ML20029E610
| ML20029E610 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/16/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20029E606 | List: |
| References | |
| NUDOCS 9405190164 | |
| Download: ML20029E610 (7) | |
Text
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UNITED STATES
{jj [ j i kr; NUCLEAR REGULATORY COMMISSION W ASHINGT ON. D.C. 2055%o001 g '.x,. c j SAFETY _LVALUATION_BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 51 TO FACILITY OPERATING _UCENSE NO. NPF-77 COMMONWEALTH EDISON COMPANY BRAIDWOOD STATION. UNIT NO. 2 DOCKET NO. STN 50-457 1.0 JNTRODUCTION 21, 1994, Commonwealth Edison Company (Ceco, the By letter dated April licensee) submitted an application for a License Amendment for Braidwood, Unit 2.
The proposed amendment would revise Braidwood, Unit 2, Technical Specification (TS) 4.7.1.1 by relieving Unit 2 of compliance with the provisions of TS 4.0.4 until initial entry into Mode 2 following a forced This one-time only change would allow Unit 2 to reach Operational outage.
Mode 3 in order to reset the lift setpoints of 17 main steam safety valves (MSSVs) which are known to have setpoint tolerances greater than the 11% TS The proposed change follows an amendment dated April 18, 1994, issued
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to Braidwood, Units 1 and 2, which granted Unit 2 approval to operate with a d3X tolerance until May 9,1994, at which time the valves were to be reset.
Unit 1, in a refueling outage at the time, was granted relief from the provisions of TS 4.0.4 so that it could reach Mode 3 to reset the valves and restart.
The April 18, 1994, amendment was originally requested after the licensee discovered that the as-left setpoints on certain MSSVs on both units were greater than the TS limit because the testing contractor, Furmanite, incorrectly calculated the valve mean-seat area used in the Trevitesting procedure.
The amendment issued on April 18, 1994, permitted Braidwood, Unit 2, to operate until May 9, 1994, with out-of-tolerance MSSVs. However, on April 5, 1994, Unit 2 experienced a reactor trip with complications due to a failed main power transformer, resulting in a forced outage. The plant must now restart from a cold shutdown condition.
Normally, TS 4.0.4 does not allow entry into a higher operational mode if the Surveillance Requirements associated with, limiting condition for operation (LCO) have not been performed.
However since the MSSVs must be tested and set at the ambient conditions corresponding to nominal plant pressure and temperature, temporary relief from TS 4.0.4 is necessary to permit resetting the valves and to allow for the startup of Unit 2.
It should be noted that the amendment under considernbr. % submitted for, and therefore applies only to, Braidwood, Unit 2.
Granting of this amendment will not remove, for Unit 1 or 2, the note added to TS Table 3.7-2 in the April 18, 1994 amendment, which granted approval of a 13% tolerance until 9405190164 940516 PDR ADDCK 05000457
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. April 21, 1994, which would have removed the note in Table 3.7-2, is not presently being considered.
If a change to this note is desired, a separate amendment request should be submitted for the applicable dockets at a later date.
2.0 EVALUATION The MSSVs at Braidwood were designed and manufactured as Class II components in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 1971 edition. Testing of the valves is performed in accordance with Section XI of the ASME Code. Operability of the MSSVs ensures that secondary system pressure is limited to 110% of its design pressure of (1200 psia for Braidwood) during a turbine trip from 102% rated thermal power with no available path to the condenser (no steam dump capability).
This represents the most severe anticipated operational transient.
An increase on the positive side of the setpoint tolerance would potentially result in the MSSV lifting at a higher pressure, increasing the maximum pressure in the secondary system.
In its submittal of April 21, 1994, CECO assessed the safety impact of plant operation with the higher setpoint tolerance. The accident analyses considered in this application are the same as those considered in the submittal of March 21, 1994, as supplemented by a submittal of March 24, 1994, for the similar amendment issued April 18, 1994.
Specifically, the licensee examined the effect of the increased MSSV setpoint tolerance on the existing licensing basis events analyses as presented in the Updated Final Safety Analysis Report (UFSAR), and concluded that the analyses remain valid with the exception of the loss-of-external load / turbine trip event.
The licensee re-analyzed this event assuming the relaxed tolerance, and determined that all applicable acceptance criteria would continue to be met and that the UFSAR conclusions would remain valid. CECO concluded that the increased as-found setpoint tolerance has no significant impact on any system, operating mode, or accident analysis.
The licensee's findings are consistent with those of other similarly designed pressurized water reactor plants which have been granted relaxed setpoint tolerance for their MSSVs. These include the Seabrook, V.C. Summer, and Fort Calhoun stations, as well as the previously issued amendment for Braidwood.
