ML20082A223
| ML20082A223 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/28/1991 |
| From: | Schuster T COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20082A225 | List: |
| References | |
| NUDOCS 9107100231 | |
| Download: ML20082A223 (26) | |
Text
- _ _ _ _ _ _ _
/ \ 'C:mmonwealth Edisen h (
O ) 1400 opus Place Downers Grove, Illinois 60$15 June 28,1991 Dr. Thomas E. Murley, Director Office Of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn: Document Control Desk
Subject:
Byron Station Unit 2 Braidwood Station Unit 2 Application for Amendment to Facility Operating Licenses NPF-66 & NPF-77
.Aopendix A, Technical Specifications Model D-5 Steam Generator Level Setpoint Changes NRCRochetNosJ0-4553nd_50-457 l
l
Dear Dr. Murley:
{
Pursuant to 10 CFR 50.90 Commonwealth Edison Company (CECO) proposes to amend Appendix A, Technical Specifications of Facility Operating Licenses NPF-66 and NPF-77 ior Byron and Braidwood Stations respectively. The proposed amendment revises a portion of the Technical Specification Tables 2.2-1 and 3.3-4, Reactor Trip System instrumentation Trip Setpomts and En ineered Safety Features Actuation System Instrumentation Trip Setpoints respectiv . New setpoints are specified for the Low-Low Steam Generator Level-Reactor rip / Auxiliary Feedwater Initiation and the High-High Steam Generator Level-Turbine Trip /Feedwater Isolation for the Unit 2 Model D-5 Steam Generators. These changes are necessitated by a 3roposed modification to relocate the lower sensing tap of the Unit 2 Steam Generator Level instrumentation to improve the exhibited level indication performance during secondary system transients and low power operations.
The description and safety analysis of the proposed changes are presented in Attachment A. The revised Technical Specification pages are contained in Attachment B.
9107100231 910628 FDR ADOCK 0500o455 p PDR
/ sci:1019:1 W 000085 ii
s Dr. Thomas E. Murley June 28,1991 The proposed changes have been reviewed and approved by both on-site and off-site review in accordance with Commonwealth Edison procedures.
Commonwealth Edison has reviewed this proposed amendment in accordance with 10 CFR 50.92(c) and has determined that no significant hazards consideration exists.
This evaluation is documented in Attachment C. An Environmental Assessment has been completed and is contained in Attachment D. The Commonwealth Edison Setpoint Study, relative to the proposed Unit 2 Steam Generator level setpoints, has been completed. Any impact on the proposed setpoints has been factored into this response. Please also note that an Amendment of similar scope has been previously submitted and approved for Vogtle and Catawba Nuclear Power Stations.
Commonwealth Edison is notifying the State of Illinois of our application for this amendment by transmitting a copy of this letter and its attachments to the designated State Official.
To the best of my knowledge and belief the statements contained here are
~
true and correct. In some respects, these statements are not based on my personal knowledge but upon information received from other Commonwealth Edison and contractor employees. Such information has been reviewed in accordance with Company practice and I believe it to be reliable.
Please direct any questions you may have concerning this matter to this Respectfully, state of d : county of s
Terence .S ster
$gebre me b/ on tNs * #M day Nuclear Licensing A'dministrator Notary Public /
y ___
Attachments: (A): Descriptbn and Safety Analysis of the Proposed Changes B): Proposed Technical Specification Changes C): Evaluation of Significant Hazards Consideration (D) Environmental Assessment Statement cc: W. Kropp - Resident inspector, Byron (w/o attachments)
S. Dupont - Resident inspector, Braidwood (w/o attachments)
A. Hsia - Project Manager (Byron), NRR R. Pulsifer - Project Manager (Braidwood), NRR _ _ , , _
W. Shafer - Branch Chief, Region lil .. g g g i e m 3g g :
Office of Nuclear Facility Safety -IDNS snsosa c t- a em f a y c ..-
I?/f C0YY i f. Eis M i . %
k~~--<--~---~
t AHACHMENT_A RESCRIP_TKMAND.SAFEILANALYSIS OEfBOEOSED_CBANGES 1.0 Descdotion.sLitteAoposed_ Change This proposed change will change values for Unit 2 steam generator (S/G) low-low and high-high level instrumentation in Tables 2.2-1 and 3.3-4.- The low low setpoint will change from 17% of the old instrument span to 36.3% of the new instrument span, and the high-high setpoint will change from 78.1% of the old instrument s aan to 80.8% of the new Instrument span. This represents an increase in t 1e operating margin between the low-low setpoint and the nominal operating level of about 22 inches, and an increase in the operating margin between the high high setpoint and the nominal operating level of about 4 inches.
