ML20237A260

Application for Amend to License NPF-3,reducing Steam Generator Low Level Trip Setpoint from Greater than 20 Inches to Greater than 16.4 Inches

ML20237A260
Person / Time
Site:	Davis Besse
Issue date:	12/07/1987
From:	Shelton D TOLEDO EDISON CO.
To:
Shared Package
ML20237A250	List: ML20237A248 ML20237A260 ML20237A298
References
1414, TAC-66007, NUDOCS 8712140366
Download: ML20237A260 (22)

Text

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I i Docket No. 50-346-License No. N?F-3 i

Serial No. 1414 Enclosure l Page 1 i

APPLICATION FOR AMENDMENT TO FACILITY OPERATING LICENSE NO. NPF-3 FOR L

l DAVIS-BESSZ NUCLEAR POWER STATION l UNIT NO. 1 l Attached are requested changes to the Davis-Besse Nuclear Power Station, i Unit No. 1 Facility Operating License No. NPF-3. Also included are the Safety Evaluatior. and Significant Hazards Consideration.

The proposed char.ges (submitted under cover letter Seria3 No. 1414) concern

• Section 3/4.3 2, Safety System Instrumentation, Safety Features Actuction System Instrumentation, Table 3.3-12, Steam and Feedwater Pupture Control S.ystem Instrumentation Trip Setpoints.

3 y M '

h D. C. Shelton, Vice President, Nuclear Sworn to and subscribed before me this 7th day of December, 1987.

i .bD{lf l0 ( i Notary Public, State of Ohio My commissica expires _ /E I B7121403$$ gyggg, ,

DR p ADOCK 05000346 PDR

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Docket No. 50-346 License No'. NFF-3 Serial No. 1414 i Enclosure 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, Section 3.3.2.2, Table 3.3-12.

l A. Time Required to Implement: This change is to be implemented within 30 days after NRC issuance of the License Amendment and prior to the beginning of Cycle 6, which is presently scheduled for August, 1983. i B. Reason for Change (Facility Change Request No. 87-0097): This change will minimize inadvertent SFRCS actuations by increasing the margin between the SFRCS Low Level Trip setpoint and the Integrated Control System (ICS) Low Level Limit, thereby improving Main Feedwater availability. This is consistent with the Davis.Besse Course of Action, Appendix C.2.2, Item 4.

C. Safety Eva.1.uation: See attached Safety Evaluation (Attachment No. 1).

D. Summary Significant Hazards Consideration: See attached Summary Significant Hazards Consideration (Attachment 2).

1 E. Significant Hazards Consideration: See attached Significant Hazards Consideration (Attachment No. 3).

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i Docket Nc. 50-346 License No~. NPF-3 Serial No. 1414

' Attachment 1 Page 1 l

SAJETY EVALUATION i o  !

L DESCRIPTION OF PROPOSED ACTIVITIES The purpose of,this safety evaluation is to review a proposed change to the Davis-Besse Nuclear Power Station (DBNPS), Unit No. 1 Qperating License,

Appendix A, Technical Specifications, Section 3.3.2.2, Table 3.3-12 to ensure that no unreviewed safety question exists. This safety evaluation l 1s being performed to meet tbe requirements of-10CFR50.59. l l

This request proposes reducing the steam generator low level Steam and l Feedwater Rupture Control System (SFRCS) trip setpoint from 1 20 inches to 2 16.4 inches. Reducing this setpoint will decrease the likelihood of spurious SFRCS actuations by providing an increased margin between an ICS low level limit of 35 inches and the SFRCS setpoint.

I SYSTEMS AFFECTED The proposed change affects the SFRCS and the Auxiliary Feedwater System (AFWS).

The SFRCS consists of two identical redundant and independent channels.

The four logic channels of the SFRCS are made up of solid state components. In addition to performing several steam and feedwater isolation functions, the SFRCS is also required to ensure an adequate feedwater supply to the steam generators to remove reactor decay heat during periods when the normal feedwater supply and/or forced circulation has been lost.

Reducing the SERCS low level trip setpoint in the Technical Specifications does not affect the present logic configuration of either the SFRCS or  ;

the AFWS. The analysis, as described below, ensures that this change does not increase the demand placed on any system.