Additionally,Section XI of the 1989 edition of the ASME Code requires that MSSVs be tested in accordance with ASME/ ANSI OM-1987, Part 1, which permits the tested setpoint pressure to exceed the nominal value by up to 3% before a test failure is declared. A higher tolerance is, therefore, consistent with recent editions of the ASME Code.
On the basis that the setpoints are within 13%, which has been granted to other plants, including the previously issued amendment for Braidwood, Unit 1, and the relatively short duration of the proposed change (until the valves are reset in Mode 3), the staff is satisfied that the MSSVs will continue to accomplish their function with a 13% tolerance, and that entering into Mode 3 with out-of-tolerance MSSVs involves minimal safety significance.
Therefore,
. the staff finds the proposed temporary revision to the TS to be acceptable.
It should be noted that any analyses used in support of future amendment requests for a permanent change of the setpoint tolerance are subject to further staff review.
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3.0 EXIGENT CIRCUMSTANCES
The Commission's regulations, 10 CFR 50.91, contain provisions for issuance of amendments when the usual 30-day public notice period cannot be met. One type of special exception is an exigency. An exigency is a case where the staff and licensee need to act promptly, but failure to act promptly does not involve a plant shutdown, dorating, or delay in startup. The exigency case usually represents an amendment involving a safety enhancement to the plant.
On April 5, 1994, a fault occurred in the main power transformer of Braidwood, Unit 2, resulting in a reactor trip.
Because a control rod stuck out of the core and the transformer was severely damaged, the plant was forced into an outage.
A previously issued amendment dated April 18, 1994, permitted Unit 2 to operate until May 9,1994, with MSSV tolerances of 13%. However, the April 5, 1994, reactor trip prevented the licensee from resetting the valves by May 9, 1994.
The valves must be set at the ambient conditions of the valve corresponding to nominal plant operating pressure and temperature.
Since TS 4.0.4 prevents the plant from changing operational mode with the valves out of tolerance, the provisions of TS 4.0.4 must be temporarily waived to allow Unit 2 to reach Mode 3 to reset the valves and allow the plant to restart.
CECO currently has two large units in forced outages, six other units in outages for equipment repairs, and has outages planned for additional units prior to June 1, 1994.
Delayed issuance of this amendment would prevent the startup of Braidwood, Unit 2, and, in view of the outages at other Ceco facilities, could result in an inadequate supply of available power upon entry into the peak power usage months. The circumstances leading to this request for a TS amendment could not have been avoided since the licensee could not have anticipated the trip which occurred at Braidwood, Unit 2, nor was the situation created by failure of the licensee to submit a timely application for a TS amendment.
Because of the aforementioned circumstances, this amendment is being treated as an exigency, in accordance with 10 CFR 50.91.
The NRC published a public notice of the proposed amendment, issued a proposed finding of no significant hazards consideration and requested that any comments on the proposed no significant hazards consideration be provided to the staff by the close of business on May 12, 1994. The notice was published in the Joliet News Herald and the Morris Daily Herald on May 9,1994. There were no public comments in response to the notices published in the local newspapers.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
The Commission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards considerations if operation of that facility in accordance with the amendment would not:
1.
Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2.
Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3.
Involve a significant reduction in a margin of safety.
Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, the licensee provided its analysis of the issue of no significant hazards consideration which states that:
A.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
In the analysis performed for a i 3% as-found MSSV setpoint, all of the applicable loss of coolant accident (LOCA) and non-LOCA design basis acceptance criteria remain valid both for the transients evaluated and the single event analyzed, Loss of External Load / Turbine Trip.
The MSSVs are actuated after accident initiation to protect the secondary systems from overpressurization.
Increasing the as-found setpoint tolerance will not result in any hardware modification to the MSSVs.
Therefore, there is not an increase in the likelihood of spurious opening of a MSSV.
Sufficient margin exists between the normal steam system operating pressure and the valve setpoint with the increased tolerance to preclude an increase in the probability of actuating the valves.
The peak primary and secondary pressures remain below 110% of design at all times.
The departure from nucleate boiling ratio (DNBR) and peak clad temperature (PCT) values remain within the specified limits of the licensing basis. Although increasing the valve setpoint tolerance may increase the steam release from the ruptured steam generator above the UFSAR value by approximately 2%, the steam generdor tube rupture (SGTR) analysis indicates that the calculated break flow is still less than the value reported in the UFSAR.
