ML20114D902
| ML20114D902 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/03/1992 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20114D898 | List: |
| References | |
| NUDOCS 9209100144 | |
| Download: ML20114D902 (21) | |
Text
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l Docket Number 50-346 'i
, License Number NPF-3 1
, Set;ial Number 2050 l
. Enclosure 1
,Page 1 i
APPLICATION FOR AMENDMEliT TO F/tCILITY OPERATING LICENSE NUMBER NPP-3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 4
Attached are the requested changes to tha Davis-Besse Nuclear Pover Station, Unit Number 1 Facility Operatirg License Number NPP-3. Also included is the Safety Assessment and Significant Hazards
, Consideration.
The proposed changes (submitted under cover letter Serial Number 2050) concern:
Technical Specifications Section 3.4.5 (Steam Generators)
Technical Specifications Bases 3/4.4.5 (Steam Generators) i
'N
[ By: l uM/
i D. C. Shelton, Vice President,-
Nuclear - Davis-Besse i
Svorn and subscribed before me this 3rd day of September, 1992.
l
}A,afin)) / p))
4 Notary S6blic, St' ate of Ohio EVELYNL DMSS ICTN1YPUBUC,STATECf CHO
% b%f.xpanMy23,1994 i
9209100144 920903 PDR i P ADOCK 05000346 PDR J
i Docket Nuwber 50-346
. L2 cense Number NPF-3
. Serial Number 2050 a
F,nclosure 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 3.4.5 and Bases 3/4.4.5.
A. Time Required to implement: This change is to be implemented within 90 days after the NF; issuance of the License Amendment.
B. Reason for Change (License Amendment Request Number 91 0019):
Permit the maximum allovable steam generator (SG) 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. The change vill allow the plant.to continue to produce full power as Steam Generator fouling occurs, while ensuring the plant response to accident conditions remains acceptable and adequate margins to safety limits are maintained.
C. Safety Jasessment and Significant flazards Consideration: See Attachment.
b w.mi e . . . . . . . . . . . m __.. _ _ ,.
Docket Number 50-346
, l.icense Number NPF-3
. Serial Number 2050
. Enc'losure 2
,Page 1 FOR INFORMATION ONLY I
l once Through Steam Generator. Operation l
FOR INFORMATION ONLY l
i
) Docket Number 50-346 !
) . License Number NPF-3 Ser.ia1 Number 2030 FOR INFORMATION ONLY Enclosure 2
,Page 2 Table of Contents
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i
[ ate ,
1 1.0 Introduction 3 i
! 2.0 Steam Generator Description 3 j
. 3.0 OTSG Level Indication 4 3.1 Operate Range 4 3.2 Startup Range 4 3.3 Full Range 4 344 Correlation Between Ranges 4 4.0 Steam Generator Operation 5 ,
4.1 Plant lleatup 5 4.2 Power Operation 5 4.3 Plant Coolefovn 5 Figures
- 1. OTSG Cross Sectional Diagram 7
- 2. Level Ins rumentation s 8 FOR INFORMATION ONLV
i Docket Number 50-346 License Number NPF-3 Serial Number 2050 enaosure 2 Page 3 FOR INFORMATION ONLY
. 1.0 Introduction The Babcock and Vilcox (B6V) 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 descript on of the OTSG and explain the fundamentals of its i
operation.
2.0 Steam Generator Description The OTSG is a vertical counter flow shell and tube heat er wager with reactor coolant on the tube-side and a secondary boil ..g mixture on the shell side (See Figures 1 and 2).
On the secondary side, subcooled main feedvater (HFV) is distributed through the MFV nozzles into the steam filled annulus between the shell and the tube bundle shroud. At the top of the annulus, the MFV 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 flov. The downtomer provides the last stage of HFV preheating as the HFV is heated to the saturation temperature corresponding to the OTSG pressure in the downcomer. Some additional MFV heating is also supplied by conductive heat transfer through the tube bundle shroud.