DOCUMENTS AFFECTED I

1) Davis-Besse Nuclear Power Station, Unit No. 1 Operating l License, Appendix A, Technical Specifications Section 3.3.2.2, j Table 3.3-12 i

2) Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Report (USAR), July, 1987, Table 7.4-1 and Sections i 15.2.8, 7.4.1.3.10

3) Davis-Besse Setpoint Index M-6205 and Instrument Index M-720I ]

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I Docket No. 50-346 License No. NPF-3 Serial No. 1414 Attachment 1 Page'2 REFERENCES

1. Davis-Besse Nuclear Power Station, Unit No. 1 Operating i License, Appendix A, Technical Specifications Section 3.3.2.2, i Table 3.3-12 l

2. Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety  !

Analysis Report (USAR), July, 1987, Table 7.4-1 and Sections 15.2.8, 7<4.1.3.10

3. Davis-Besse Setpoint Index M-620S, and Instrument Index M-720I

4. Babcock and Wilcox (B&W) Document 32-1159090-01, Davis-Besse i Unit 1 SFRCS Accident Analysis with Licensing Assumptions l 5. Instrument Society of America (ISA) Standard S67.04, Setpoints  !

for Nuclear Safety-Related Instrument Channels in Nuclear Power Plants SAFETY FUNCTIONS OF SYSTEMS AFFECTED The safety function of the SFRCS is to isolate the unaffected steam l generator from either a main steam line break or main feedwater line break, to automatically start the AFWS in the event of a main steam line or main feedwater line rupture, to automatically start the AFWS on the  !

loss of both main feed pumps or the loss of all four RCPs, and to prevent steam generator overfill and sesequent spill ove- into the main steam lines.

The safety function of the AFWS is to provide feedwater to the steam generators for the removal of reactor decay heat in the absence of maiu feedwater and to promote natural circulation of the Reactor Coolant System (RCS) in the event of a loss of all four RCPs.

EFFECTS ON SAFETY  ;

The functional requirement of the SFRCS affected by this Technical Specification change is the initiation of the AFWS following a Loss of .

Feedwater (LOFW) event in a timely manner in crder to remove reactor decay i heat without exceeding an RCS pressure of 2750 psig and without fuel ,

damage.

The performance of the AFWS is described in the USAR Chapter 15 accident analyses for the Loss of Feedwater transient (Section 15.2.8). The norraal LOFW transient puts a more severe design requirement on the AFW system than other USAR Section 15.2.8 transients. The LOFW transient is the only l design basis accident that takes credit for SFRCS actuation due to low steam generator level. The USAR analysis used an AFW flowrate of 800 gpm j to be delivered within 40 seconos of actuation of the SFRCS. Reducing the SFRCS low level trip setpoint increases the time for initiation of AFW.

i Docket No. 50-346 License No. NPF-3 Serial No. 1414 Attachment 1 Page 3 i The analysis performed by B&W in support of reducing the SFRCS low level trip setpoint and the AFW flow requirement, B&W document 32-1159090-01, is a re-analysis of the LOFU described in USAR Section 15.2.8. This new B&W analysis includes the following conservative assumptions:

1) Initial reactor power at 102% Full Power

2) No credit for PORV, pressurizer sprays, or make-up flows

3) 1.2 times ANS 5.1 (1979). decay heat curve

4) Full AFW slow of 600 gpm delivered to the steam generator 40 seconds after SFRCS low level setpoint is reached

5) SFRCS low level setpoint of 10 inches actual level above lower tube sheet

6) Offsite power is available during event. Therefore, RCS pumps continue to operate which is conservative for this analysis and is consistent with the existing Safety Analysis Report.

