ML20237F817
| ML20237F817 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/07/1987 |
| From: | Shelton D TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20237F807 | List: |
| References | |
| TAC-65685, NUDOCS 8708130168 | |
| Download: ML20237F817 (16) | |
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Dock;t No. 50-346 License No. NPF-3 Serial No. 1400 Enclosure Page 1 APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NO. 1 Attached are requested changes to the Davis-Besse Nuclear Power Station, Unit No. 1 Facility Operating License No. NPF-3. Also included are the Safety Evaluation and Significant Hazards Consideration.
The proposed changes (submitted under cover letter Serial No. 1400) concern: l^
Section 3/4.3.2, Safety System Instrumentation; Table 3.3-5, Sarety j Features System Response Times; and ;
Section 3/4.6.3, Containment Isolation Valves, Table 3.6-2, Containment Isolation Valves.
By-D. C. Shelton, Vice President, Nuclear Sworn and subscribed before me this 7th day of August, 1987.
h(O Ik Notary Public, State of Ohio My commission expires [
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8708130168 870007 ;
PDR ADOCK 05000346 ;
P PDR j i
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. 1 Docket No. 50-346 License No. NPF-3 Serial No. 1400
. Enclosure f Page 2 The following information is provided to support issuance of the requested changes to the Davis-Besse Nuclear Power Station, Unit No. 1 Operating License No. NPF-3, Appendix A, Technical Specifications Sections t 3/4.3.2, Safety System Instrumentation and 3/4.6.3, Containment Isolation-l Valves. '
A. Time Required to Implement: This change is to be effective 30 days after issuance of the License Amendment.
B. Reason for Change (FCR No. 87-0110): Revise Technical Specifications by deleting Safety Features Actuation System response times and con-tainment isolation closure time requirements from the valves on the secondary side of the steam generaters in Table 3.3-5, Safety Features System Response Times and Table 3.6-2, Containment Isolation Valves. These valves also receive redundant Steam and Feedwater Rupture Control System closure signals. By deleting the SFAS signal, unnecessary levels of redundancy will be eliminated and plant reli-ability will be improved by reducing the effect of an inadvertent. 1 Safety Features Actuation System actuation.
C. Safety Evaluation: See attached Safety Evaluation (Attachment 1).
D. Significant Hazards Consideration: See attached Significant Hazards !
Consideration (Attachment 2).
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l .Do'cket No. 50-346 License No. NPF-3 Serial No. 1400 Attachment 1 Page 1 l
l SAFETY EVALUATION l
l INTRODUCTION The purpose of this safety evaluation is to review proposed changes to the Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A Technical Specifications. This safety evaluation is being performed to meet the requirements of 10 CFR 50.59 to ensure that no unreviewed safety questions exist with the proposed changes.
The proposed changes involve removing all Safety Features Actuation System (SFAS) and containment isolation Technical Specification closure time requirements from all the valves attached to the secondary sides of the steam generators (SG). These valves (see Table 1) also receive Steam and Feedwater Rupture Control System (SFRCS) signals.
The design requirements for these valves are detailed in 10CFR50, Appendix A, General Design Criterion 57:
Closed System Isolation Valves. Each line that penetrates primary reactor containment and is neither part of the reactor coo 16nt pressure boundary nor connected directly to the containment I atmosphere shall have at least one containment isolation valve which i shall be either automatic, or locked closed, or capable of remote manual operation. This valve shall be outside containment and located as close to the cor.tainment as practical. A simple check valve may not be used as the automatic isolation valve.
Technical Specification 3.3.2, Safety System Instrumentation (Safety Features Actuation System Instrumentation), requires that each valve listed in Technical Specification Table 3.3-5, Safety Features System Response l Times, be operable with the response time listed. The Main Steam Isolation I Valves (MS 100 and 101), Main Steam Warmup Valves (MS 100-1 and 101-1),
l the Main Steam Warmup Drain Valves (MS 375 and MS 394), the Atmospheric l Vent Valves (ICS 11A and B), and the Main Feedwater Stop Valves (FW 601 and l 612) are listed in Table 3.3-5. The Bases sectic.a of 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety System Instrumentation, states that the function of the system is to " provide the overall reliability, redun-dance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses."
