed by the NRC as being bounding. Consequently, increasing the bypass leakage from 0.015 La to 0.03 La does not increase the consequences previously analyzed for the HHA even though a more limiting :ase is provided in the USAR.
The following is a discussion of the administrative changes. For specification 3.6.1.2.c, Table 3.6-1 is being deleted since the secondary containment bypass leakage peths are identified in USAR Section 6.2.4. This change is similar to License A6.endment Number 147 which relocated the list of coLtainment isolation valves from the Technical Specifications to the USAR. Also, Bases Section 3/4.6.1.2 is being revised to reference US;.R Section 6.2.4. For Surveillance Requirement 4.6.1.2.d, air locks (Item 1.)
is being relocated to the text and Items 2. and 3. are deleted since the Davis-Besse Nuclear Power Station does r.ot have Items 2. or 3. as part of its design. Consistent with this change Surveillance Requirement 4.6.1.2.g and .h are deleted since they apply only to Items 2. and 3. and the remainder of the Surveillance Requirements are relabeled accordingly.
Surveillance Requirement 4.6.1.2.j also includes editorial changes and the elimination of the reference to 4.6.1.2.d since this would only apply to Iters 2. and 3. which are being deleted.
SIGNIFICANT HAZARDS CONSIDERATION The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordancs vitn d'e proposed changes vould: (1) Not involve a significant increase in the probability or consequences of an accident previously evaluatedt (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. Toledo Edison has reviewed the proposed change and determined that a significant hazards consideration does not exist because operaticn of the Davis-Desse Nuclear Power Station Unit 1 in accordance with these changes vould:
la) Not involve a significant increase in the probability of an accident previously evaluated because there are no design modifications or hardware changes proposed.
lb) Not involve a significant increase in the consequences of an accident previously evaluated because the proposed change does not increase the consequences above those previously analyzed and found acceptable by the NRC in NUREG-0136, Supplement 1.
2a) Not create the possibility of a new kind of accident from any accident previously evaluated because there are no design moditications or hardware changes proposed.
2b) Not create the possibility of a different kind of accident from any accident previously evaluated because there are no design modifications or hardvare changes proposed.
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Docket Numbor 50-346
.= ' License Number NPF-3
.. Serial Number 1900 4ttachment Pa,ge 4 of 9
- 3) Not involve a significant reduction in a margin of safety as defined in the basis for any Technical Specification since the TS vill continue to limit the allowed secondary containment bypass leakage rate and maintain appropriate surveillance requirements.
CONCLUSION On the basis of the above, Toledo Edison has determined that the License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specifications that must be reviewed by the Nuclear Regulatory Commission, this License Amendment Request does not constitute an unreviewed safety question.
ATTACHMENT Attached is the proposed marked-up change to the Operating License.
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