Th t. momernum of the downward directed MFV stream and the gra ity het , of the liquid in the downcomer provide the driving head for the ateam generator. This head in the downcomer balances the gravity head of the boiling mixture in the tube bundle and the frictional losses ins (1) the lover downcomer (primarily the 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 bc :.;nidered to be made up of several zones. At the bottom of the u bundle, a zone of essentially saturated liquid exists. In the boiling zone of the OTSG, a steam-vater mi>.ture 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 l contributions from the mass of v o . and steam and the flow l FOR INFORMATION ONLY
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Docket Number 50-346 License Number NPF-3 Serial Number 2050 2
Enclosure 2 FOR INFORMATION ONLY Page 4 induced frictional losses between the level taps. The dP contribution from the mass of vater and steam is commonly refttred to as the collapsed liquid level. The dP contribution from flov is due to the frictional losses, primarily at the orifice plate, the tube support plates, and the tube surfaces. This flov induced dP varies eith the square of the velocity of the fluid in the OTSG, which varies with plant's power level. For instance, at 100% power an indicated dP, such as the startup range discussed '
belov, vould have approximately 1/3 of its total contribution from frictional and momentum effects and 2/3 from the mass in the tube region. Vhereas, at hot zero power in Hode 3, the same indicated level vould result from almost entirely the mass in the tube region. Therefore the mass in the OTSG for a given dP reading is much greater in Mode 3 than in Hode 1.
The various le uels measured in the OTSG are discussed belov.
3.1 Operate Range The operate range (OR) has a lover 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 lover :ube sheet. The OR measures the diffetential pressure between the taps and converts this to a percentage of the differential pressure which would exist between the taps if the entire space vere filled with saturated water. Therefore, the OR is generally interpreted as a percentage by volume of the veter in the downcomer above the lover 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 HPV momentum and the pressure losses of the aspirating steam flov through the aspirator port.
3.2 Startup Range The startup (SU) range has a lover tap 6" above the lover tube sheet. The upper tap is the same te? used by the OR. The SU range 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 lov 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 lover tube sheet. The full tange is used when '. acing the OTSG in vet layup.
FOR INFORMATION ONLY
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Docket Number 50-3%
, License Number NPF-3 Mu%*!" 2 5 FOR INFORMATION ONLY
,Page 5 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 without frictional losses is valid. At high power (high stern flov) conditions any assumed mass in the OTSG can result in a co. >1ete spectrum of indicated levels dependent on the fouling (dP) of the OTSG, which affects the frictional losses, and the condition of the MFV 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 OTSG vater inventory.
The above discussion explains why it is difficult to define {
operating limits, based solely on SU and OR indicated dPs i (levels).
4.0 Steam Generator pe 9a ,
The various methocs d e ' m ar.g ths OTSG's are described below.
4.1 Plant Ileatup As the plant is heated up from H0DE $ 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 H0DE ', to MODE 3, the OTSG 1evel is reduced. Depending on the chemical content of the OTSG inventory, the generator may be near)y drained, refilled with pure water, and allowed to soak. The draining and refilling process continues until the desired 0TSG chemistry is obtained. This also '
aids in removing contaminates which may have been deposited within the OTSG.
4.2 Power Operation The OTSG level is established at " low level limits" in preparation for changing to MODE 2. This is a level controlled by procedures.
The OTSG 1evel is held constant-at this level until the plant's pover level is above approximately 28 percent of rated thermal power. This method of level control allows the average RCS-
- **) to be raised from the zero power value of 532'F temperature (T to 582*F.
FOR INFORMATION ONLY
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Docket Number 50-346
, License Number NPT-3
!*su""E" 2 5 FOR INFORMATION ONLY
,Page 6 Once T reaches 582'F, the OTSG 1evel is allowed to rise as requir$Pto maintain T at 582'F. This method of operation continues up to 100 pe9cInt rated thermal power or until a OTS::
level limit is reached. It is during power operation that the chemical deposition at higher elevations in the OTSG occurs.
These deposits degrade the thermal-hydraulic characteristics of the OTSG and may eventually cause the plant to become "pover limited," 1.e., the maximum permissible OTSG vater level limit may be reached before the reactor is at 100 percent power.
4.3 Plant Cooldovn The OTSG 1evel is allowed to decrease to the lov level. limit as power is reduced. At approximately 28 percent power, the OTSG i level is held constant and T is decreased to the zero power l.
temperature of 532'F. Onceib! plant is in H0DE 3, the OTSG 1evel l may again be elevated to dissolve as much of the impurities which vere 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 dovn, the Steam and Feedvater Rupture Control System (SFRCS) Lov 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 Feedvater 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 HSLB or MPVLB are not harmful to public health and safety.
As the plant is cooled down to MODE 4, the OTSG 1evel may be raised up to limit the entrance of oxygen into the OTSG. This reduces the oxidation of the OTSG materials. As steam production ceases the plant cooldovn is continued using the Decay Heat Removal system. When the plant enters H0DE 5, the OTSG 1evel is adjusted as required to support any planned activities.
FOR INFORMATION ONLY
Docket Number 50-346
. License Number NPP-3 seriei Numuer 20so Euclosure 2 FOR INFORMATION ONLY Page 7 Primary inlet --
Tube (typ) - -
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Umer Tubesheet
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Shell ----> cro Superheating Region Steam Outlet c='*
Aspitator Port
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Tube Suppen 00 % ospanure From Plate (typ) \ 00 Nucleate Ming
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Orifce Plate #. .
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- - Saturated Uquid Primary Outlet k.,- -> Lower Tube Sheet
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AdN Figure 1: Once Through Steam Generator Cross Sectional Diagram FOR INFORMATION ONLY 9
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et se and SG weier sampmanne et s20F 6
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Stearn Outlet i m in Fee & veaar Nozzles ;
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2,1,: 2, ~ Figure 2: SG Level Instrumentation :
-32 3 U Z- Zk 11 0 3 wh e-e in to "O C st! O I M Ume4 G U U k U to Oe & C r4 O J LA LJ L -
Dock.et Number 50-346 License Number NPF-3
. Serial Number 2050
. Attachment Page 1
- SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS rCNSIDERATION FOR LICENSE AMENDHENT REQUEST NUMBER 91-0019 Title A proposed change to the Davis-Besse Nuclear Power Station. Unit 1 Operating License, Appendix A, Technical Specification 3.4.5, steam Generators, and Bases 3/4.4.5, Steam Generators.
Description 5
The purpose of this Safety Assessment and Sign!ficant Hazards Considerat.on is to review a proposed change to the Davis-Desse Nuclear Power Station Unit 1 Operating License Technical Specifications to ensure the change does not have an adverse effect on safety and does not involve a significant hazrrds consideration. The following-change to the Technical Specifications is proposed:
Revise Technical Specification (T.S.) 3.4.5 to permit-the maximum-allovable Steam Generator (SG) level to be a variable limit based on the plant's Mode of operation. The Operational Nodes are cefined in Table 1.1 of the Technical Specifications. A graph of Acceptable SG Operate Range Level versus Main Steam Superheat-during Modes 1 and 2 is to be incorporated into the Technical Specifications as Figure 3.4-5.
The Limiting Condition for Operation vill also specify the maximum acceptable Steam Generator level when the plant is in Mode 3 based on the status of the Hain Feedvater Pumps and the Steam and Feedvater Rupture Control System (SFRCS) and specify the maximum acceptable Steam Generator level when the plant is in Mode 4.
The Bases Section 3/4.4.5 of the Technical Specifications is to be updated to reflect that the SG vater level limits cre consistent with the initial assumptions =of the analyses in the Updated Safety Analysis Report (USAR) rather than the Final Safety-Analysas. Report (FSAR).
Examples of incapable Main Feedvater Pumps under this proposed T.S.
l 3.4.5 are also provided in the Bases.
This change to T.S. 3/4.4.5 is being.made to allov-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 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 The proposed change affects the maximum allovable SG 1evel asLspecified in the T.S. 3.4.5 Limiting Condition for Operation (LCO)~and the Basis for this LCO in Bases Section 3/4.4.5.
Docket Number 50-346 I License Number NFF-3 {
, Serial Number 2050 Attachment Fage 2 l
' Safety Functions of the Affected Systems, Components, and Activities The safety function of the Steam Generators is to convert the thermal !
energy of the reactor coolant into steam for use in the turbinc !
generator, to act as a heat sink for the reactor, and to act as a RCS pressure boundary. }
The existing Limiting Condition for Operation for the Steam Generators !
ensures that the Steam Generaters have sufficient inventory to remove heat from the Reactor Coolant Systen (RCS) and to_ limit the amount of inventory to be consistent with the assumptions made in the Final Safety Analysis Report (FSAR).
The Bases Section of the Technical Specifications-provides the technical bases upon'which the Technical Specification requirements are formed.- This ensures that the design bases of the' plant is preserved.
Effects On Safety i
The proposed change 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 area is provided below.
The minimum SG inventory Limiting Condition for Operation,-
Action requirements, and SG inventory related Surveillance Requirements are to remain the same as those currently found in the existing Technical Specification 3/4.4.5.
A. Effects on-Safety of LCO Chaage '
1.- Containment Integrity i i 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 vas initially operating at 102 percent rated thermal power. Toledo. '
Edison has evaluated the mass and energy ~ released to'the containment for the various SG levels 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~1ess'than the 62',500 lbm in the stated-USAR analyses 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 in support.of the development of the B&V Revised Standard Technical. Specification, Topical. Report BAV 2076, issued in April, 1989.
Docket Number 50-346 License Number NPF-3
. Serial Number 2050 Attachment Page 3 In Mode 3, the SG inventory is limited to 30 inches Startup Range if a Main Feedvater Pump is capable of supplying vater to the SG j and the SFRCS Lav Pressure Trip-is bypassed. This limits the amount of energy available for release to the containment to less 3 1
than that released during a MSLB at 100% full power.
, When the SFRCS Lov Pressure Trip is protecting the plant or once '
the possible feedvater flov to the SG is limited to that available from the Motor Driven Feedvater Pump (HDFP), the mass to be permitted in the SG may be increased until the Operate Range SG level indicates 96 percent. Toledo Edison has completed analyses l
vhich demonstrate that the total mass and energy released to the containment by this mass of water and ten minutes of continued feed from the HDFP-(the vorst single failure) is less than the mass and ,
energy released by the HSLB analyzed in the USAR. cit is concluded that the containment pressure and temperature response vill be i
bounded by the USAR results. Therefore, containment integrity vill not be more severely challenged.-
Since the water in the SG has such a_lov specific enthalpy when the plant is in Mode 4, there is no need to limit the SG inventory, with respect to containment integrity reasons. 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 r
Several pipe breaks outside of containment have been evaluated to determine the_lmpact on the environmental qualification profiles for equipment important to safety which could be exposed to a harsh
, environment. The effects of the proposed change on each break is discussed below.
2.1 Main Steam Line Break i In Modes 1 and 2, the water inventory in the SG will-be limited to the mass assumed in the Main Steam Line Break (HSLB) analysis which supports USAR Section 3.6.2,7.1.4, _ Main Steam to the Turbinr Generator. _ Consequently, the plant response is bounded by the USAR
-results in Modes 1 and 2.
-In Mode 3, the SG inventory ic 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 a HSLB to less than that released during a MSLB_from full power. This ensures _that the environmental conditions are bounded by the values reported in USAR Section 3.6.2.7.1.4.
While the SFRCS Lov Pressure trip'is protecting the plant or once the plant is in-Hode 3 with no Main Feedvater Pumps-feeding the SG's, the proposed Limiting Condition for Operation vill permit Operate Range. levels upLto 96 percent.- Analyses completed by
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Docket Number 50-346 License Number NPF-3 Serial Number 2050 At.tachment Page 4 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, when the above conditions are met. Therefore, the environmental effects of the Mode 3 MSLB break with elevated SG levels are concluded to be 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 Limiting Condition For Operation has.an upper limit on SG level in Mode 4 to ensure the SG's remain capable of decay heat removal by reaintaining 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, are still met.
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 lover than vould 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.
Vhile the SFRCS Lov Pressure trip is not bypassed or once the possible feedvater f)3v to a SG-is limited to that available from the MDFP, regardless af 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 Steam Generator 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 MDFP, 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.
Therefore,-the effects of a-Main Feed Water _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:
~
i remain capable of decay heat removal-by maintaining a steam flow path (e.g., to the Atmospheric Vent Valves).
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Docket Number 50-346 License Number NPF-3 Serial Number 2050
' 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 pover. The referenced ar lyses for both breaks have assumed that full power operation contirmes for 10 minutes before plant operators take mgnual action to mitigate the event. This would result in 3.5 x 10 lbm of full power temperature and pressure $ 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 vith 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 is bounded by the USAR reported results.
The SG 1evel is limited to 50 inches-Startup Range while the plant-is in Mode 3 vith either Hain Feedvater Pump capable-of_ supplying vater to the SG and the SFRCS Lov 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 Pover conditions. This ensures that the environmental conditions are bounded by the results reported in USAR Section 3.6.1.7.1.5.
Vhen the plant is in Mode 3 vith the HDFP supplying-the SG's or with the SFRCS Lov Pressure Trip active, the environmental effects of these breaks are judged to be no more severe _ than the cases currently presented in the.USAR. While'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 lower 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 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 SGLis very lov so that it is judged that the ef fects 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 Atmospheric Vent Valves).
2.4 Steam Generator Blowdown System Break USAR Section 3.6.2.7.1.15, Steam Generator Blovdown System, presents =an evaluation of the effects of a Steam Generator Blovdovn=
Line' Break'in Mode 1. The analysis referenced by the USAR assumed-
, Docket 14 umber 50-346 '
License Number NPF-3 i
. Serial Number 2050 Atischment f Page 6 l
30 minutes of full power operation occur prior to plant operators .;
taking action to mitigate the break. The extended period of '
continued power operation was assumed because it vould be difficult i 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 anjpressurewouldnotbeaffected. This vould result in over 3 x l 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 i inventory assumption made in the analysis referenced by the USAR is met. 1 In Mode 3, with a-Main Feedvater Pump capable of supplying water to the SG's and the SFRCS Lov Pressure trip bypassed. the SG .
inventory is limited to less than.50 inches Startup Range level, j This limits the energy which could be released to a value below the !
full power condition. This ensures that_the resulting environmental conditiens are bounded by the results reported in ,
USAR Section 3.6.2.7.1.15. '
If the plant is in Mode 3, with no Main Feedvater Pump capable of supplying vater to the SG's or with a Main Feedvater Pump running ;
with the SFRCS Low Pressure trip. active, the SG inventory vill 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 flov exceeds MDFP capacity (initially), the SG i level and pressure vill decrease and alert operators of the problem 1 prior-to the 30-minutes of-continued blowdown which vas postulated-for the Mode 1 analysis. This would result in faster operator response.to mitigate the accident. If the Main Feedvater (HFW) pumps are supplying feedvater to the SG with the SFRCS Lov Pressure trip active, the SG vill-quickly 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'vould decrease rapidly. !
When the plant is in.Hode 4, the specific enthalpy of: the SG inventory is very lov. Consequent 1.y, it is concluded that the -
effects of a Steam Generator 31ovdown Line Break:in Mode 4 vith any_ '
SG inventory would be bounded by the USAR_ analyses. -Theref ore', : 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).
3.0 Reactivity and Core Cooling Effects '
A HSLB results in a rapid overcooling of the-RCS, which~ adds positive reactivity to the reactor due to the negative temperature'-
coefficient. In order te, prevent the reactor from becoming critical, adequate shutdown margin must be maintained.-. Toledo-1
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- Docket Number 50-346 License Number NPF-3
. Serial Number 2050 At.tachment Page 7 Edison has evaluated the RCS cooldown associated with a HSLB vhile in Mode 3 vith the SG inventory at 96 percent on the Operate Range- ,
and the Motor Driven Feedvater Pump supplying feedvater to the SG and the SFRCS Lov 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 bfRCS Lov Pressure trip bypassed. The conditions permitted by the proposed Technical Specification for that plant condition have also been evaluated for cooldown effects.
The cooldovn calculations have included conservative assumptions which result in overestimating the RCS cooldovn. It was assumed I that the RCS cooled down to the feedvater temperature or that no I decay heat from the reactor is added to the RCS, the worst case single failure occurs which results in continued feeding of the S0 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 vill be established to ensure that there is adequate shutdown margin to prevent the reactor from:
becoming-critical during any Mode 3 HSLB. The administrative controls include determining-the boron concentration requir4 to compensate for the calculated cooldown and procedural. requirements i to establish the necessary boron concentration in the RCS prior to '
raising the SG 1evel above the lov level limits. These controls-ensure that the acceptance criteria of USAR Section 15.4.4, steam Line Break, are met for HSLBs.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 Technical Specification 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).
Vith respect to the 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 vater 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 amotint!of secondary. inventory in.the SG has no ef net on the DNBR. Consequently,.the proposed change has no.
effect on keeping the reactor fuel adequately cooled. .
4.0 Radiological Consequences-Of the accidents analyzed in USAR Chapter 15, Accident Analysis,
, only-tvo have' radiological-consequences which are potentially.
affected by an-increased SG inventory. These are a~ Steam Generator
~ Tube Rupture and-a Steam Line Break. Each is evaluated below.
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Docket Number 50-346
- Licensa Number NPF-3 i . Sorial Number 2050 l ,
Attachment i -
Page 8 i .
1 4.1 Steam Generator Tube Rupture l The consequences of a Steam Generator Tube Rupture (SGTR) are presented in USAR Section 15.4.2, Steam Generator Tube Rupture.
1 This analysis assumed that all the fission products contained in j the RCS inventory which transfers to the secondary side of the j steam generator are directly released to the environment until the i RCS has been depressurized below the lovest Main Steam Safety Valve
! (HSSV) lift pressure. An increased inventory in the steam i
generator does not affect the ti.ae required to depressurize the RCS
! to this pressure. This is because the amount of time required to
{ reduce the RCS pressure to below the lovest HSSV setpoint only
! depends on the initial conditions of the RCS (which are at vorst t case conditions in the USAR analysis) and the energy removal rate
{-
of the secondary side of the steam generator, which is not affected by this proposed change. Therefore, the total mass and radioactive-
_ contamination released are independent of the initial SG inventory.
j 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 i that for breaks of pipes smaller than the Hain Steam pipe, the ,
consequences ate bounded by the Main Steam Line Break (HSLB) analysis. This remains true, since the HSLB releases the entire inventory of the SG. -Smaller line breaks may not release the !
entire inventory.
4 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 with a SG inventory of 96 percent Operate Range level. The results are presented in Table 1.- Vhile the ;
thyroid doses are higher than the analysis presented in the USAR, i 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 vnlues 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- s decreased from the values reported in'the USAR due to changes in '
the dose factors for the analyred isotopes. . '
B. Effects on Safety of Bases Ch'ange 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 or;iginal assumptions of the FSAR
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7 Docket Number 50-346-License Number NPF-3 ;
. Serial Number 2050 l At.tachment Page 9 are included in the USAR, so that there is no loss of information regarding the permissible SG vater levtic. Examples of incapable Main Feedvater Pumps are also proposed in this revised Bases text and have no adverse effects on safety.
C. Conclusion of Effects on Safety
{
Based on the above discussion, it is concluded that the proposed- ,
change to T.S. 3/4.4.5 and its Bases does not have an adverse effect on safety.
i 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 voulds (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 th; ' avis-Besse Nuclear Power Station, Unit 1 in accordance with this change vouldt-la. Not involve a significant increase in the probability _of an accident previously evaluated because the inventory contained in the Steam Generator does 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 change have been determined to be within the acceptance criteria for previously evaluated accident analyses.
2a. Not create the possibility of a nev_ kind of accident from any accident previously evaluated because no new failure modes are being introduced, and therefore no nev accident scenarios can be postulated.
2b. Not create the-possibility of a different kind of accident from any accident previously.evalrm ed because no new failure modes are being introduced, and therefore no d!fferent-accident scenarios can be postulated.
- 3. Not involve a significant reduction'in a margin of safety since the original accident analysis acceptance criteria are still met.
, . _ ~ _ _ . _ . - _ _ _ . _ _ _ . _ _ _ _ _ - , _ . , _ _ . . . , , _ _ , . , - . _ , _ _ _ - . -
Docket Number 50-346 License Number NPF-3
. Serial Number 2050 Attachment Page 10
' Conclusion i
, Bated on the above, Toledo Edison has determined that this License Amendment Request has no adverse effect on safety and does not involve i a significant hazards consideration. As this License Amendment Request '
concerns a proposed change to the Technical Specifications that must be l cevieved by the Nuclear Regulatory Commission, this License Amenoment '
Request does not constitute an unreviewed safety question. j Attachment i
.> -)
Attached is the proposed marked-up change to the Operating License, i
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4 I: r i
i mem-,-,a~e.--,w e ws -
,- Docket Number 50-346 License Number NPP-3
, Serial Number 2050 At.tachment Page 11 Table la i Offsite Dose Consequences of a Main Steam Line Break l l
Event Plant Condition / Location Thyroid Dose Vhole Body Dose ;
Documentation Time (FEM) - (REM) ,
I MELB Mode 3, Elevated Site Boundary 0.951 0.00.3 (New Inventory / 2 Hr.
Limiting Toledo Edison Event) Calculation LPZ/30 Day 0.063' O.0002 MSLB Hode 1, 100% Site Boundary 0.79 0,007 l (Current Pover/USAR 2 Hr.
Limiting Section 15.4.4 ,
Event)
LPZ/30 Day 0.041 0.0003 MSLB Hode 1, 100% Power Site Boundary <1.0 <1.0 (NRC SER 15.3 Analysis 2 Hr.
Accept-ance LPZ/30 Day --- - --
Criteria) 4., - 4.v.,- ,,m.- . , . , , ,.,,..,.,,.,,.r.,.m,_ _ _ . . . . _ . , . . . , , - , , - , , . . . . . - 4 .r,.. ....,,..__,.....,,_.m-