ML20148J729
| ML20148J729 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/21/1988 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 1500, TAC-65685, NUDOCS 8803300365 | |
| Download: ML20148J729 (3) | |
Text
og Docket No. 50-346 TOLEDO License No. NPF-3 EDISON Serial No. 1500
, c,,,,,,, cm March 21, 1988 DONALO C. SHELTON wnum-ww (4191249 2300 United States Nuclear Regulatory Commission Document Control Desk Vashington, D.C.
20555 Subj ec t : Supplemental Information Regarding the License Amendment Application to Revisc Technical Specification Sections 3/4.3.2 and 3/4.6.3 (TAC No. 65685)
Gentlemen:
In response to a request made during a March 3, 1988 conference call between Toledo Edison representatives, Mr. A. V. DeAgazio (NRC/NRR Davis-Besse Project Manager), and other members of the NRC Staff, Toledo Edison is providing additional information to assist in the review of the subject License Amendment Application. This License Amendment Application was submitted to the NRC on August 7, 1987 (Serial No. 1400) and proposed removing closure time requirements for ten valves connected to the secondary side of the steam generators listed in Technical Specification Tables 3.3-5, Safety Features System Response Times, and 3.6-2, Containment Isolation Valves.
Each NRC question, followed by Toledo Edison's response, is listed below.
Question:
Provide a discussion of the philosophy and objective of this License Amendment Application, including a comparison of the proposed action to other Babcock and Vilcox (B&V) nuclear plant configurations.
Response
Toledo Edison proposed deleting the Safety Features Actuation System (SFAS) closure signal for ten valves (Serial No. 1400) which cause the steam generators to be isolated, primarily following a Large Break Loss of Coolant Accident (LOCA).
As discussed in Serial No. 1400, the specific closure of these valves is not one of the input assumptions to the analysis of this event. Currently, these secondary system valves receive closure signals from both the SFAS and the Steam and Feedvater Rupture Control System (SFRCS).
In the original design for Davis-Besse, the SFAS vas the only system relied upon to actuate the safety related systems. The SFRCS vas added during the later stages of the plant design prior to receipt i
of the Operating License.
Since the SFRCS provides the closure required by 10CFR50 Appendix A, General Design Criteria (GDC) 5', the use of the SFAS to isolate the l
secondary systen valves is not needed.
The existence of multiple isolation signals to these valves results in an increased potential for their spurious closure.
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c Docket No. 50-316 License No. NPF-3 Serial No. 1500 Page 2 Toledo Edison has previously experienced inadvertent SFAS actuations, both at power and while shutdown.
Following the Davis-Besse June 9, 1985 loss-of-all-feedvater event, the NRC acknowledged the safety significance of a spurious closure of the HSIVs.
Item 122.1.C of NUREG-0933, A Prioritization of Generic Safety Issues, concluded that closure of the Main Steam Isolation Valves (HSIVs) vill shut off all main feedvater flov and vill require initiation of auxiliary feedvater to remove the decay heat.
Moreover, once the HSIVs are closed, the reopening of these valves is a rather elaborate procedure. The loss of main feedvater is not easily recoverable under this scenario. Additionally, Item 122.1.C concluded that efforts to reduce the probability of interruption of feedvater flov to the steam generators should be given high priority. Removal of the SFAS closure signals to these valves contributes directly to th!.n goal by preventing inadvertent valve closures, eliminating certain sources of problems which the operator must recognize and respond to, and reducing the potential for the loss of main feedvater flov.
Toledo Edison has surveyed other plant designs for steam generator isolation following a Large Break LOCA, which presently causes steam generator isolation due to a containment vessel high-high pressure condition. Vhile other B&V Nuclear Steam Supply System (NSSS) design plants do not have the specific \\?AS design that Davis-Besse has, this survey indi.ates th.t other B&V plants do not isolate steam generator' on a containment vessel high-high pressure condition.
Additionally, it is noted that the Combustion Engineering NSSS design does not isolate the steam generators based on containment building high-high pressure. The designs considered isolated the steam generators on a lov steam generator pressure or lov main steam line pressure condition.
Since the SFRCS provides this safety function et Davis-Besse, the proposed amendment request is consistent ;1th the Technical Specifications for several other B&V plants and the requirements of GDC 57.
Question:
Confirm that the Steam Generator Tube Rupture (SGTR) Accident Analysis presented in Chapter 15, Accident Analysis, of the Davis-Besse USAR is still valid for offsite dose assessment.
Response
During a steam generator tube rupture event, the preferred mode of cooling is to use the turbine bypass system and the condenser.
This mode of cooldovn minimizes radiation releases to the environment.
If the condenser is not available, the Atmospheric Vent Valves (AVVs) vill be used to control the cooldovn and to minimize the Main Stean Safety Valve (HSSV)
Docket No. 50-346 License No. NPF-3 Serial No. 1500 Page 3 lifts during this event. Thus, if the AVVs are closed by a SFAS signal due to a lov Reactor Coolant System (RCS) pressure signal during this event, the operators vill need to block the SFAS signal so that the AVVs can be used to control the cooldown. The Davis-Besse SGTR analysis has assumed that the affected steam generator is isolated when the RCS pressure is reduced below the losest MSSV setpoint. The radiation dose analysis results presented in USAR Section 15.4.2, Steam Generator Tube Rupture, are based on the assumption that all activity released to the affected steam generator while the RCS is above this pressure is directly released to the environment. Thus, the offsite dose analysis results presented in Chapter 15 of the USAR are not affected and remain valid even if SFAS closure of the AVVs is removed.
Question:
If the closure time requirements for the affected valves are revised to "N/A" (not applicable) in Table 3.6-2, Containment Isolation Valves, and are deleted from Table 3.3-5, Safety Features System Response Times, what assurance is provided that the valve stroke time vill continue to meet other Technical Specification requirements and that valve operability is demonstrated.
Response
At a minimum, operability is required to be demonstrated by the Surveillance Requirements of Technical Specification 3/4.6.3, Containment Isolation Valves.
Stroke time performance vill be monitored as a part of the in-service inspection and testing program, as required by 'lechnical Specification 4.0.5, Surveillance Requirements for inservice inspect: ;n of ASME Class 1, 2 and 3 components.
For those valves with specific stroke time requirements in other Technical Specifications, Toledo Edison vill continue to ensure those requirements are met.
To provide a footnote referencing other Technical Specifications vould require the addition of at least two footnotes to the Table, which already has four footnotes. This vould add to operator burden and is inconsistent with the current format of the Technical Specifications. Consequently, Toledo Edison requests that the License Amendment be approved as proposed in Serial No. 1400.
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Toledo Edison believes the above addresses the NRC request.
Should there be additional questions, please contact Mr. R. V. Schrauder, Nuclear Licensing Manager, at (419) 249-2366.
I Very truly yours, f
DRB/bam l
cc: DB-1 NRC Resident Inspector A. B. Davis, Regional Administrator A. V. DeAgazio, NRC/NRR Davis-Besse Project Manager