Technical Specification 3.6.3, Containment Isolation Valves, requires that all valves listed in Technical Specification Table 3.6-2, Containment Isolation Valves, be operable with the listed isolation times. All the l
valves affected by the proposed Technical Specification revision are i listed in Table 3.6-2. Bases Section 3/4.6.3, Containment Isolation )
Valves, states this requirement ensures "that the containment atmosphere i
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Docket No. 50-346 License No. NPF-3 Serial No. 1400 1 Attachment 1 l Page 2 )
I will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of
~t he containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA."
The analyses in the Final / Updated Safety Analysis Report (FSAR/USAR) have assumed that the secondary side of.the SG is isolated automatically following a large break Loss of Coolant Accident (LOCA). For Davis Besse, this is the only class of accident which can cause.a SFAS initiation on containment high-high pressure (Incident Level 4) containment isolation signal, which is the only automatic containment isolation closure signal i provided to any of the affected valves that would be normally open. The Main Steam Warmup Drain Valves (MSDV) (MS 375 and 394) and the Atmos-pheric Vent Valves (AVV) (ICS 11A and B) receive SFAS Level 2 close signals; however, they are normally closed during normal plant operation. Since they are already shut, the SFAS Level 2 signal serves to provide only a backup to procedural requirements of maintaining the valve in a closed position. Additionally, Davis-Besse has taken credit for the automatic ,
SFAS closure of the Main Steam Isolation Valves (MSIV) to cause a SFRCS SG/ main feedwater high reverse differential pressure trip. This will cause any of the affected valves which do not directly receive a SFAS Level 4 signal to close. The NRC accepted this method of establishing containment isolation in the Safety Evaluation Report supporting License l Amendment No. 79 for Davis-Besse (Log No. 1659). I I
In order to improve reliability and availability of the Main Feedwater System, i which is the primary means of SG heat removal, and to minimize challenges to the Auxiliary Feedwater System, this application proposes to change Technical Specification Table 3.3-5 by eliminating all reference to the affected valves and to revise Technical Specification Table 3.6-2 by deleting all closure time requirements for the affected valves. This will I improve plant reliability by reducing the effect of inadvertent SFAS actu- I ations which can cause a loss of main feedwater. Since the affected i valves receive SFAS and SFRCS signals during large break LOCAs, deleting
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one signal also eliminates unneeded redundancy. Each proposed change is fully described and justified in the following sections of this evaluation.
SYSTEMS AFFECTED Safety Features Actuation System (SFAS)
Containment System Steam and Feedwater Rupture Control System (SFRCS)
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Dockee No. 50-346 License No. NPF-3 Serial No. 1400 Attachment 1 Page 3 REFERENCES Davis-Besse Nuclear Power Station, Unit No. 1, Updated Safety Analysis Report, June, 1986; Davis-Besse Nucicar Power Station, Unit No. 1, Operating License, Appendix A, Technical Specifications.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 79 to Facility Operating License NPF-3, Attachment to NRC Letter to Toledo Edison, dated December 11, 1984 (Log No. 1659)
FUNCTIONS OF SYSTEMS AFFECTED 1
The SFAS senses adverse containment conditions which indicate a LOCA may have occurred. Based on predetermined severity levels, the SFAS, among other actions, automatically establishes containment leaktight integrity by performing its containment isolation function. A design basis LOCA will cause essentially complete containment isolation. A l SFAS Incident Level 4 condition, which is caused by a high-high con-tainment pressure (38.4 psia), is the containment isolation level which causes all the valves listed in Table 1 to close automatically cxcept for l the MSDvs (MS 375 and MS 394), the AVVs (ICS 11A and B), and the Auxiliary Feedwater (AFW) isolation valves (MS 107, MS 106A, and AF 599 or MS 106, i MS 107A, and AF 608). The MSDVs and the AVVs receive a SEAS Incident Level i
2 signal (Reactor Coolant System low pressure (1650 psig) or containment l' high pressure (18.4 psia)). Operation of the AFW isolation valves is con-trolled by the SFRCS, as described below.