For the Unit 2 S/G high-high level setpoint, new values will replace the current values for Total Allowance, Z, Sensor Error, Trip Setpoint, and Allowable Value.
For the Unit 2 S/G low-low level setpoint, new values will replace the current values for Trip Setpoint and Allowable Value. The current values for Total Allowance, Z, and Sensor Error will be replaced with N.A.
2.0 Current ReoulremBat The current design of Unit 2 model "D-5" steam generators for the Byron /Braidwood Nuclear Power Stations provides ta as for these instruments at elevations 566" and 438". Therefore, the associated nstruments can only measure water levels in the 128" span between the two taps. The low-low signal is used to initiate a reactor trip signal and to initiate the motor driven and diesel driven auxiliary feedwater pumps. The high-high signalinitiates a turbine trip and feedwater isolation. The Technical Specidcations express these setpoints as a aercentage of instrument span. The narrow span associated with the tap locations
- n conjunction with allowances for instrument error and uncertainty results in a narrow range of allowable level operation.
i 3.0 Begueate.d_BeYision L The steam generators at the Byron /Braldwood Units are model"D" type steam generators. Each Unit 1 has model"D-4" while each Unit 2 has model"D-5".
These steam generators are similar in that they are both preheat steam generators. However, the Narrow Range Span (NRS) currently used for control and protection functions for each of the steam generators is different. Specifically, the lower tap location of the NRS is located above the downcomer region on the D-5 steam generators and below the downcomer region on the D-4 steam generators. This results in significantly different operating characteristics due to the difference in recirculation flow rate (velocity head) at these locations. Figure 1 Illustrates the current level tap locations for the model D-4 and model D-5 steam
_ generators, along with the proposed change to the lower tap on D-5.
/scl:1019:3
4 3.0 Bequested Bevision (con't)
The velocity head error on the model D-4 steam generators is 2.6% of narrow range span. This error is accounted for in the selection of the setpoints. For the model D S steam generators, this error was calculated to be as high as 10% of span at the equivalent elevation. Locating the level tap above the transition cone eliminated the measurement error introduced by the velocity head, but as a consequence the narrow range span is reduced. Relocation of the level sensing line to the same elevation as the model D 4 steam generators requires that the velocity head error (which in effect masks the actual level changes in such a manner as to compensate for " shrink" and " swell" effects) be accounted for lie the determination of trip and operating level setpoints. It has been determined that this can be accomplished without reducing any operating margin to trip.
Velocity head error is a result of flow from the downcomer past the lower tap.
This flow decreases the pressure seen at the lower tap and, as a consequence, results in an increase in the difference in 3ressure seen by the t,ansmitter. When the pressure in the lower tap is comparec to the filled reference leg, the maximum differential pressure is present at 0% span and the minimum differential pressure is present at 100% span. Therefore, any increase in differential pressure in response to a higher flowrate past the lower tap results in an indication of lower than actual water level.
Steam generators exhibit a " shrink" or " swell" characteristic at low power levels (when feedwater temperature is low) when feedwater flow is initially increased or decreased, respectively. This makes plant control difficult and more susceptible to trips. The " shrink" and " swell" in the Model D-4 is much less pronounced due to the location of the lower tap in the downcomer region. At the time that the level
" swells", the recirculation flow rate increases thus creating a lower pressure at the lower instrument tap and significantly reducing the magnitude of the indicated level increase.
