ML20154A806
| ML20154A806 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 05/02/1988 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 1515, TAC-65685, NUDOCS 8805160173 | |
| Download: ML20154A806 (5) | |
Text
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i TOLEDO EDISON A Cesaw trw3, Cmm DonAto C SHELTON
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Docket No. 50-346 (N" "
t License No. NPF-3 Serial No. 1515 May 2, 1988 United States Nuclear Regulatory Commission Document Control Desk Vashington, D. C.
20555 Subj ect:
Supplemental Information Regarding the License Amendment Application to Revise Technical Specifications 3/4.3.2 and 3/4.6.3 (TAC No. 65685)
Gentlemen:
In response to your Request for Additional Information dated March 24, 1988 (Log No. 2528) Toledo Edison is providing additional information to assist in the review of the subject License Amendment ap.11 cation. This License Amendment application was submitted to the NRC on August 7, 1987 (Serial No.
1400), supplemented on March 21, 1988 (Serial No. 1500), and discussed during a meeting in Rockville, Maryland between Toledo Edison representatives.
Mr. A. V. DeAgazio (NRC/NRR Davis-Besse Project Manager), and other members of the NRC Staff. This License Amendment proposes removing closure time requirements for valves connected to the secondary si!e of the steam generators listed in Technical Specification Table 3.3-5, Safety Features System Rt.sponse Times, and Table 3.6-2, containment Isolation Valves.
Each NRC question, followed by Toledo Edison's response, is listed below:
1.
Question:
The proposed changes to the plant Technical Specifications (TS) in the licensee's letter dated August 7, 1987 vill revise Table 3.3-5 Safety Features System Response Times to delete reference to the main steam varmup drain valves and atmospheric vent valves (AVV) receiving a high containment pressure or lov reactor coolant system pressure SFAS automatic signal.
It was indicated that the purpose of this change vas to improve reliability and availability of the Main Feedvater System by reducing the chance of plent trips resulting from an inadvertent SFAS. The primary justification for this change vas that those valves are normally closed during power operations. The SFAS signal serves to provide only a backup to procedural requirements for maintaining the valves in a closed position.
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License No. NPF-3 Serial No. 1515 Page 2 The staff has two concerns with the above proposed TS change:
(1) These valves are normally closed, and an automatic closure of these valves does not isolate the feedvater system.
Therefore, hov enn the elimination of the SFAS automatic signal for MS varsup drain valves and AWs improve reliability of the main feedvater system?:
1 (2) It is required in NUREG-0737, Item II.E.4.2 that following an accident all nonessential systems penetrating containment be automatically isolated. No credit can be given for i
operator action.
By eliminating the SFAS state how this l
requirement is satisfied, or justify why those containment i
isolation valves can be granted a deviation from this j
requirement.
Responses (1) In the Safety Evaluation submitted with the License i
Amendment application, Toledo Edison stated that the primary purpose of removing the Safety Features Actuation System (SFAS) closure signal to the steam generator secondary isolation valves is to improve the reliability and availability of the Main Feedvater System and to minimize challenges to the Auxiliary Feedvater (ATV) System. The r
reliability of the Main Feedvater (MTV) System is impacted by l
l the large valves affected by the proposed application, specifically the Main Steam Isolation Valves (MSIV) and the l
Main Feedvater Isolation Valves.
The position of other valves i-(AVV's, MSIV bypass valves and MS varsup drain valves) is I
inconsequential from an accident analysis standpoint if the
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MSIV remains open and does not receive a closure signal.
l Deleting SFAS closure of the AVV's simplifies operator I
response and control during a Steam Generator Tube l
Rupture (SGTR).
Specifically, with an SFAS Level II closure of the AVV's and the Turbine Bypass System unavailable during a SGTR, the operators vould have to override the AVV SFAS L
signal in order to cool down and depressurize the plant to below the Main Steam Safety Valve set pressure.
Removing the SFAS closure of the MS varsup drain valves benefits the Human Factors engineering of the Control Room by maintaining consistency in the manner by which the Steam Generator secondary side is isolated following postulated accidents, i
Provision of one status and control location for major secondary side valves simplifies operator response to transients.
i (2) In reviewing the valves and systems affected by this change, it has been concluded that for small break Loss of Coolant Accidents (SBLOCA), the availability of the Main Steam, the Main Feedvater, and the Auxiliary Feedvater Systems
(
is desirable for event mitigation and, therefore, these i
systems should not be isolated during a SBLOCA.
During a large break LOCA, automatic isolation vill occur when the r
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dockotNo.50-346 License No. NPF-3 Serial No. 1515 Page 3 Steam Feedvater Rupture Control System (SFRCS) lov steam line pressure condition occurs.
Until the SFRCS induced isolation is completed, the secondary side of the steam generator becomes effectively isolated during a large break LOCA due to the pressure gradients which vill develop between the Reactor Coolant System (RCS) and the steam generator secondary side, i
and the containment vessel to steam generator differential pressures. Automatic closura through the SFRCS, given a Main Steam Line Break or Main Feedvater Line Break, where such isolation is indeed required, vill continue to be available.