The Containment System is designed to withstand the worst case Design l Basis Accident (DBA) postulated for the Davis-Besse plant in order to l limit the consequences to the public of such accidents. For breaks in the Reactor Coolant System (LOCAs), the containment accomplishes this function by establishing predetermined levels of leak tight integrity. The contain-ment structure has also been analyzed to ensure that it can withstand the effects of a Main Steam Line Break (MSLB) or main feedwater line rupture l within it. As demonstrated in USAR Section 6.2, the pressurization con-sequences of these accidents are less severe than the worst case LOCA pressure transient. Further the radiological consequences of a LOCA are much more severe than that due to a MSLB or a main feedwater line rupture as established in USAR Sections 15.4.6, 15.4.4, and 15.2.8. The function of the containment is to establish and maintain designated levels of leak tight integrity following such high energy line faults or large releases of radiation occurring within the containment.
The SFRCS is designed to detect and mitigate the effects of major main steam / main feedwater upsets including MSLBs, main feedwater line ruptures, l loss of main feedwater events, SG overfeeding events, and a loss of Reactor Coolant System (RCS) forced circulation flow. The SFRCS performs
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. l Docket No. 50-346 License No. NPF-3 )
- Serial No. 1400 Attechment 1 '
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1 its design functions by automatically positioning valves and initiating AFW to the SGs, as required. The system detects a MSLB by sensing a main ]'
steam line Icw pressure condition in the faulted SG's steam line. The system senses a feedwater line rupture by detecting a steam line low i pressure on the faulted SG or a high reverse differential pressure between i the unaffected SG and its feedwater line. A steam generator overfill I event will also cause MSIV closure. A SG low level signal, indicative of l a loss of main feedwater event and a loss of all reactor coolant pump signal (indicative of a loss of RCS forced circulation event) cnly initiates AFW to the SGs and does not cause any SG isolation action.
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EFFECTS ON SAFETY J The effect of revising the SFAS and containment isolation requirements provided by Technical Specifications 3.3.2 and 3.6.3 must be' evaluated for impact on the USAR Chapter 15 analyzed events and the environment.
I As previously stated, except for the MSDVs the AVVs, and the AFW isolation j valves, only a SFAS Level 4 signal can cause any of the affected valves, which are normally open, to close for containment isolation purposes.
A Level 4 signal can only be caused by a large break LOCA. During this accident, the RCS will rapidly depressurize to a value below the saturation pressure of the water in the secondary side of the SG. This effec tively j seals the SG as a leakage path from the RCS. As the SG secondary side pressure falls to approximately 600 psia, a SFRCS steam line low pressure trip will occur. The SFRCS trip causes all SG isolation valves associated with the Feedwater and Main Steam System including the SG blowdown valves (MS 603 and MS 611) to automatically close on both SGs. Additicaally, the AFW isolation valves for the first SG, (MS 106, MS 107A, AF 608 or MS 106A, MS 107, .d AF 599) depressurizing below the low pressure setpoint, will automatically close due to the SFRCS trip. Because the SGs are not being used as a heat sink during this accident, remote manual operator action can be used to close any of the valves which did not automatically close. This action is not time dependent however, since the SGs will continue to stay above RCS pressure for an extended period of tire thereby preventing leakage from the RCS to the SG due to the pressure gradient.
The containment pressure response to a large break LOCA, discussed in USAF Section 6.2, will never exceed the pressure which exists in the SG and Main Steam System. This prevents any leakage to the environment from the containment atmosphere. Again, remote manual operator action can follow the automatic SFRCS closure of the affected valves to ensure complete containment isolation does occur. Consequently, the valves affected by the proposed Technical Specification meet all the requirements of General Design Criterion 57. Therefore deletion of all SFAS Level 4 closure
( signals, from the Technical Specifications, for the affected valves listed
! in Table 1 is considered to be acceptable from the stand point of containment integrity and radiological health and safety of plant personnel and the public.