7) Turbine trip due to reactor trip Excluding the reduced low level SFRCS setpoint, the reduced AFW flow and the use of the 1979 ANS decay heat curve, the above assumptions are consistent with assumptions utilized in the original USAR Chapter 15 LOFW transient analysis. j Cor,sistent with the USAR, Chapter 15, the acceptance criteria for this B&W analysis are:

1) No fuel damage

2) RCS pressure does not exceed 2750 psig In this analysis, the initiating event is a failure of the main feedwater control valves with a subsequent ramp reduction to zero flow in seven seconds. During the feedwater reduction, the temperature and pressure in the RCS begin to increase and cortinue to increase until a high pressure Reactor Protection System (RPS) trip occurs when 2400 psia is reached and the reactor trips at approximately 14 seconds into the event. RCS pressure is then controlled by the pressurizer code safety valves as was the case in the original LOFW analysis. This analysis initiates the turbine trip one second following the reactor trip with the turbine stop valves ramping closed during the next second for a total of two seconds from reactor trip to valve closure. At this point, the secondary side pressure is maintained by the main steam code safety valves while the steam generator inventory is boiling down. When the low level safety .

setpoint.of 10 inches (collapsed liquid level above the lower tube sheet) {

in the steam generator is reached, SFRCS is actuated. This occurs at j approximately 25.6 seconds into the event. Full flow AFW reaches the l steam generator 40 seconds 1 ster with AFW feeding only one steam generator j due to single failure considerations. The AFW flow curve used provides l for a flow of 600 gpm at a 1050 psig steam generator pressure.

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. Docket No.'50-346 License No. NPF-3

. Serial No. 1414 Attachment 1 Page 4 As can be seen from Figures A and B respectively, by 170 seconds into the event a suf ficient water level is established in the steam generator to terminate the temperature-rise. in the RCS and to estrblish RCS cooldown.  ;

This establishes that the AFW flow with the reduced SFRCS low level setpoint is' sufficient for the removal of core decay heat and subsequent cooldown to 280 F. Since the RCS remains subcooled thrcughout the event with no temperature or power excursions which approach the 112% full power limit, there is no fuel damage. Figure C shows the peak RCS pressure is limited to approximately 2590 psia which is below the safety limit of 110% of design pressure (2750 psig). Therefore, the USAR Chapter 15 acceptance criteria have been met. Figure D shows the maximum pressurizer level reached is approximately 350 inches. The pressurizer level does exceed the high end of the scale. However, a steam bubble is maintained at all times in the pressurizer during the event  !

(approximately four vertical feet remaining). This' analysis is conservative in that the modeling techniques'do not consider any cooling

'of the RCS due to AFW wetting on the steam generator tubes.

Additionally, since steam generator pressure remains above 1000 psig at the time of AFW initiation, sufficient steam inventory exists in the steam generators to supply 600 gpm to the steam generator by use of the AFW turbines.

, In order to ensure that the 10 inch low level safety 1-imit is nat

! violated, a calculation (Toledo Edison calculation C-IC-63.01-001) was

! performed in accordance with the . guidelines of ISA Standard S67.04 used for setpoint aualysis to determine the proper low level trip setpoint.

The calculation. determined the total loop uncertainty for the instrument string and the device allowance, which is a combination of the bistable and transmitter drift, for the string. These two terms are then added to the safety limit to yield the Technical Specification Trip Setpoint of 116.4 inches (actual level above the lower tube sheet). Since 0.8 inch has been added to the safety limit to account for bistable drift, the "as found" allowable value for channel functional test is 115.6 inches (actual level above the lower tube sheet). Also, since 2.7 inches have been added to the safety limit to account for transmitter drift the "as found"

allowable value for channel calibration test is 212.9 inches (actual level above the lower tube sheet). Technical Specification Table 3.3-12, therefore, allows for any drif t which may have occurred between calibrations while still ensuring that the safety limit used in the analysis is not being violated. Since the setpoints are defined in terms of the actual level of water above the lower tube sheet, Note (1) of Table 3.3-12 is also being changed for clarity.

Since the LOFW is a bounding transient for the AFWS requirements and has been analyzed with acceptable results, the consequences of all other accidents analyzed in the safety analysis report requiring AFW with the reduced SFRCS low level setpoint are also acceptable.

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Docket No. 50-346 g License No'. NPF-3 i Serial No. 1414 Attachment 1 Page 5 UNREVIEWED SAFETY QUESTION EVALUATION I

The proposed action would not increase the probability of an accident previously evaluated in the USAR because there have been no hardware changes or design modifications which would affect the probability of an accident (10CFR50.59(a)(2)(i)).