For the reasons cited in the response above, it is also desired to make all associated valves respond consistently to a postulated accident.
It is noted that the proposed design is consistent with General Design Criterion 57.
Consequently, it is concluded that the requirements of 10CFR50, Appendix A and NUREG 0737, Item II.E.4.2 vill still be met.
2.
Question: The proposed TS change vill revise TS 3/4.3.2, Table 3.3-5 to delete reference to the atmospheric vent valves, main steau varsup drain valves, main steam isolation valves, main
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feedvater stop valves, and main steam line varsup valves receiving a manual STAS.
It was indicated in a telecon of March 3, 1988 between the licensee and staff that those valves were also listed in Table 3.6-2, Containment Isolation Valves, under TS section 3/4.6.3.
Therefore, the licensee considered it redundant and unnecessary to list those valves in Table 3.3-5.
The staff finds that the surveillance requirements under TS 3/4.3.2 are not the same as the requirements under TS 3/4.6.3.
For exauple, a monthly CHANNEL FUNCTIONAL TEST is required by TS 3/4.3.2 but not required by TS 3/4.6.3.
Identify the dif ferences between these TS requirements and justify your proposed TS for the above valves.
Response
It is noted that the channel check, channel functional test and channel calibration requirements stipulated in Technical Specification Surveillance Requirements (SR) 4.3.2.1.1 and 4.3.4.2.1 only apply to instrument channels and not the actuated equipment (e.g., valves) except for the 18 month response time measurement which does require surveillance testing of the actuated equipment. Vith the proposed change the only instrumentation system applicable to automatic isolation of the affected valves vill be STRCS.
The Survuillance Requirements for the STRCS provided in SR 4.3.2.2.1, therefore, replace the requirements of SR 4.3.2.1.1 for instrument string and output logic surveillance. This vill continue to ensure that the sensors and logic channels which are depended upon for containment isolation are still tested in a manner and on a schedule comparable to that which nov exists.
dockotNo.50-346 License No. NPF-3 Serial No. 1515 Page 4 Actuated equipment vill continue to be tested at least every eighteen months in fulfillment of SR 4.6.3.1.2.
The time response requirement vill be consistent with the most limiting value used in Safety Analysis Report (SAR) analyses. Vhere no specific time assumption was made in the SAR, the only response that is required to be verified is that the valve closes in response to its automatic containment isolation initiation signal. This is consistent with the existing requirements.
3.
Question: The proposed change vill revise TS section 3/4.3.6, Table f
3.6-2 to delete isolation time requirements for the MSIV, MS varmup valves, MFV stop valves, AVV, MS varmup drain valves, and steam generator blevdown valves, along with the deletion of SFAS actuation. The licensee's evaluation of unrevieved safety question has been focused on large break LOCA and MSLB.
The licensee should verify whether there are any other unidentified safety concerns or accident ar.alyses that may be impacted by the proposed changes? For example, confirm the dose consequences for a steam generator tube rupture accident arc within acceptable limits. Confirm that the environmental effects for a small MSLB inside or outside containment are not adversely affected.
Verify that the small break LOCA accident analysis is not adversely affected by the proposed change.
Provide additional diccussion and/or analysis to justify that there is no unreviewed safety question resulting from the proposed change.
Response: The Safety Evaluation submitted by Serial No. 1400 provides discussion and rationale for the primary focus on large break LOCA3 and Main Steam Line Breaks (MSLB).
Vith the present plant configuration, the only accident which can cause an SFAS Level 4 signal is a large break LOCA.
Consequently, when comparing the proposed configuration to the existing plant configuration for analysis impact, only the large break LOCA need be considered for the majority of the valves. The Safety Evaluation also discusses the valves that receive an SFAS Level 2 confirmatory close signal, and why that signal is inconsequential when the MSIV's are still open.
For SBLOCAs, steam generator isolation is not desired to aid in accident mitigation. As stated in the March 21, 1988 supplemental letter (Serial No. 1500), the dose consequences of a Steam Generator Tube Rupture (SGTR) are not affected by this change.
Per the Davis-Besse Updated Safety Analysis Report (USAR),
Section 15.4.2, the consequences of any steam line break which is beyond the capability of the Integrated Control System are bounded by the MSLB analysis. Chapters 3 and 6 of the USAR fully examine the environmental effects of steam line breaks inside and outside of containment.
The only mitigating
bockstNo.50-346 License No. NFF-3 Serial No. 1515 Fage 5 isolation features assumed in these analyses are from SFRCS.
These features are not affected by the proposed change.
Since there is no change being made tc the SFRCS which mitigates such Steam Line Breaks, there is no impact on the USAR analyses.
The applicable accidents have, therefore, been reviewed and, as summarized in the Safety Evaluation contained in Serial No.
1400, an unreviewed safety question does not exist.
Toledo Edison believes the above addresses the NRC request.
If you have any additional questions, please contact Mr. R. V. Schrauder, Nuclear Licensing Manager, at (419) 249-2366.
Very trul
- ours, i
CAB tit cc: DB-1 Resident Inspector A. V. DeAgazio, NRC/NRR Davis-Besse Project Manager A. B. Davis, Regional Administrator State of Ohio