Docket No. 50-346 ;
License No. NPF-3 Serial No. 1400 ;
Attachment 1 Page 5 ;
I Any LOCA which can gause a SFAS Level 4 signal falls into the large break (greater than 0.5ft ) LOCA classificat'on. The USAR Chapter 15 accident analyses, Section 15.4, demonstrate that for large break LOCAs the SGs rapidly become an energy source due to the depressurization of the RCS. ,
The SGs are therefore not relied upon to help mitigate the large break LOCA. Consequently, the changes to the-secondary side valves in response to a large break LOCA as proposed above, have conservative effects on the 1 accident analysis presented in the USAR. Therefore deletion of all SFAS l and containment isolation requirements for the affected valves from the Technical Specifications and removal of automatic closure signals from any i of the affected valves is acceptable from an accident mitigatiou perspective. 'l The AVVs and t,he MSDVs are normally closed during power operations. l They presently receive an SFAS Level 2 signal, which is caused by etuar a i containment vessel high pressure (18.4 psia) or a RCS low pressure (approx- )
imately 1650 psia). Since these valves are aligned normally closed, and l significant other flow paths from the SGs, such as the main steam lines j (which do not get isolated on an SFAS Incident Level 2 signal) exist during J accidents wh1ch could cause a Level 2 SFAS signal, there is no need for an 1 SFAS Level 2 closure signal to the AVVs and MSDVs. Additionally, these valves l would receive close signals on either an SFRCS low steam pressure signal l or SG/ main feedwater high reverse differential pressure signal and can also be closed manually from the control room. Consequently, removal of the SFAS Level 2 signal from these valves, and fran the Technical Specifi-cations for these valves is acceptable from both an environmental and an accident analysis perspective and still satisfy the requirements of General Design Criteria 57.
UNREVIEWED SAFETT QUESTION EVALUATION Revising the Technical Specification requirements as proposed and removing SFAS closure signcis from any of the valves listed in Table 1 will not increase the probability of occurrence of any previously analyzed accident because the plant will be operated and tested the same as before the change except that automatic SFAS closure for the affected valves will not be tested. Proper operability of all the affected valves will still be established during SFRCS testing. Consequently, there is no change in
- he probability of occurrence of any previously analyzed accident (10CFR50.59 l ( a) (2) (i)) .
1 Revising the Technical Specification requirements as proposed and removing SFAS closure signals from any of the valves listed in Table 1 will not increase the consequences of any event previously analyzed in the USAR.
As discussed above, due to the pressure gradients which exist following a l large break LOCA, no additional leakage path for radioactive contamination l to reach the environment will exist as a result of this change for those valves which would close following SEAS Level 4 signals. The response of the plant as a result of the change will be no more severe than previously analyzed. Therefore, the consequences of the events analyzed in the USAR ,
are not increased (10CFR50.59(a)(2)(1)) .
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Revising the Technical Specification requirements as proposed and removing I SFAS closure signals from any of the valves listed in Table 1 will not 1 l increase the probability of occurrence of a malfunction of equipment l important to safety previously evaluated in the USAR. All equipment )
1 important to safety will still be required to demonstrate proper operability at the same frequency as before so that there is no increased probability of its malfunctioning (10CFR50.59(a)(2)(1)).
Revising the Technical Specification requirements as proposed and removing SFAS closure signals from any of the valves listed in Table 1 will not increase the consequences of a malfunction of equipment important to !
safety previously evaluated in the USAh, since all previously analyzed I events, with appropriate failures included, are still within the bounds of l the USAR. There is no increase in the consequences of the previously analyzed events which included all appropriate failures; therefore, the consequences of these failures have teen found acceptable (10CFR50.59 (a) (2) (1)) .