This difference in level indication response between units creates concerns for the operators when the feedwater control system is in manual control. A variation in the main feedwater flow changes the actual water level in the steam generators. However, the indicated level as a iesult of the actual water level change initially differs In the two steam generator types. This is a direct result of the locat;on of the lower tap. In the past, changes in the steam generator water level associated with normal plant operation have resulted in unnecessary trip signals. In order to reduce the possibility of future trips, Commonwealth Edison has elected to extend the span of the levelinstruments by lowering the lower tap from 438" to 333". This allows a wider operating margin to both the low-low setpoint and the high-high setpoint. By moving the current D-5 lower tap to a location consistent with the D-4 steam generators, the two steam generators will behave similarly. Specifically, a change in the feedwater flow will produce the same indicated level effect in both types of steam generators.
/scl:1019:4
4 3.0 Requested Revision (con't)
Additionally, following a reactor trip from power, the steam generators experience a significant " shrink" charactoristic. Emor00ncy procedures require that the operator establish indicated level in the safety grade narrow range instrumentation prior to throttling auxiliary foodwater flow. Due to the wider span on the model D-4 steam generators, if Indicated level does fall off scale it is generally recovered very quickly. In the case of the D-5 steam generators, more time is required for the indicated aarrow range level to recover. This increases the potential of post trip over cooling and depressurization.
The determination of the revised sotpoints; including now values for Total Allowanco (TA), parameters not measured on a ponoJic basis (Z), Sensor Error (SE), and Allowable Value;is consistent with the methodology outlined in WCAP 12523, " Bases Document for Westinghouse Setpoint Methodology for Protection Systems" and WCAP-12583, " Westinghouse Setpoint Methodology for Protection Systems". The new values for the U2 S/G hl0 i high lovel sotpoint are provided in Attachment B. For the U2 S/G low-low level sotpoint, a minimum amount of excess margin has been incorporated. As a result, use of equation 2.2-1 (Z +RE + SE <; TA) in Action Statomonts for Technical Specification 2.2,
" Limiting Safety System Settings" and 3/4.3.2, " Engineered Safety Features Actuation System Instrumentation", is not applicable. Therefore, to ensure that the aparopriate action statements are entered, the values in columns TA, Z and SE wil: be marked N.A.
In summary, relocating the narrow rango instrumentation lower sensing tap on the Westinghousa model D-S steam generators to the same elevation as the model D-4 steam generators will provido the following enhancomonts:
- 1. The offects of " shrink" and " swell" at low power levels will be greatly reduced, thus reducing the potential for reactor trips.
- 2. The operating margin between nominal level and the low-low level trip will be Increased, thus reducing the potential for unnecessary reactor trips at power.
- 3. A wider instrument span will result in more benign indication of steam generator level transients.
- 4. LevelIndication of steam generator transients on Unit 1 and Unit 2 will be the same, thus el;minating the need for unit-specific training and operating practices and reducing the risk of inappropriate operator manual response.
- 5. The time necessary to recover indicated level following a reactor trip will be greatly reduced. thus reducing the potential for an over cooling due to excessive auxiliary feedwater.
1
/scl:1019:5
t EIGVBEJ i
/scl:1019:6
e FIGURE 1 MODEL D4 MODEL D5 STEAM GENERATOR STEAM GENERATOR Steam Outtet Steam Ownet To Terbine To Tm.h CE5IG4 MCD1MED
( % A.=..e.:.,:e',-
s'-
~
C ' ^ *; = '
! = , - ~
- "'I Menwer g '
566 Upper Tac Prim,.ary ,
se,e.N,- {
Me .
ei. if L-Upper Tap 566 g fMe.e.
Primary e
,,,. : . tevei rr Trie 5x *-"' '""" s 8 h O O b e<eie, nominal Level SC2 522 hl-w! - tevel TT Tr g 462 montaal te et Lo-to Level Tric 460 - " ~ " "
..,, to-er Tap a3a O C5 Tw of - wr*88'"
1 .
[
ta) Jao-(s) 360- % -
,W (2) 333- **"*"
[N l
a.et.vn ,ess 333 tower Tap
,-,...... ,7.-.-e
/ ' .
1 Top of tubesheet -
[ h"es*e( [ [ Pte *(
j 2- Lower marrou range tap 3 Top of streipt section of downctuer Pre heet Asse
"' Pre-hose Ares
! Feed.ater =-
eedwetee
$. Upper narrow revge top /
W "-" '
7 W l
Te,be Sheet ,_, _
f oteTSheet (1) 0* ,e. --
g- j-, Too of Tubesheet 0 i F Menwer R$enway p*eelant 't t Cootent anset b
r
4.0 ImpatLoBroposed_ Change By moving the current Unit 2 steam generator lower tap to a location consistent with the Unit 1 steam generator lower tap elevation, the two steam generators will behave similarly. Specifically, a change in the feedwater flow will produce the same indicated level effect in both types of steam generators. This would allow for consistent operator response for both Byron /Braidwood Units, thereby enhancing their response to steam generator transient conditions.
-The proposed narrow range configuration provides more operating margin and thereby reduces the challenges to the safety systems. Table 1 identifies the current narrow range span configuration and the revised configuration. The revised setpoints are appropriately compensated for the effects of volocity head.
.The velocity head was actually measured on a model D 5 steam generator tyalcal of the Byron steam generators. The velocity head is a function of power leve but it is not directly proportional. That is, there is no velocity head effect at 0%
power; there is 20 inches of velocity head error at 100% power. The maximum possible velocity head error is 23 inches. This means that the actual wner level can be as much as 23 inches h!gher than the indicated water level. The nominal lovel setpoint was selected such that at 100% power the actual water level in the steam generator would be the same before and after installation of the modification. Therefore, since the velocity head e ror at 100% power is 20 inches, the nominal level is set at 482" (482" + 20" - 502"). The low-low level setpoint was reduced by 42 inches from the existing value due to lowering of the lower level tap and the Incoiporation of a new safety analysis limit. The high-high level setpoint was reduced by 16 inches due to the incorporation of a new safety analysis limit in conjunction with the expected velocity head effect.
Velocity head e*fects result in an indicated leve! less than or eq ual to actual level for any power level, therefore it is conservative to use indicatecl level for the low-low level set aolnt. The safety analysis limit setpoint for the high-high steam generator water evel turbine trip /feedwater isolation function is assumed to be at the upper tap elevation (566") and the Technical Specification setpoint is adjusted to account for the velocity head.
The low-low and high-high steam generator water level setpoints were selected based on a statistical analyses of the errors associated with the particular l Instrument loop. The statistical analyses are consistent with the methodology l outlined in WCAP-12583, "Westinghouso Setpoint Methodology for Protection
! Systems" and WCAP-12523, " Bases Documer t for Westinghouse Setpoint Methodology for Protection Systems." Refer to Table 2 for a summary of the i results of the statistical analyses. Note that the errors considered when I determining a setpoint are unique to their function, i.e. velocity head is considered when determining the high-high steam generator water level setpoint, but it is not l considered when determining the low-low level setpoint for conservatism; the low-low level setpoint considered for Feedline Break incorporates an l environmental allowance due to the ensuing adverse environment, whereas the Loss of Normal Feedwater does not, and so on.
i
/scl:1019:7
- - . .. . -. -. . -- ~ ~.
4.0 ImpactoLPJQPOSed_Cha0ge (con't)
Certal_n accident analyses in the UFSAR rely on the low low level reactor trip (e.g.
Feedline Break, Loss of Normal Feedwater, and Steamline Break), certain analyses rely on the high-high level trip (e.g. Feedwater System Malfunction
'Causin nomh. glevel, an increase in Feedwater (e.g. Secondary SystemFlow), Pipe Ruptures and certain analyses inside assume Containment, Steaman initial System Piping Fa lure, Feedwater System Pipe Break, Steamline Break Outside Containment. Feedwater System Malfunction, etc.). Determination of the low-low and high-high level setpoints is discussed above. With consideration to those ,
analyses which assume an initial nominal level (mass), the velocity head was accounted for depending on the power level at which the anal sis was performed. At 0% power there is no compensation for velocit head; 100%
power,20 inches of volocity head error are accounted for, ditionally, for certain analyses which assume an initial nominal level, a typical +5% NRS control error is applied for conservatism to account for any deviations from nominal. For example, 5% NRS is conservatively applied to the initial nominal level in the Feedline Break Analysis (less water avaliable in the intact steam generators at the time of trip to be used for long-term heat removal). - Note that control system errors are increased because of an increase in NRS.
The current non-LOCA safety analyses aresented in the Byron /Braidwood-UFSAR are all based on the level setpo nts and aerformance of the model D 4 steam generators. As will be discussed below,11ere are several transient .
analyses that will not require new analyses because the current D 4 data bounds the new D-5 level assumptions. Note that the same water level (actual) in both the D-4 and the D 5 steam generators does not necessarily resuit in the same mass inventory (which forms the basis of the input values used in the safety analyses) due to the internals of each generator. See Figure 1.
-The following evaluation was aerformed by Westinghouse and was reviewed and ap3 roved by Commonwealth Edison, it addresses the impact on the UFSAR sa"ety analyses (LOCA, non-LOCA, SGTR).
4.1 UCRDsingRasla The criteria stated in the accident analysis chapters of the Byron /Braidwood UFSAR were used to establish the continued applicability of the licensing basis safety analyses by demonstrating that the conclusions In the UFSAR remain
!= valid. The folloiwng evaluation was performed to provide the bases for the
' determination that the proposed modification to the steam generator level tap
- does not involve an unreviewed safety question.
L L
1
/scl:1019:8
- - . , , . . - . - . , - - . - - , . . . - . ~ . - - . . - - - . - . - , . . - . - - . _ - , - - - . - . . . , -
i i
4.2 Evaluation 4.2.1 Non:LO.CA Accidents The following safety evaluation has been prepared to justify the relocation of the lower tap on the Byron /Braidwood Unit 2 D 5 steam generators to a location consistent with the Byron /Braldwood Unit 1 D-4 steam generators. The safety analysis assumptions affected by the steam generator modification include the initial steam generator water level, the steam generator watec level low-low reactor trip, and the steam Oenerator water level high-high turbine trip /feodwater isolation function. The following sections provide discussions for each of the UFSAR events.
4.2.1.1 Transients Not. Requiring Reanalysis The following transients were not reanalyzed because they are not affected by the changes to the safety analysis assumptions, 4.2.1.1.1 Mass and Energy. Release for Postulated Secondary System Pipe RupturesJnside_ Containment (presented.in.UESAR Section 02,1.4)
The steamline break mass / energy releases are generated to ensure that the peak containment pressure and temperature limits are not exceeded. The mass / energy release data is arimarily dependent upon secondary side parameters such as the brea < size, the initial steam generator inventory, steam pressure, auxiliary feedwater flow, and feedwater temperature. An increase in the steam generator inventory will result in more limiting mass / energy data. The mass / energy release data currently presented in the UFSAR is based upon the current D-4 steam generator level program. The revised D-5 steam generator nominal lowel will remain bounded by the current mass & energy releases calculations. Thus, no analysis is required and the current mass / energy release data presented in the UFSAR remains bounding, 4.2.1.1.2 Excesslye jncreasejo .Socondary_SteamfjowJprese nted jn_U ES AB SectionJ 5,1,3)
For this ANS Condition ll ovent, cases are analyzed at beginning and end of life conditions both with and without automatic rod control Since the steam Generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not aff ected by the steam generator modification. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
/scl:1019:9
4
- 4.2.1.1.3 Inadvertent Opening ota Steam GeneratotBelletotSafety_ Valve (prese nted.in_UESAB_Sectionl5J A)
This ANS Condition ll event is bounded by " Steam System Piping Failure"
. (15.1.5). The DNB design basis is met and the conclusions of the UFSAR remain valid.
4.2.1.1.4 Steant. System _Eip!ag _Eailu re_(pres a ntedj a_UES AB_SectiorL1515)
For this ANS Condition IV event, the ANS Condition ll criterion of meeting the DNBR limit is applied. The analyses are performed at zero power and the results are primarily dependent upon secondary side parameters such as the break size, the initial steam generator inventory, steam pressure, auxiliary feedwater flow, and feedwater temperature. An increase in the steam generator inventory will result in a higher return to power, thereby maximizing the core heat flux and minimizing the DNBR Since the initial steam generator inventory assumed for the current analysis of the Steam System Piping Failure is larger than the inventory associated with the revised indicated level, the results will not be adversely affected. Therefore, no analysis is required and the conclusions of the UFSAR remain valid. Note that there is no velocity head effect at 0% power.
4.2.1.1.5 Losa oLExternaLLoad.(presentedJtLUESAB_Sectionl5,2.2)
This ANS Condition 11 event is bounded by " Turbine Trip"(15.2.3). The DNB design basis is met and the conclusions of the UFSAR remain valid.
4.2.1,1,6 Iurbinelrip_(presentedla_UESAB_Sectionl5,213)
For this ANS Condition ll event, cases are analyzed at beginning and end of life conditions both with and without pressurizer control. For the four cases analyzed, the reactor trip signals were generated by either the high pressurizer pressure or over temperature AT reactor trip functions. Thus, a change to the low-low steam generator reactor trip setpoint will not affect this transient. The transient is also not that sensitive to the initial steam generator level and the change in the nominal program level will not adversely im aact this event given the low-low level trip is not required. Therefore, no analysis s required and the conclusions of the UFSAR remain valid.
4.2.1.1.7 Inady.erLClosu rn_oLM ain SteantlsolationYalves_(pres entedJnil ESAB Section_1fi,214)
This ANS Condition ll event is bounded by " Turbine Trip" (15.2.3). The DNB design basis is met and the conclusions of the UFSAR remain valid.
/scl:1019:10
4 4.2.1.1.8 EeedwateLSysicaLEipafkeaMaresentedja_UES AASecti.onl528)
For this ANS Condition IV event, two cases, with and without offsite power available, are analyzed to demonstrate that the core remains intact and in a coolable geometry. This is accomplished by showing that no hot leg boiling occurs pnor to auxiliary feedwater heat removal turnaround. The evaluation performed to address the changes in the D-5 steam generators concluded that, when compared to the D 4 steam generator massos used in the current UFSAR analysis, the initial masses would decrease by no more than 2% in both the faulted and the intact steam generators.
For the faulted loop, this would result in reaching the reactor trip setpoint (on low-low level) earlier in the transient which is a benefit. However, since the intact steam generators are also at a lower initial water level, there could be less mass available in the intact steam generators at the time of trip to be used for long-term heat removal.
The evaluation also concluded that the mass value corresponding to the low-low steam generator level trip setpoint (including adverse environmental errors) would increase by more than 7% mass (from the current D-4 setpoint to the new D-5 setpoint). With the increase in the low-low level setpoint, a reactor trip signal will actually be generated earlier in the transient when compared to the current UFSAR analysis, and consequently, more mass will be available in the intact steam generators to remove decay heat. Based on engineering judgement, the increase in the low-low level trip setpoint will more than compensate for the changes in the initial levels.
Thus, when the effects on the initial levels in both the intact and faulted steam generators are considered in combination with the higher reactor trip setpoint, a new analysis would remain bounded by the analysis currently presented in the UFSAR. