ML20072J587
| ML20072J587 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/22/1994 |
| From: | Stetz J, Jeffery Wood CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20072J536 | List: |
| References | |
| NUDOCS 9408290102 | |
| Download: ML20072J587 (21) | |
Text
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Docket Number 50-346 1 License Number NPF-3 Serial Number 2231 ,
Enclosure 1 1 APPLICATION FOR AMENDMENT I
1 TO FACILITY OFERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 Attached are the requested changes to the Davis-Besse Nuclear Power Station, Unit .iumber 1 Facility Operating License Number NPF-3. Also included is the Safety Assessment and Significant Hazards Consideration. ,
t The proposed changes (submitted under cover letter Serial Number 2231) concern: i Technical Specifications Section 3.1.1.1 (Shutdown Margin)
Technical Specifications Section 3.4.5 (Steam Generators)
Technical Specifications Bases 3/4.4.5 (Steam Generators)
For: J. P. Stetz, Vice Pre ident - Nuclear By: c~ [tu ,
J7 K. V6od, ~ Plant Mahager Sworn and subscribed before me this 22nd day of August, 1994.
8 wLu) W./
Notary Public, State of Ohio My Commission Expires July 28, 1999 t
Docket Number 50-346 '
License Number NPF-3 Serial Number 2231 Enclosure 1 Page 2 The following information is provided to support issuance of the requested change to the Davis-Besse Nuclear Power Station, Unit 1 Operating License Number NPF-3, Appendix A, Technical Specifications )
3.1.1.1, 3.4.5 and Bases 3/4.4.5. '
A. Time Required to Implement: This change is to be implemented l vithin 90 days after the NRC issuance of the License Amendment.
B. Reason for Change (License Amendment Request Number 91-0019 Revision 2): Add a footnote to the MODE applicability section of ,
TS 3.1.1.1 which refers to additional requirements for SHUTD0VN MARGIN vhich are: also being added to TS 3.4.5, Steam Generators; permit the maximum allowable steam generator (SG) water level to be a variable limit based on the plant's mode of operation and the status of the Main Feedvater Pumps and the Steam and Feedvater Rupture Control System (SFRCS), as applicable. These changes vil]
allow the plant to continue to produce full power as future Steam Generator fouling occurs, while ensuring the plant response to accident conditions remains acceptable and adequate margins to safety limits are maintained.
C. Safety Assessment and Significant Hazards Consideration: See Attachment.
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Docket Mmnber 50-346 -I License Ihrnber NPF-3 Serial thunber 2231 Page 1 INFORMATION ONLY Once Through Steam Generator Operation INFORMATION ONLY
Docket Number 50-346 License Number NPF-3 Serial Number 2231
- ;"< 2 INFORMATIC \ 0MLY Table of Contents Page 1.0 Introduction 3 2.0 Steam Generator Description 3 3.0 OTSG Level Indication 3 3.1 Operate Range 4 3.2 Startup Range 4 3.3 Full Range 4 3.4 Correlation Between Ranges 4 4.0 Steam Generator Operation 5 4.1 Plant Heatup 5 4.2 Power Operation 5 4.3 Plant Cooldown 6 Figures
- 1. OTSG Cross Sectional Diagram 7
- 2. Level Instrumentation 8 INFORMATION ONLY
Docket Ntmiber 50-346 License Number NPr-3 Serial Number 2231 g;;;-2 ATl0NCMy 1.0 Introduction The Babcock and Wilcox (B&W) Nuclear Steam Supply System (NSSS) is designed with a unique steam generator, the Once Through steam Generator (OTSG). The purpose of the document is to provide a basic description of the OTSG and explain the fundamentals of its operation.
2.0 Steam Generator Description The OrSG is a vertical counter flow shell and tube heat exchanger with reactor coolant on the tube-side and a secondary boiling mixture on the shell side (See Figures 1 and 2).
On the secondary side, subcooled main feedwater (M N) is ,
distributed through the M N nozzles into the steam filled annulus between the shell and the tube bundle shroud. At the top of the annulus, the M m is heated by direct contact condensation of steam which is aspirated from the tube bundle through the aspirator port in the tube bundle shroud. At 100% power, the aspirating steam is approximately 15% of the total main steam line flow. The downcomer provides the last stage of M m preheating as the M m is heated to the saturation temperature corresponding to the OTSG pressure in the downcomer. Some additional M N heating is also supplied by conductive heat transfer through the tube bundle shroud.
The momentum of the downward directed M N stream and the gravity head of the liquid in the downcomer provide the driving head for the steam generator. This head in the downcomer balances the gravity head of the boiling mixture in the tube bundle and the frictional losses in: (1) the lower downcomer (primarily the i
orifice plate), (2) the tube bundle (primarily at tube support plates), and (3) the aspirator port.
The water in the tube region of an OTSG can be considered to be made up of several zones. At the bottom of the tube bundle, a zone of essentially saturated liquid exists. In the boiling zone of the J OTSG, a steam-water mixture of varying quality exists until a zone of totally saturated steam environment is reached. Above this zone, a region of superheated steam of increasing temperature exists. The length of the boiling zone varies depending on the power level of the reactor and the thermal-hydraulic conditions in the region. Because the length of the boiling zone changes, the length of the superheating zone also varies. This affects the amount of superheat added to the steam before it leaves the OTSG.
3.0 OTSG Level Indication Several " levels" are measured in the steam generator. These level measurements are actually differential pressure (dP) measurements across different physical regions of the OTSG. These dPs have contributions from the mass of water and steam and the flow induced frictional losses between the level taps. The dP contribution from the mass of water and steam is commonly referred to as the l
Docket Number 50-346 g ye License Number NPF-3 Serial Number 2231 l { '"s= . s hh O L sg (p
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Enclosure 2 Page 4 collapsed liquid level. The dP contribution from flow is due to
! the frictional losses, primarily at the orifice plate, the tube l support plates, and the tube surfaces. This flow induced dP varies with the square of the velocity of the fluid in the OrSG, which varies with plant's power level. For instance, at 100% power an indicated dP, such as the startup range discussed below, would have approximately 1/3 of its total contribution from frictional and l
momentum effects and 2/3 from the mass in the tube region.
l Whereas, at hot zero power in MODE 3, the same indicated level l l would result from almost entirely the mass in the tube region. l Therefore the mass in the OTSG for a given dP reading is much greater in MODE 3 than in flODE 1. l l
The various levels measured in the OTSG are discussed below.
3.1 Operate Range The operate range (OR) has a lower tap 102" above the tube sheet in the downcomer (above the orifice plate). The upper tap is in the tube region just above the aspirator port at 394" above the lower !
tube sheet. The OR measures the differential pressure between the l taps and converts this to a percentage of the differential pressure which would exist between the taps if the entire space were filled with saturated water. Therefore, the OR is generally interpreted as a percentage by volume of the water in the downcomer above the lower tap. The OR is temperature compensated for the lower downcomer temperature. It is the only level indication which is temperature compensated. It should be noted that the indicated OR level is higher than actually exists during power operation. This is due to the net effect of both the MEW momentum and the pressure )
losses of the aspirating steam flow through the aspirator port. ]
3.2 Startup Range The startup (SU) range has a lower tap 6" above the lower tube l sheet. The upper tap is the same tap used by the OR. The SU range l indicates the head of the water and steam mass and the frictional losses primarily at the tubes and tube support plates in the tube region below the aspirator port.
The SU range provides an SFRCS low level trip and input to ICS for low level limits.
3.3 Full Range The full range shares a bottom tap with the SU range. The top tap is located 625" above the lower tube sheet. The full range is used when placing the OTSG in wet layup.
3.4 Correlation Between Ranges It should be noted that at zero power (MODE 3) the indicated levels can be correlated relatively easily between the three different indications because the assumption of a collapsed liquid level
l Docket Number 50-346 License Number NPF-3 Serial IAimber 2231 V QQ anlyg Enclosure 2 Page 5 without frictional losses is valid. At high power (high steam flow) conditions any assumed mass in the OTSG can result in a complete I spectrum of indicated levels dependent on the fouling (dP) of the OTSG, which affects the frictional losses, and the condition of the MFW nozzles. Also, since the level tap locations are different and the calibration reference conditions are different, there is not a one-to-one correlation between changes in indicated levels among the three ranges for known changes in OrSG water inventory.
- The above discussion explains why it is difficult to define operating limits, based solely on SU and OR indicated dPs (levels).
4.0 Steam Generator Operation The various methods of operating the OTSG's are described below.
4.1 Plant Heatup As the plant is heated up from MODE 5 to MODE 4 the OTSG's are required to be capable of removing heat from the RCS. In order to accomplish this plus to remove air from the main steam lines, a vacuum is typically established in the main steam system, including the steam generators.
As the RCS continues to heat up from MODE 4 to MODE 3, the OTSG 1evel is reduced. Depending on the chemical content of the OrSG inventory, the generator may be nearly drained, refilled with pure water, and allowed to soak. The draining and refilling process continues until the desired OTSG chemistry is obtained. This also aids in removing contaminates which may have been deposited within the OrSG.
4.2 Power Operation The OTSG 1evel is established at " low level limits" in preparation for changing to MODE 2. This is a level controlled by procedures.
The OTSG level is held constant at this level until the plant's power level is above approximately 28 percent of rated thermal ,
power. This method of level control allows the average RCS temperature (T'" ) to be raised from the zero power value of 532*F to 582'F.
Once T reaches 582'F, the OTSG level is allowed to rise as .
requirell*to maintain T at 582'F. This method of operation continues up to 100 peri:ent rated thermal power or until a OTSG level limit is reached. It is during power operation that the chemical deposition at higher elevations in the OISG occurs. These deposits degrade the thermal-hydraulic characteristics of the OTSG and may eventually cause the plant to become " power limited," i.e.,
the maximum permissible OrSG water level limit may be reached before the reactor is at 100 percent power.
Docket 1Annber 50-346 License Number NPF-3 i Serial Number 2231 l
=;; r 2 INFORMATION, ONLY l
4.3 Plant Cooldown l i
The OTSG level is allowed to decrease to the low level limit as power is reduced. At approximately 28 percent power, the OTSG level is held constant and T is decreased to the zero power !
temperature of 532*F. Once tFe plant is in MODE 3, the crrSG level )
may again be elevated to dissolve as much of the impurities which were deposited during power operation as possible. This process is .
the same as is used during plant heatup to adjust the OTSG !
chemistry.
As the plant cools down, the Steam and Feedwater Rupture Control System (SFRCS) Low Pressure trip is manually bypassed so the cool down can be continued. This disables the plant's primary protection against a Main Steam Line Break or a Main Feedwater Line Break. Consequently, the proposed Technical Specification limits the plant configuration, while allowing for continued OTSG cleaning, so as to ensure that the consequences of a MSLB or MEWLB are not harmful to public health and safety.
As the plant is cooled down to MODE 4, the OTSG level may be raised up to limit the entrance of oxygen into the CyrSG. This reduces the oxidation of the OTSG materials. As steam production ceases the plant coo 160wn is continued using the Decay Heat Removal system.
When the plant enters MODE 5, the OTSG level is adjusted as required to support any planned activities.
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Docket rAutar 50-346 ucense number ter-3 Serial Number 2231 FOR INFORMATION ONLY Enclosure 2 Page 7 Primary inlet -
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Tube (typ) - -
Upper TuNsheet V V
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Shell l
- 3 < Superheating Region Steam Outlet c=>
Aspirator Port c======>
d la Main Feed ' i Inlet - -
>0 m m %
O I l
O o % Film Boiling Ow oO % % 1 Tube Support
\ OO % Departure From Plate (typ) OO Nucleate Boiling
%u oc Downcomer -* oo oO oo Og- -
Nucleate Boiling
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oo O Orifice Plate . . *
- - Saturated Uquid Primary Outlet k
-> Lower Tube Sheet
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M Figure 1: Once Through Steam Generator Cross Sectional Diagram FOR INFORMATION ONLY
or $ $ $ $
- c. . ~ w nc.ie,r - ,.... g e. ;1 g g.
of 105 and SG wowe tempensane of 68F 0Wg Upper Tube Sheet
- C**'*d b * *ac* % **P*' *""* k of 68 and SG womr temperasne of 120F (D "
/\ U Steam "k" Outlet Nonles N g Main Feedwater wmi Header and Nonles " l'w o, aussa
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Lower Tube Sheet M
Figure 2: SG Level Instrumentation
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Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 1 Safety Assessment and Significant Hazards Consideration for License Amendment Request Number 91-0019, Revision 2 Title A proposed change to the Davis-Besse Nuclear Power Station, Unit 1 Operating License, Appendix A, Technical Specification 3.1.1.1, Shutdown Margin; 3.4.5, Steam Generators; and Bases 3/4.4.5, Steam Generators.
Description The purpose of this Safety Assessment and Significant Hazards Consideration is to review proposed changes to the Davis-Besse Nuclear Power Station Unit 1 Operating License Technical Specifications (TS) to ensure the changes do not have an adverse effect on safety and do not involve a significant hazards consideration. The following changes to the TS are proposed:
A footnote is to be added to the MODE applicability section of TS 3.1.1.1.
l This footnote is being added to refer to additional requirements for SHUIDOWN
! MARGIN which are being added to TS 3.4.5, Steam Generators (SG). The additional requirements are being located in TS 3.4.5 rather.than in TS 3.1.1.1, because the conditions which cause the need for additional SHUTDOVN MARGIN are controlled by TS 3.4.5.
Revise TS 3.4.5 to permit the maximum allowable SG level to be a variable limit based on the plant's H0DE of operation. The Operational MODES are defined in Table 1.1 of the TS, A graph of Acceptable SG Operate Range Level '
versus Main Steam Superheat during MODES 1 and 2 is to be incorporated into the TS as Figure 3.4-5. The Limiting Condition for Operation (LCO) vill also specify the maximum acceptable SG 1evel when the plant is in MODE 3 based on the status of the Main Feedvater Pumps and the Steam and Feedvater Rupture Control System (SFRCS) and specify the maximum acceptable SG level when the plant is in MODE 4. A footnote is to be added to require maintaining an adequate SHUTDOVN MARGIN while in MODE 3.
The Bases Section 3/4.4.5 of the TS is to be updated to reflect that the SG vater level limits are consistent with the initial assumptions of the analyses in the Updated Safety Analysis Report (USAR) rather than the Final Safety Analysis Report (FSAR). Examples of incapable Main Feedvater Pumps, as proposed in TS 3.4.5, are also provided in the Bases. The Bases will also note that the shutdown margin requirements are provided in plant operating procedures.
This change to TS 3/4.4.5 is being made to allow the plant to continue to produce full power with continued SG fouling while ensuring the plant response to accident conditions is acceptable. Adequate margins to safety limits will be maintained by this change. Since the SG aspirator ports become flooded at approximately 97 percent Operate Range level, the change also ensures that power operation with flooded aspirator ports is strictly prohibited by always restricting the SG level to 96 percent Operate Range.
Systems, Components, and Activities Affected TheLproposed change affects the maximum allowable SG level as specified in the TS 3.4.5 LCO and the Basis for this LCO in Bases Section 3/4.4.5. The change .
to the TS 3.1.1.1 affects the SHUTDOVN MARGIN requirements for the reactor.
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Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 2 Safety Functions of the Affected Systems, Components, and Activities The safety function of the SGs is to convert the thermal energy of the reactor coolant into steam for use in the turbine generator, to act as a heat sink for the reactor, and to act as a Reactor Coolant System (RCS) pressure boundary.
The requirement of TS 3.1.1.1 ensures that there is sufficient SHUTD0VN MARGIN to make the reactor subcritical from all operating conditions, to ensure the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and to ensure the reactor vill be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition. The SHUTD0VN MARGIN requirements are consistent with the FSAR safety analysis assumptions.
4 The existing LCO for the SGs ensures that the SGs have sufficient inventory to remove heat from the RCS and to limit the amount of inventory to be consistent with the assumptions made in the FSAR.
The Bases Section of the TS provides the technical bases upon which the TS requirements are formed. This ensures that the design bases of the plant are preserved.
Effects On Safety The proposed change to TS 3.1.1.1 serves to reference the SHUTD0VN MARGIN requirements existing under TS 3.4.5. The footnote is being added as a clarification and has no effect on safety.
The proposed change to TS 3.4.5 has been evaluated for its effect on the containment's integrity for accidents inside containment, the effects of accidents on Auxiliary Building environments, the reactivity and core cooling effects for all accidents, and the radiological consequences for all accidents. The proposed MODE 1,2, and 3 limits are more restrictive than the current limit of 348 inches. A discussion of each topic is provided below.
The minimum SG inventory LCO, Action Requirements, and SG inventory-related Surveillance Requirements are to remain the same as those currently found in the existing TS 3/4.4.5.
A. Effects on Safety of LC0 Change
- 1. Containment Integrity The proposed change has no effect on the containment's integrity.
Chapter 6.2, Containment Systems, and Chapter 15.4.4, Steam Line Break, of the USAR, present the analysis of a Main Steam Line Break (MSLB) inside containment. The analysis assumed that the reactor was initially operating at 102 percent rated thermal power. Toledo Edison has evaluated the mass and energy released to the containment for the various SG 1evels permitted by the proposed change. For the levels permitted by proposed Figure 3.4-5 in MODES 1 and 2, the total mass in the faulted SG is less than the 62,500 lbm in the stated USAR analyses l
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Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 3 assumptions. The mass of water in the SG at O'F Superheat is based on the calculated height of a pool of water with no boiling occurring.
The mass of water at higher levels of superheat are based on calculations performed by Babcock and Vilcox (B&V) in support of the development of the B&V Revised Standard TS, Topical Report BAV 2076, issued in April, 1989.
In MODE 3, the SG inventory is limited to 50 inches Startup Range if a Main Feedvater Pump is capable of supplying water to the SG and the SFRCS Lov Pressure Trip is bypassed. This limits the amount of energy available for release to the containment to less than that released during a MSLB at 100 percent Full Power.
When the SFRCS Lov Pressure Trip is protecting the plant or once the possible feedvater flow to the SG is limited to that available from the Motor Driven Feedvater Pump (MDFP), the mass permitted in the SG may be increased until the Operate Range SG level indicates 96 percent.
Toledo Edison has completed analyses which demonstrate that the total mass and energy released to the containment by this mass of water and ten minutes of continued feed from the MDFP (the worst single failure) is less than the mass and energy released by the MSLB analyzed in the USAR. It is concluded that the containment pressure and temperature response. vill be bounded by the USAR results. Therefore, containment integrity will not be more severely challenged.
Since the water in the SG has such a low specific enthalpy when the plant is in MODE , there is no need to limit the SG inventory, with respect to containment integrity. However, a maximum limit is specified to ensure the SG's remain capable of decay heat removal while in MODE 4 by maintaining a steam flow path (e.g., to the Atmospheric Vent Valves).
- 2. Environmental Effects of Breaks Outside Containment Several pipe breaks outside of containment have been evaluated to determine the impact on the environmental qualification profiles for equipment impor 1.h to safety which could be exposed to a harsh environment, c' effects of the proposed change on each break are discussed belt,.
2.1 Main Steam Line Breau In MODES 1 and 2, the water inventory in the SG will be limited to the mass assumed in the MSLB analysis which supports USAR Section 3.6.2.7.1.4, Main Steam to the Turbine Generator. Consequently, the plant response is bounded by the USAR resalts in MODES 1 and 2.
In MODE 3, the SG inventory is limited to less than 50 inches Startup Range Level while a Main Feedvater Pump is operating with the SFRCS Lov Pressure trip bypassed. This limits the amount of energy available for release during MSLB to less than that released during a
Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 4 MSLB from full power. This ensures the environmental conditions are bounded by the values reported in USAR Section 3.6.2.7.1.4.
While the SFRCS Low Pressure trip is protecting the plant or once the plant is'in MODE 3 with no Main Feedvater Pumps feeding the SG's, the proposed LCO vill permit Operate Range levels up to 96 percent.
Analyses completed by Toledo Edison have determined that the mass and energy release from the faulted SG are bounded by the analysis referenced by USAR Section 3.6.2.7.1.4, whe the above conditions are met. Therefore, the environmental effecte ; the MODE 3 MSLB break with elevated SG levels are concluded to se no more severe than the MODE 1 MSLB case.
When the plant is in MODE 4, the water in the SG has a lov specific enthalpy. Consequently, it is concluded that the environmental conditions following a MSLB vould be bounded by the USAR analyses, regardless of the initial inventory of the SG. However, the proposed LCO has an upper limit on SG level in MODE 4 to ensure the SG's remain capable of decay heat removal by maintaining a steam flow path (e.g.,
to the Atmospheric Vent Valves).
2.2 Main Feedvater Line Break The MODE 1 and 2 inventory limits of proposed Figure 3.4-5 vill ensure that the analysis for a Main Feedvater Line Break referenced by USAR Section 3.6.2.7.1.6, Main Feedvater System, is still bounding.
The SG Level is limited to 50 inches Startup Range Level when a Main Feedvater Pump is capable of feeding the SG in MODE 3 and the SFRCS Lov Pressure Trip is bypassed. This limits the amount of energy which would be released during a MODE 3 Main Feedvater Line Break to a value lower than would be released during a Main Feedvater Line Break from full power. This ensures the environmental conditions are bounded by the results reported in USAR Section 3.6.2.7.1.6.
While the SFRCS Low Pressure trip is not bypassed or once the possible feedvater flow to a SG is limited to that available from the MDFP, regardless of the SFRCS status, the SG level is permitted to be as high as 96 percent Operate Range in MODE 3. Toledo Edison has completed calculations which demonstrate that the energy released in the event of a Main Feedvater Line Break with the MODE 3 SG inventory at 96 percent, is less than the energy release discussed in the analysis referenced by the USAR. The calculations assumed that feedvater to the SG, supplied by the HDFP, continued for ten minutes.
This represented the vorst case single failure. Consequently, the peak temperatures and pressures which would occur remain bounded by the existing USAR results.
In MODE 4, the energy content of the SG inventory is very lov.
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Therefore, the effects of a Main Feedvater Line Break are deemed negligible with any inventory in the SG's. Consequently, upper SG inventory limit in MODE 4 is specified only to ensure the SG's remain capable of decay heat removal by maintaining a steam flow path (e.g.,
to the Atmospheric Vent Valves).