The proposed action would not increase the consequence of an accident i previously evaluated in the USAR because the @nalysis, performed for the j LOFW Transient, has shown the reduced SFRCS setpoint and reduced flow is still capable of removing decay heat and meeting the USAR Chapter 15 acceptance criteria. Thus, although the results are different, the consequences are acceptable as they are bounded by other Chapter 15 analyses,(i.e., rod withdrawal accident for peak RCS pressure and LOCA for peak fuel temperature) (20CFR50.59(a)(2)(i)).

The proposed action would not increase the probability of a malfunction of equipment important to safety because there have been no hardware changes or design modifications which would af fect the probability of a malfunction (10CFR50.59(a)(2)(i)).

The proposed action would not increase the consequence of a malfunction of equipment important to safety because the analysis has shown the reduced SFRCS setpoint and reduced flow is still capable of removing decay beat and meeting the USAR Chapter 15 acceptance criteria. Thue the consequences are acceptable as they are bounded by the uther Chapter 15 analyses (i.e., rod withdrawal accident for peak RCS pressure and LOCA for peak fuel temperature) (10CFR50.59(a)(2)(i)).

The proposed action would not create a possibility for an accident of a different type than any evaluated previously in the USAR because there have been no hardware changes and analysis has shown that the SGs, with t reduced SFRCS setpoint and the reduced AFW flow, are capable of removing .

decay heat and cooling the core (10CFR50.59(a)(2)(ii)).

The proposed action would not creste a possibility for a malfunction of  ;

equipment of a different type than any evaluated previously in the USAR 1 because there have been no hardware changes or design modifications which would affect the possibility for a malfunction of equipment of a different type than evaluated previoursly (10CFR50.59(a)(2)(ii)).

The proposed action would not reduce the margin of safety as defined in the bases for the Technical Specifications. Reducing the SFRCS low level setpoint and reducing the AFW flow do increase the RCS temperature and I pressurizer level over that previously analyzed for this specific event I but the results meet the USAR Chapter 15 acceptance criteria and are bounded by other Chapter 15 analyses (i.e., rod withdrawal accident for peak RCS pressure and LOCA for peak fuel temperature). Therefore, reducing the SFRCS low level setpoint does not reduce the overall margin j of safety as analyzed for this plant and consequently does not impact the I intent of the Technical Specification (20CFR50.59(a)(2)(iii)).

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Dochet No. 50-346

' License No. NPF-3 Serial No. 1414 Attachment 1 Page 6 CONCLUSION Pursuant to the above, this change to Technical Specification 3.3.2.2, Table 3.3-12 does not involve an unreviewed safety question.

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-Docket No.'50-346

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Docket No. 50-346  ;

License No. NPF Serial No'. 1414 Attachment 2 Page 1

SUMMARY

SIGNIFICANT HAZARDS CONSIDERATION

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Description of Amendment Request: Thie amendment request proposes reducing the steam generator low level Steam and Feedwater Rupture Control System (SFRCS) trip setpoint from greater than or equal to 20 inches to greater than or equal to 16.4 inches. .Tbls requirement is conta.ined in Technical Specification 3.3.2.2, Table 3.3-12. A clari-Vication of note (1) to this table is also proposed.

Basis for Proposed No Significant Hazards Consideration Determination:

The purpose of the change is to avoid unnecessary challenges to the SFRCS by increasing the margin between the SFRCS low level trip setpoint and the Integrated Control System (ICS) lower level limit. This change is consistent with the Davis-Besse Course of Action Appendix C.2.2, Item 4 in response to the June 9,1985 Loss of Feedwater Event.

Babcock & Wilcox-(B&W) has completed a re-aualysis of the Loss of Feedwater Accident described in the Davis-Besse Updated Safety Analysis Report (USAR), Section 15.2.8 based on the new setpoint. The results of thir analysis clearly meet the USAR Chapter 15 acceptance criteria which are:

1) No fuel damage

2) RCS pressure does not exceed 2750 psig The Commission has provided guidance concerning the application of standards for determining whether license amendments involve significant hazards considerations by providing certain examples, published in the Federal Register on March 6, 1986. One of the examples (vi) of an action involving no significant hazards considerations is a change resulting from the application of a refinement of a previously used calculational model or design method. This change is not the result of a refined calculational method, per se, but rather a refined calculation based on -

more realistic plant specific assumptions. The B&W re-analysis has shown that this change does not exceed design basis bounding values and there- ,

fore, involves no increase in the probability or consequences of an i accident, and no reduction of the margin of safety.