Therefore, the radiological analysis indicates that the slight increase in the steam release is offset by the decrease in the break flow such that the offsite radiation doses are less than those reported in the UFSAR. The evaluation also concluded that the existing mass releases used in the offsite dose calculation for the
9 existing mass releases used in the offsite dose calculation for the remaining transients (i.e., steamline break, rod ejection) are still applicable. Therefore, based on the above, there is no increase in the dose releases.
The effects of increased tolerances for MSSV setpoints on the LOCA safety analyses has been previously performed for VANTAGE 5 fuel.
Calculations performed to determine the response to a hypothetica' large break LOCA do not model the MSSVs, since a large break LOCA is characterized by a rapid depressurization of the reactor coolant system below the pressure of the steam generators. Thus, the calculated consequences of a large break LOCA are not dependent upon assumptions of MSSV performance.
Therefore, the large break LOCA analysis results are not adversely affected by revising setpoint tolerances.
The small break LOCA analyses presentcJ in Appendix C of the Byron /Braidwood Stations, Units 1 ana 2, VANTAGE 5 Reload Transition Safety Report were performed using a 3% higher safety valve setpoint pressure.
The standard 3% accumulation between valve actuation and full flow was also accounted for in the analyses.
These analyses calculated peak cladding temperatures well below the allowed 2200*F limit as specified in 10 CFR 50.46 demonstrating that the change to the MSSV setpoint tolerance can be accommodated for small break LOCAs.
Neither the mass and energy release to the containment following a postulated LOCA, nor the containment response following the LOCA analysis, credit the MSSV in mitigat'.ng the consequences of an accident. Therefore, changing the MSSV lift setpoint tolerances would have no impact on the containment integrity analysis.
In addition, based on the conclusion of the transient analysis, the change to the MSSV tolerance will not affect the calculated steamline break mass and energy releases inside containment.
The loss of load / turbine trip event was analyzed in order to quantify the impact of the setpoint tolerance relaxation. As was demonstrated in the evaluation, all applicable acceptance criteria for this event have been satisfied and the conclusions presented in the UFSAR remain valid. The conclusions presented in the Overpressure Protection Report remain valid.
Therefore, the probability or consequences of an accident previously evaluated in the UFSAR would not be increased as a result of increasing the MSSV lift setpoint as found tolerance to 3% above or below the current Technical Specification lift setpoint value.
The probability of an accident occurring will not be affected by granting this amendment request.
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- B.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
No new system configurations are introduced, and no equipment is being operated in a new or different manner than has been previously analyzed. Accordingly, no new or different failure modes are being created.
Increasing the as-left setpoint tolerance on the MSSV does not create the possibility of an accident which is different than any already evaluated in the UFSAR.
Increasing the as-left lift setpoint tolerance on the MSSVs does not introduce a new accident initiator mechanism.
No new failure modes have been defined for any system or component important to safety nor has any new limiting single failure been identified.
No accident will be created that will increase the challenge to the MSSVs and result in increased actuation of the valves. Therefore, the possibility of an accident different than any already evaluated is not created.
C.
The proposed change does not involve a significant reduction in a margin of safety.
Although the proposed amendment is requested for equipment utilized to prevent overpressurization on the secondary side and to provide an additional heat removal path, increasing the as-left lift setpoint tolerance on the MSSVs will not adversely affect the operation of the reactor protection system, any of the protection setpoints or any other device required for accident mitigation.
The proposed increase in the as-left MSSV lift setpoint tolerance will not invalidate the LOCA and non-LOCA conclusions presented in the UFSAR accident analyses.
The new loss of load / turbine trip analysis concluded that all applicable acceptance criteria are still satisfied.
For all the UFSAR non-LOCA transients, the departure from nucleate boiling (DNB) design basis, primary and secondary pressure limits and dose release limits continue to be met.
Peak cladding temperatures remain well below the limits specified in 10 CFR 50.46.
Thus, there is no reduction in the margin of safety.
The staff has completed its review of the licensee's proposed no significant hazards consideration and concludes that the amendments meet the three standards of 10 CFR 50.92(c). Therefore, the staff has made a final determination that the proposed amendment does not involve a significant hazards consideration.
5.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Illinois State official was notified of the proposed issuance of tae amendment.
The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.
The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and that there has been no public ccmment on such finding.
The proposed finding was issued in the local media described in Section 3.0 of this Safety Evaluation. 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 amendment.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that because the requested changes do not involve a significant increase in the probability or consequences of an accident previously evaluated, do not create the possibility of an accident of a type different from any evaluated previously, and do'not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration 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 Contributor:
H. Dawson Date:
May 16, 1994 4
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