Docket Number 50-346 Licensa Number NPF-3 Serial Number 2231 Attachment Page 5 2.3 Main Steam to the Auxiliary Feed Pump Turbines The USAR Section 3.6.2.7.1.5, Main Steam to the Auxiliary Feed Pump Turbines, presents the evaluation of a 2-1/2 inch line and a 6 inch line break in the Main Steam to Auxiliary Feedvater Pump Turbine pipes while at 100 percent power. The referenced analyses for both breaks have assumed that full power operation continues for 10 minutes before plantoperatorstagemanualactiontomitigatetheevent. This would result in 3.5 x 10 lbm of full power temperature and pres 3ure steam being released for the 2 1/2 inch line break and 2.17 x 10 lbm of steam being released from the 6 inch line break during the 10 minute interval with no operator action.
In MODES 1 and 2, the proposed Figure 3.4-5 limits the SG to an inventory less than that assumed in the analyses referenced by USAR Section 3.6.2.7.1.5. This ensures the total blovdown mass and energy are bounded by the USAR reported results.
The SG 1evel is limited to 50 inches Startup Range while the plant is in MODE 3 with either Main Feedvater Pump capable of supplying water to the SG and the SFRCS Low Pressure trip is bypassed. This limits the amount of energy which would be released during this accident to less than vould be released by the same accident starting from full power conditions. This ensures the environmental conditions are bounded by the results reported in USAR Section 3.6.2.7.1.5.
When.the plant is in MODE 3 with the MDFP supplying the SG's or with the SFRCS Low Pressure Trip active, the environmental effects of these breaks are judged to be no more severe than the cases currently presented in the USAR. Vhile the initial mass of water in the SG may be larger than was assumed released in the analysis referenced by USAR Section 3.6.2.7.1.5, the energy content of the steam exiting the break is always. lover at any given time in the transient because the transient'begins in MODE 3 rather than staying at full power for ten minutes. Also, the SG pressure starts falling immediately in the MODE 3 case, whereas the pressure stays constant for 10 minutes in the USAR referenced analysis. This results in lover flow rates out the break for the MODE 3 case.
When the plant is in MODE 4, the energy content of the SG is very lov so that it is judged that the effects of a Main Steam to Auxiliary Feedvater Pumps Line Break are bounded by the USAR referenced case, regardless of the SG inventory in MODE 4. Therefore, an upper limit is specified in MODE 4 only to ensure the SG's remain capable of decay heat removal by maintaining a steam flow path (e.g., to the AtmospF ric Vent Valves).
2.4 Steam Generator Blovdown System Break.
USAR Section 3.6.2.7.1.15, Steam Generator Blowdown System, presents an evaluation of the effects of a Steam Generator Blovdown Line Break in MODE 1. The analysis referenced by the USAR assumed 30 minutes of full power operation occur prior to plant operators taking action to mitigate the break. The extended period of continued power operation i
Docket Number-50 346 License Number NPF-3 Serial Number 2231 Attachment Page 6 was assumed because it vould be difficult for the control room operators to diagnose this small break. This is because the Main Feedvater Pumps can provide sufficient feedvater flow to compensate for the break, so that the SG 1evel and pregsure vould not be affected. This vould result in over 3 x 10 lbm of normal SG operating temperature water being discharged out the break.
The proposed MODE 1 and 2 limits of Figure 3.4-5 ensure that the SG inventory assumption made in the analysis referenced by the USAR is met.
In MODE 3, with a Main Feedvater Pump capable of supplying water to the SG's and the SFRCS Low Pressure trip bypassed, the SG inventory is limited to less than 50 inches Startup Range level. This limits the energy which could be released to a value below the full power condition. This ensures that the resulting environmental conditions are bounded by the results reported in the USAR Section 3.6.2.7.1.15.
If the plant is in MODE 3, with no Main Feedvater Pump capable of supplying water to the SG's or with a Main Feedvater Pump running with the SFRCS Low Pressure trip active, the SG inventory will be permitted to be as high as 96 percent on the Operate Range instrument. This condition is judged to be bounded by the USAR analyses because of the same effects discussed in the section on the Main Steam to Auxiliary Feedvater Pump Turbine line break.
Also, because break flow exceeds MDFP capacity (initially), the SG level and pressure vill decrease and alert operators of the problem prior to the 30 minutes of continued blevdown which was postulated for the MODE 1 analysis. This vould result in faster operator response to mitigate the accident.
If the Main Feedvater (MFV) pumps are supplying feedvater to the SG with the SFRCS Low Pressure trip active, the SG vill eventually be isolated by the SFRCS. This vould occur since there is very reduced heat input from the RCS in MODE 3, so that the SG pressure would decrease. This also helps limit the environmental effects of the accident.
When the plant is in MODE 4, the specific enthalpy of the SG inventory is very lov. Consequently, it is concluded that the effects of a Steam Generator Blowdown Line Break in MODE 4 with any SG inventory would be bounded by the USAR analyses. Therefore, the proposed upper SG inventory limit in MODE 4 is only to ensure the SG's remain capable of decay heat removal by maintaining a steam flow path (e.g, to the Atmospheric Vent Valves).
i 3.0 Reactivity and Core Cooling Effects A MSLB tesults in a rapid overcooling of the RCS, which adds positive I reactivity to the reactor due to the negative moderator temperature !
coefficient. In order to prevent the reactor from becoming critical, )
adequate SHUTDOVN MARGIN must be maintained. Toledo Edison has i evaluated the RCS cooldown associated with a MSLB while in MODE 3 with l
l
Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 7 the SG inventory at 96 percent on the Operate Range and the MDFP supplying feedvater to the SG and the SFRCS Low Pressure Trip bypassed. This plant condition bounds all other MODE 3 scenarios except the case of the Main Feedvater Pumps supplying the SG's with the SFRCS Low Pressure trip bypassed, which has also been evaluated for cooldown effects.
The cooldown calculations have included conservative assumptions which result in overestimating the RCS cooldown. It was assumed that the RCS cooled down to the feedvater temperature and that no decay heat from the reactor is added to the RCS, the vorst case single failure occurs which results in continued feeding of the SG for ten minutes, and all of the faulted SG inventory and feedvater flashes to steam due to heat transferred from the RCS.
Administrative control requirements, as well as a specific footnote in TS 3.4.5, vill ensure that there is adequate SHUTD0VN MARGIN to prevent the reactor from becoming critical during any MODE 3 MSLB.
The footnote in TS 3.1.1.1 serves to reference the additional SHUTD0VN MARGIN requirements in TS 3.4.5. The administrative controls include determining the boron concentration required to compensate for the calculated cooldown and procedural requirements to establish the necessart boron concentration in the RCS prior to raising the SG level above the low level limits. These controls ensure that the acceptance criteria of USAR Section 15.4.4, Steam Line Break, are met for MSLBs in MODE 3 with elevated SG levels.
When the plant is in MODE 4, the SG's can only induce a very limited cooldown of the RCS following any secondary line breaks. Therefore, no reactivity requirements beyond the TS definition of MODE 4 are necessary. In addition, no specific Feedvater Pump requirements are needed for the same reason. The maximum SG inventory limit is provided to ensure the SG's remain capable of decay heat removal while in MODE 4 by maintaining a steam flow path (e.g, to the Atmospheric Vent Valves).
The USAR Chapter 15 discussion of Steam Line Breaks evaluated the potential for departure from nucleate boiling. Vith respect to the l Departure from Nucleate Boiling Ratio (DNBR), the results of the analyses reported in USAR Section 15.4.4.2.3, Results of Analysis, remain valid. Under the proposed change, the mass of water in the SG's vill remain consistent with the USAR assumptions in MODES 1 and
- 2. When the plant is in MODES 3 or 4, the reactor's heat flux is so lov that departure from nucleate boiling cannot occur even if the RCS pressure was reduced to saturation, so the amount of secondary inventory in the SG has no effect on the DNBR. Consequently, the proposed change has no effect on keeping the reactor fuel adequately cooled.
Docket Number 50-346 License Number NPF-3 Serial Number 2231 i Attachment l l
Page 8 4.0 Radiological Consequences Of the accidents analyzed in USAR Chapter 15, Accident Analysis, only two have radiological consequences which are potentially affected by i an increased SG inventory. These are a Steam Generator Tube Rupture (SGTR) and a Steam Line Break. Each is evaluated below.
i 4.1 Steam Generator Tube Rupture The consequences of a SGTR are presented in USAR Section 15.4.2, Steam Generator Tube Rupture. This analysis assumed that after the reactor trips, all the fission products contained in the RCS inventory which transfers to the secondary side of the steam generator are directly released to the environment until the RCS has been depressurized below the lovest Main Steam Safety Valve (MSSV) lift pressure. An increased inventory in the SG does not affect the time required to depressurize the RCS to this pressure. This is because the amount of time required to reduce the RCS pressure to below the lowest MSSV setpoint only depends on the initial conditions of the RCS (which are at vorst case conditions in the USAR analysis) and the energy removal rate of the secondary side of the SG, which is not affected by this proposed change. Therefore, the total mass and radioactive contamination released are independent of.the initial SG inventory. Consequently the proposed increased inventory does not affect the results presented in the USAR.
4.2 Steam Line Break The radiological consequences of a Steam Line Break are presented in USAR Section 15.4.4, Main Steam Line Break. The USAR states that for breaks of pipes smaller than the Main Steam pipe, the consequences are bounded by the Main Steam Line Break (MSLB) analysis. This remains true, since the MSLB releases the entire inventory of the SG. Smaller line breaks may not release the entire inventory.
The assumptions used in the USAR radiological evaluation include a 1 gpm tube leak in the faulted SG and that all the activity in the SG inventory, the feedvater, and leaked RCS inventory are released to the environment. Toledo Edison has calculated the radiological consequences of a MSLB vith a SG inventory of 96 percent Operate Range level. The results are presented in Table 1. While the thyroid doses are higher than the analysis presented in the USAR, they are clearly below the NRC acceptance criteria included in the Davis-Besse Operating License Safety Evaluation Report, NUREG-0136, Section 15.3.
Therefore, it is concluded that the higher values do not represent a significant increase in the consequences of the accident. The primary reason for the increased thyroid dose is that the new calculation assumed that the SG vas stagnant for two hours, with a 1 gpm RCS tube leak, prior to the break occurring. This reflects the desired method of removing chemical impurities and deposits from the SG's. The calculated whole body doses have decreased from the values reported in the USAR due to changes in the dose factors for the analyzed isotopes.
Docket Number 50-346 ;
License Number NPF-3 l Serial Number 2231 ]
Attachment Page 9 B. Effects on Safety of Basis Change The proposed change to the Bases Section 3/4.4.5 revises the text of the Bases to show that the design basis of the level requirements is in the USAR. The original assumptions of the FSAR are included in the USAR, so that there is no loss of information regarding the permissible SG vater levels. Examples of incapable Main Feedvater Pumps are also proposed in this revised Bases text and have no adverse effects on safety. The Bases will also state that the adequate SHUTDOWN MARGIN required in the Technical Specification vill be provided by plant operating procedures.
This statement has no adverse effect on safety.
C. Conclusion of Effects on Safety Based on the above discussion, it is concluded that the proposed changes to TS 3.1.1.1, 3.4.5, and Bases 3/4.4.5 do not have an adverse effect on safety.
SIGNIFICANT HAZARDS CONSIDERATION The Nuclear Regulatory Commission has provided standards in 10CFR50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would: (1) Not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) Not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Not involve a significant reduction in a margin of safety. Toledo Edison had reviewed the proposed change and determined that a significant hazards consideration does not exist because operation of the Davis-Besse Nuclear Power Station, Unit 1 in accordance with this change would:
la. Not involve a significant increase in the probability of an accident previously evaluated because the inventory contained in the Steam Generator dSes not affect the probability of experiencing any accident initiator.
Ib. Not involve a significant increase in the consequences of an accident previously evaluated because the consequences of the proposed changes have been determined to be within the acceptance criteria for previously evaluated accident analyses.
2a. Not create the possibility of a new kind of accident from any accident previously evaluated because no new failure modes are being introduced, and therefore no new accident scenarios can be postulated.
2b. Not create the possibility of a different kind of accident from any accident previously evaluated because no new failure modes are being introduced, and therefore no different accident scenarios can be postulated.
- 3. Not involve a significant reduction in a margin of safety since the original accident analyses acceptance criteria are still met.
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Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 10 Conclusion Based on the above, Toledo Edison has determined that this License Amendment Request has no adverse effect on safety and does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specification that must be reviewed by the Nuclear Regulatory Commission, this License Ameridment Request does not constitute an unreviewed safety question.
Attachment Attached is the proposed marked-up change to the Operating License.
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Docket Number 50-346 License Number NPF-3 Serial Number 2231 Attachment Page 11 Table 1:
Offsite Dose Consequences of a Main Steam Line Break Event Plant Condition / Location Thyroid Dose Whole Body Dose Documentation Time (REM) (REM)
MSLB MODE 3, Elevated Site Boundary 0.951 0.003 (New Inventory / 2 Hr.
Limiting Toledo Edison Event) Calculation LPZ/30 Day 0.063 0.0002 MSLB MODE 1, 100% Site Boundary 0.79 0.007 (Current Power /USAR 2 Hr.
Limiting Section 15.4.4 Event)
LPZ/30 Day 0.041 0.0003 i
MSLB MODE 1, 100% Power Site Boundary <1.0 <1.0 (NRC SER 15.3 Analysis 2 Hr.
Accept-ance LPS/30 Day -- --
Criteria) ,
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