The change to Note (1) of Table 3.3-12 provides clarification of the datums for the measurement of Steam Generator level.

It is therefore concluded that the amendment application involves no significant hazards consideration.

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Docket No. 50-346 License No. NPF-3 ,

Serial No. 1414 1 Attachment 3 Page 1 SIGNIFICANT HAZARDS CONSIDERATION _

DESCRIPTION OF PROPOSED ACTIVITIES The purpose of this Significant Hazards Consideration is to review a proposed change to the Davis-Bessc Nuclear Power Station (DBNPS), Unit No. 1 Operating License, Appendix A, Technical Specifications Section 3.3.2.2, Table 3.3-12, to ensure that no significant hazards consideration existr..

This request proposes reducing the steam generator Icw level Steam and I Feedwater Rupture Control System (SFRCS) trip setpoint from 120 inches to 116.4 inches. In the Davis-Besse Course of Action Report, Appendix C.2.2, Item 4, it was stated that the Decay Heat Removal Task Force recommended increasing the margin between the SFRCS Low Level Trip Setpoint and the integrated Control System (ICS) Low Levc1 Limit prior to the beginaing of fuel. Cycle 6. This action was recommended in order to minimize spurious challenges of the SFRCS during operation at an ICS steam generator low level limit of 35 inches and thus improve main feedwater availability.

This recommendation will be satisfied through the implementation of this proposed Technical Specifications amendment.

SYSTEMS AFFECTED The proposed change affects the SFRCS and the Auxiliary Feedwater System j (AFWS).

The SFRCS consists of two identical redundant ana independent channels.

The four logic channels of the SFRCS are made up of solid state components. In addition to performing several steam and feedwater i isolation functions, the SFRCS is also required to ensure an adequate j feedwater supply to the steam generators to remove reactor decay heat during periods when the normal feedwater supply and/or forced circulation ,

has been lost.

Reducing the SFRCS low level trip setpoint in the Technical Specifications does not affect the present logic configuration of either the SFRCS or the AFWS. The analysis, as described below, ensures that this change does not increase the demand placed on any system.

DOCUMENTS AFFECTED

1) Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A, Technical Specifications Section 3.3.2.2, Table 3.3-12

2) Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Report (USAR), Jtly, 1987, Table 7.4-1 and Sections 1 15.2.8, 7.4.1.3.10

3) Davis-Besse Setpoint Index M-620S and Instrument Index M-7201 l

Docket No. 50-346 License No. NPF-3 l' Serial No. 1414 Attachment 3  !

Page 2 REFERENCES

1. Davis-Besse Nuclear Power Station, Unit No. 1 Operating License, Appendix A, Technical Specifications Section 3.3.2.2, Table 3.3-12

2. Davis-Besse Nuclear Power Station, Unit No. 1 Updated Safety Analysis Rcport (USAR), July, 1987, Table 7.4-1 and Sections 15.2.8, 7.4.1.3.10

3. Davis-Besse Setpoint Index M-620S, and Instrument Index M-720I

4. Babcock and Wilecx (B&W) Document 32-1159090-01, Davis-Besse Unit 1 SFRCS Accident Analysis with Licensing Assumptions

5. Instrument Society of America (ISA) Standard S67.04, Setpoints for Nuclear Safety-Related Instrument Channels in Nuclear Power Plants S_AFETY FUNCTIONS OF. SYSTEMS AFFECTED The safety function of the SFRCS is to isolate the unaffected steam generator from either a main steam line break or main feedwater line break, to automatically start the AFFS in the event of a main steam line or main feedwater line rupture, to automatically start the AFWS on the loss of both main feed pumps or the loss of all four Reactor Coolant Pumps (RCPs), and to prevent steam generator overfill and subsequent spill over into the main steam lines.