Revising the Technical Specification requirements as proposed and removing SFAS closure signals from any of the valves listed in Table 1 will not create the possibility for an accident of a different type than any )
previously evaluated in the USAR since proper operability of the valves is !
i still assured, operation of the plant will be the same as before, and all failure modes of equipment are the same as before. Consequently an j j accident of a different type than previously analyzed can not occur 1
l l Revising the Technical Specification requirements as proposed and removing 4
! SFAS closure signals from any of the valves listed in Table 1 will not create the possibility of a malfunction of a different type than rny l previously evaluated in the USAR. The new configuration will still be ]
bounded by the existing single failure analyses in the USAR and the
, equipment will still complete its assigned safety function (10CFR50.59 (a)(2)(ii)'
Revising the fechnical Specification requirements as proposed and removing I SFAS closure signals from any of the valves listed in Table 1 does not I reduce the margin of safety as defined in the bases for any Technical I Specification. This is because the currently installed automatic SFAS signals are redundant with the SFRCS low steam line pressure trip signals supplied to these valves and the remote manual operation capability .
I available for these valves. By utilizing the SFRCS trip, and the actual existence of the pressure gradients which exist following a large break LOCA, all the assumptions made in the USAR accident analyses are met and the consequences are within the previously analyzed bounds. Therefore, the results of those analyses, which form the bases of the Technical Specifications are still valid with no decrease in the margins provided by the assumptions of the analyses (10CFR50.59(a)(2)(iii)).
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Docket No. 50-346 License No. NPF-3 Serial No. 1400 Attachment 1 Page 7 1
CONCLUSION Based on the above, it is concluded that the proposed Technical Specification changes do not constitute an unreviewed safety question.
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Docket No. 50-346 License'No. NPF-3 Serial No. 1400
~ Attachment 1 Page 8 TABLE 1: SECONDARY SYSTEM CONTAINMENT ISOLATION VALVES i
SFRCS Low -
SFAS Steam Line. Capability Normal Valve Valve Level Pressure Trip of Remote Operating
- Number Function Signal Position Operation- Position j
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MS 100 MSIV 4 Close Yes Open MS 100-1 MS1V Bypass #1 4 Close Yes Closed MS 101 MSIV 4 Close Yes Open MS 101-1 MSIV Bypass #2 4 Close Yes Closed MS 375 MS Drain #2 2 Cloce Yes' Closed I MS 394 MS Drain #1 2 Close .Yes closed ICS 11A AVV #2 2 Close Yes Closed ICS 11B AVV #1 2 Close Yes Closed MS 603 SG Blowdown N/A Close Yes Closed MS 611 SG Blowdown N/A Close Yes Closed; FW 601 MFW Isolation 4 -Close Yes Open FW 612 MFW Isolation 4 Cicse Yes Open-AF 599 AFW #2 to SG #2 N/A Open/Close .Yes Open AF 608 AFW #1 to SG #1 N/A Close/Open Yes Open MS 107 #2 MS to AFPT #2 N/A Open/Close Yes Closed i MS 106 #1 MS to AFPT #1 N/A Close/Open Yes Closed-MS 106A #2 MS to AFPT #1 N/A Open/Close Yes Open MS 107A #1 MS to AFPT #2 N/A Close/Open Yes Open l
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Decket No. 50-346 License No. NPF-3 Serial.No. 1400 Attachment 2 Page 1 SIGNIFICANT HAZARDS CONSIDERATION l _I_ INTRODUCTION The purpose of this significant hazatda consideration is to review proposed changes to the Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A Technical Specifications.
The proposed changes involve removing all Safety Features Actuation System (SFAS) and containment isolation Technical Specification closure time requirements from all the valves attached to the secondary sides of the steam generators (SG). These valves (see Table 1 of the safety evaluation) also receive Steam and Feedwater Rupture Control System (SFRCS) signals.
The design requirements for these valves are detailed in 10CFR50, Appendix A, General Design Criterion 57:
Closed System Isolation Valves. Each line that penetrates primary l reactor containment and is neither part of the reactor coolant
! pressure boundary nor connected directly to the containment atmosphere l shall have at least one containment isolation valve which shall be l
either automatic, or locked closed, or capable of remote manual ,
operation. This valve shall be outside containment and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve.
Technical Specification 3.3.2, Safety System Instrumentation (Safety Features Actuation System Instrumentation), requires that each valve listed in Technical Specification Table 3.3-5, Safety Features System Response Times, be operable with the response time listed. The Main Steam Isolation Valves (MS 100 and 101), Main Steam Warmup Valves (MS 100-1 and 101-1), 1 the Main Steam Warmup Drain Valves (MS 375 and MS 394), the Atmospheric Vent Valves (ICS 11A and B), and the Main Feedwater Stop Valves (FW 601 and 612) are listed in Table 3,3-5. The Bases section of 3/4.3.1 and 3/4.3.2, Reactor Protection System and Safety System Instrumentation, states that the function of the system is to " provide the overall reliability, redun-dance and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses."
Technical Specification 3.6.3, Containment Isolation Valves, requiree that all valves listed in Technical Specification Table 3.6-2, Containment Isolation Valves, be operable with the listed isolation times. All the valves affected by the proposed Technical Specification revision are Ifsted in Table 3.6-2. Bases Section 3/4.6.3, Containment Isolation Valves, states this requirement ensures "that the containment atmosphere ,
will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of
Docket No. 50-346 License No. NPF-3 j Serial No. 1400 1 Attachment 2 l Page 2 I I
i the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA."
j The analyses in the Fincl/ Updated Safety Analysis Report (FSAR/USAR) )
i have assumed that the secondary side of the SG is isolated automatically following a large break Loss of Coolant Accident (LOCA). For Davis-Besse, this is the only class of tecident which can cause a SFAS initiation on j containment high-high pressure (Incident Level 4) containment isolation signal, which is the only automatic containment isolation closure signal ]
provided to any of the affected valves that would be.normally open. The i Main Steam Warmup Drain Valves (MSDV) (MS 375 and 394) and the Atmos- j pheric Vent Valves (AVV) (ICS 11A and B) receive SFAS Level 2 close signals; j however, they are normally closed during normal plant operation. Since they are already shut, the SFAS Level 2 signal servys to provide only a backup to procedural requirements of maintaining the valve in a closed position. Additionally, Davis-Besse has taken credit for the automatic SFAS closure of the Main Steam Isolation Valves (MSIV) to cause a SFRCS SG/ main feedwater high reverse differential pressure trip. This will cause any of the affected valves which do not directly receive a SFAS Level 4 signal to close. The NRC accepted this method of establishing l containment isolation in the Safety Evaluation Report supporting License !
l Amendment No. 79 for Davis-Besse (Log No. 1659).
In order to improve reliab111ty and availability of the Main Feedwater System, j which is the primary means of SG heat removal, and to minimize challenges
! to the Auxiliary Feedwater System, this application proposes to change Technical Specification Table 3.3-5 by eliminating all reference to the ;
affected valves and to revise Technical Specification Table 3.6-2 by -j l deleting all closure time requirements for the affected valves. This will
, improve plant reliability by reducing the effect of inadvertent SFAS actu-ations which can cause a loss of main feedwater. Since the affected valves receive S?AS and SFRCS signals during large break LOCAs, deleting one signal also aliminates unneeded redundancy. Each proposed change is fully described end justified in the following sections of this evaluation.
SYSTEMS AFFECTED Safety Features Actuation System (SFAS)
Containment System St;am and Feedwater Rupture Control System (SFRCS)
REFERENCES Davis-Besse Nuclear Power Station, Unit No. 1, Updated Safety Analysis Report, June, 1986; i
Docket No. 50-346 License No. NPF-3 l Serial No. 1400 l Attachment 2 Page 3 Davis-Besse Nuclear Power Station, Unit No. 1, Operating License, j Appendix A, Technical Specifications.
Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 79 to Facility Operating License NPF-3, Attachment to NRC Letter to Toledo Edison, dated December 11, 1984 (Log No. 1659)
FUNCTIONS OF SYSTEMS AFFECTED The SFAS senses adverse containment conditions which indicate a LOCA may have occurred. Based on predetermined severity levels, the SFAS, among other actions, automatically establishes containment leaktight integrity by performing its containment isolation function. A design basis LOCA will cause essentially complete containment isolation. A SFAS Incident Level 4 condition, which is caused by a high-high con-tainment pressure (38.4 psia), is the containment isolation level which causes all the valves listed in Table 1 to close automatically except for the MSDVs (MS 375 and MS 394), the AVVs (ICS 11A and B), tnd the Auxiliary Feedwater (AFW) isolation valves (MS 107, MS 106A, and AF 599 or MS 106, MS 107A, and AF 608). The MSDVs and the AVVs receive a SEAS Incident Level 2 signal (Reactor Coolant System low pressure (1650 psig) or containment high pressure (18.4 psia)). Operation of the AFW isolation valves is con-trolled by the SFRCS, as described below.
The Containment System is designed to withstand the worst case Design Basis Accident (DBA) postulated for the Davis-Besse plant in order to limit the consequences to the public of such accidents. For breaks in the l Reactor Coolant System (LOCAs), the containment accomplishes this function j by establishing predetermined levels of leak tight integrity. The contain-l ment structure has also been analyzed to ensure that it can withstand the effects of a Main Steam Line Break (MSLB) or main feedwater line rupture l within it. As demonstrated in USAR Section 6.2, the pressurization con-sequences of these accidents are less severe than the worst case LOCA pressure transient. Further the radiological consequences of a LOCA are i much more severe than that due to a MSLB or a main feedwater line rupture
! as established in USAR Sections 15.4.6, 15.4.4, and 15.2.8. The function i
of the containment is to establish and maintain designated levels of leak tight integrity following such high energy line faults or large releases of radiation occurring within the containment.
The SFRCS is designed to detect and mitigate the effects of major main steam / main feedwater upsets including MSLBs, main feedwater line ruptures, loss of main feedwater events, SG overfeeding events, and a loss of Reactor Coolant System (RCS) forced circult. tion flow. The SFRCS performs its desL.m ? unctions by automatically positioning valves and initiating AFW to be SGs, as required. The system detects a MSLB by sensing a main
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steam line low pressure condition in the faulted SG's steam line. The nystem senses a feedwater line rupture by detecting a steam line low pressure on the faulted SG or a high reverse differential pressure between the unaffected SG and its feedwater line, A steam generator overfill
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l Docket No. 50-346 License No. NPF-3 Serial No. 1400 Attachment 2 ;
l Fage 4 j event will also cause MSIV closure. A SG low level signal, indicative of a loss of main feedwater event and a loss of all reactor coolant pump signal (indicative of a loss of RCS forced circulation event) only initiates AFW to the SGs and does not cause any SG isolation action. I EFFECTS ON SAFETY The effect of revising the SFAS and containment isolation requirements provided by Technical Specifications 3.3.2 and 3.6.3 must be evaluated for impact on the USAR Chapter 15 analyzed events and the environment.
As previously stated, except for the MSDVs, the AVVs, and the AFW isolation i valves, only a SFAS Level 4 signal can cause any of the affected va3ves, )
which are normally open, to close for containment isolation purposes. l A Level 4 aignal can only be caused by a large break LOCA. During this j j accident, the RCS will rapidly depressurize to a value below the saturation ;
l pressure of the water in the secondary side of the SG. This effectively I l seals the SG as a leakage path from the RCS. As the SG secondary side l i pressure falls to approximately 600 psia, a SFRCS steam line low pressure j l trip will occur. The SFRCS trip causes all SG isolatinn valves associated with the Feedwater and Main Steam System including the SG blowdown valves l (MS 603 and MS 611) to automatically close on both SGs. Additionally, the AFW isolation valves for the first SG, (MS 106, MS 107A, AF 608 or MS 106A, MS 107, and AF 599) depressurizing below the low pressure setpoint, l will automatically close due to the SFRCS trip. Because the SGs are not l being used as a heat sink during this accident, remote manual operator action can be used to close any of the valves which did not automatically close. This action is not time dependent however, since the SGs will continue to stay above RCS pressure for an extended period of time thereby preventing leakage from the RCS to the SG due to the pressure gradient.
The containment pressure response to a large break LOCA, discussed in USAR Section 6.2, will never exceed the pressure which exists in the SG and Main Steam System. This prevents any leakage to the environment from the containment atmosphere. Again, remote manual operator action can follow the automatic SFRCS closure of the affected valves to ensure complete containment isolation does occur. Consequently, the valves affected by the proposed Technical Specification meet all the requirements of Get.eral Design Criterion 57. Therefore deletion of all SFAS Level 4 closure signals, from the Technical Specifications, for the affected valves Ifsted in Table 1 is considered to be acceptable from the stand point of containment integrity and radiological health and safety of plant personnel and the public.
Any LOCA which can gause a SFAS Level 4 signal falls into the large break (greater than 0.5ft ) LOCA classification. The USAR Chapter 15 accident analyses, Section 15.4, demonstrate that for large break LOCAs the SGs rapidly become an energy source due to the depressurization of the RCS.
The SGs are therefore not relied upon to help mitigate the large break LOCA. Consequently, the changes to the secondary side valves in response to a large break LOCA as proposed above, have conservative effects on the
Docket No.-50-346 License No. NFF-3 Serial No. 1400 Attachment 2 Page 5 I
I accident analysis presented in the USAR. Therefore deletion of all SFAS I and containment isolation requirements for the affected valves from the f Technical' Specifications and removal of automatic closure signals from any I of the affected valves is acceptable from an accident mitigation perspective.
The AVVs and the MSDVs are normally closed during power operations.
They presently receive an SFAS Level 2 signal, which is caused by either a containment vessel high pressure (18.4 psia) or a RCS low pressure (approximately 1650 psia). Since these valves are aligned normally closed, and significant other flow paths from the SGs, such as the main .
steam lines (which do not get isolated on an SFAS Incident Level 2 signal) exist during accidents which could cause a Level 2 SFAS signal, there is no need for an SFAS Level 2 closure signal to the AVVs and MSDVs. Addi-tionally, these valves would receive close signals on either an SFRCS low steam pressure signal or SG/nain feedwater high reverse differential !
pressure signal and can also be closed manually from the control room. ,
Consequently, removal of the SFAS Level 2 signal from these valves, and from the Technical Specifications for these valves is acceptable from ]'
both an environmental and an accident analysis perspective and still satisfy the requirements of General Design Criteria 5/.
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SIGNIFICANT HAZARDS CONSIDERATION "
Tne proposed changes do not involve a significant hazards consideration because the operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in accordance with this change could not:
Involve a significant increase in the probability or consequences of an accident previously evaluated because the plant will be operated and 3 l tested the same as before except automatic SFAS closure for the affected j valves will not be tested, however proper operability of all the affected i valves will still be established during SFRCS testing (10CFR50.92(c)(1)) . j Create the possibility of a new or different kind of accident previously evaluated because proper operability of the valves is still assured, operation of the plant will be the same as before, and all failure modes of equipment are the same as before (10CFR50.92(c)(2)). ;
Involve a significant reduction in a margin of safety because the currently installed automatic SFAS signals are redundant with the SFRCS low steam pressure trip signals supplied to these valves and the capability for remote, manual operation is available for these valves. By utilizing the SFRCS trip and the pressure gradients which exist following a large break LOCA, all the accident analysis assumptions made in the USAR are met and the consequences are within previously analyzed bounds. Therefore, the results of those analysis which form the Bases of the Tachnical Specifications are still valid with no decrease in the margins provided by l the analysis assumptions (10CFR50.92(c)(3)) . j i
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i Docket No. 50-346 License No. NPF-3 ,
Serial No. 1400 Attachment 2 Page 6 CONCLUSION Based on the above, it is concluded that the proposed Technical Specific-ation changes do not constitute a significant hazards consideration.
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