The safety function of the AFWS is to provide feedwater to the steam generators for the removal of reactor decay heat in the absence of main feedwater and to premote natural circulation of the Reactor Coolant Syetem (RCS) in the event of a less of all four RCPs.

EFFECTS ON SAFETY The functional requirement of the SFRCS affected by this Technical Specification change is the initiation of the AFWS following a Loss of Feedwater (LOFW) event in a timely manner in order to remove reactor decay heat without exceeding an RCS pressure of 2750 psig and without fuel damage.

The performance of the AFWS is described in the USAR Chapter 15 accident analyses for the Loss of Feedwater transient (Section 15.2.8). The normal LOFW transient puts a more severe design requirement on the AFWS than

other USAR Section 15.2.8 transients. The LOFW transient is the only design basis accident that takes credit f or SFRCS actuation due to low steam generator level. The USAR analysis used an Auxiliary Feedwater (AFW) flowrate of 800 gpm to be delivered within 40 seconds of actuation of the SFRCS. Reducing the SFRCS low level trip setpoint increases the time for initiation of AFW.

Docket No. 50-346 License No. NPF-3 Serial No. 1414 Attachment 3 Page 3 The analysis performed by B&W in cupport of reducing the SFRCS low level trip setpoint and the AFW flow requirement, B&W document 32-1159090-01, is a re-analysis of the LOFW described in USAR Section 15.2.8. This new B&W analysis includes the following conservative assumptions:

1) Initial reactor power at 102% Full Power

2) No credit for PORV, pressurizer sprays, or make-up flows [

3) 1.2 times ANS 5.1 (1979) decay heat curve {

4) Full AFW flow of 600 gpm delivered to the steam generator 40 seconds after SFRCS low level setpoint is reached (The AFW flowrate of 600 gpm is consistent with the Technical Specification Bases change propo;ed by Serial 1360, dated 03/23/87)

5) SFRCS low level setpoint of 10 inches actual level above lower tube sheet

6) Offsite power is available during event. Therefore, RCS pumps I

continue to operate which is conservative for this analysis and is consistent with the existing SAR safety analysis.

7) Turbine trip due to reactor trip l Excluding the reduced low level SFRCS setpoint, the reduced AFW flow, and the use of the 1979 ANS decay heat curve, the above assumptions are l consistent with assumptions utilized in the original USAR Chapter 15 LOFW l transient analysis.

Consistent with USAR Chapter 15, the acceptance criteria for this B&W analysis are:

1) No fuel damage

2) RCS pressure does not exceed 2750 psig In this analysis, the initiating event is a failure of the main feedwater control valves with a subsequent ramp reduction to zero flow in seven seconde. During the feedwater reduction, the temperature and pressure in the RCS begin to increase and continue to increase until a high pressure Reactor Protection System (RPS) trip occurs when 2400 psia is reached and the reactor trips at approximately 14 seconds into the event. RCS pressure is then controlled by the pressurizer code safety valves as was the case in the original LOFW analysis. This analysis initiates the turbine trip one second following the reactor trip with the turbine stop valves ramping closed during the next second for a total of two seconds from reactor trip to valve closure. At this point. the secondary side pressure is maintained by the main steam code safety valves while the steam generator inventory is boiling down. When the low level safety setpoint of 10 inches (collapsed liquid level above the lower tube sheet) in the steam generator is reached, SFRCS is actuated. This occurs at approximately 25.6 seconds into the event. Full AFW flow reaches the steam generator 40 seconds later with AFW feeding only one steam generator due to single failure considerations. The AFW flow curve used, based on previous analyses (and as presented in the proposed Technical Specification Bases change, transmitted by Toledo Edison Serial No. 1360 dated March 23, 1987) provides a flow of 600 gpm at a 1050 psig steam generator pressure.

Docket No. 50-346 License No. NPF-3 Serial No.'1414 Attachment 3 Page 4 i As can be seen from Figures A and B respectively, within 170 seconds irao the event a sufficient water level is established in the steam generator to terminate the temperature-rise in the RCS and to establish RCS cooldown.

This establishes that the AFW flow with the reduced SFRCS low level setpoint is sufficient for the removal of core decay heat and subsequent cooldown to 280*F. Since the RCS remains subcooled throughout the event with no temperature or pcuer excursions which approach the 112% full power limit, there is no fuel damage. Figure C shows that the peak RCS pressure is limited to approximately 2590 psia which is below the safety I limit of 110% of design pressure (2730 psig). Therefore, the USAR i Chapter 15 acceptance criteria are met. Figure D shows the maximum  !

pressurizer level reached is approximately 350 inches. The pressurizer )

level does' exceed the high end of the scale; however, a steam bubble is maintained at all times in the pressurizer during the event (approximately four vertical feet remaining). This analysis is conservative in that the modeling techniques do not consider any cooling of the RCS due to AFW wetting on the steam generator tubes. Additionally, since steam generator .

pressure remains above 1000 psig at the time of AFW initiation, sufficient l steam inventory exists in the steam generators to supply 600 gpm to the l steam generator by use of the AFW turbines. j In order to ensure that the 10 inch low level safety limit is not violated, '

a calculation (Toledo Edison calculation C-IC-63.01-001) was performed in accordance with the guidelines of the Instrument Society of America (ISA)

Standard S67.04 used for setpoint analysis to determine the proper low level trip setpoint. The calculation determined the total loop uncertainty for the instrument string and the device allowance, which is a combination of the bistable and transmitter drift, for the string. These two terms are then added to the safety limit to yield the Technical Specification Trip Setpoint of >16.4 inches (actual level above the lower tube sheet).

Since 0.8 inch has been added to the safety limit to account for bistable drift, the "as found" allowable value for channel functional test is 115.6 inches (actual level above the lower tube sheet). Also, since 2.7 inches have been added to the safety limit to account for transmitter drift, the "as found" allowable value for channel calibration test is 212.9 inches (actual level above the lower tube sheet). Technical Specification Table 3.3-12, therefore , allows for any drif t which may have occurred between calibrations while still ensuring that the safety limit used in the analysis is not being violated.

Since the setpoints are defined in terms of the actual level of water above the lower tube sheet, Note (1) of Table 3.3-12 is also being changed

) to. clarify the datums for measuring SG 1evel.

Since the LOFW is a bounding transient for the AFWS requirements and has been analyzed with acceptable results, the consequences of all other accidents analyzed in the safety analysis report requiring AFW with the reduced SFRCS low level setpoint are also acceptable.

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Docket No. 50-346 License No. NPF-3 Serial No. 1414 Attachment 3 Page 5 SIGNIFICANT HAZARDS CONSIDERATION The 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 these changes would: I

1. Not involve a significant increase in the probability or conr,equences of an accident previously evaluated because there have been no hardware changes or design modifications which would affect the probability of an accident, and because the analysis, performed for the LOFW transient, has shown the reduced SFRCS setpoint and reduced flow is still capable of removing decay heat and meeting the USAR Chapter 15 acceptance criteria. Thus the consequences are acceptable '

since they are bounded by other Chapter 15 analyses (i.e., rod withdrawal accident for peak RCS pressure and LOCA for peak fuel temperature) (10CFR50.92(c)(1)).

2. Not create the possibility of a new or different kind of accident I from any accident previously evaluated because there have been no hardware changes and analysis has shown that the SGs, with ,

reduced SFRCS setpoint and reduced AFW flow, are capable of removing I decay heat and cooling the core (10CFRSC .92(c)(2)). ]

3. Not involve a significant reduction in a margin of safety because, although reducing the SFRCS low level setpoint and reducing the AFW flow do increase the RCS temperature and pressurizer level over that previously analyzed for this specific event, the results meet the USAR Chapter 15 acceptance criteria and are bounded by other Chapter 15 analyses (i.e., rod withdrawal accident for peak RCS pressure and LOCA for peak fuel temperature). Therefore, reducing the SFRCS low level setpoint does not reduce the overall margin of safety as analyzed for this plant. (10CFR50.92(c)(3)).

CONCLUSION On the basis of the above, it is concluded that the proposed change does not involve a significant hazards consideration.

Docket No. 50-346

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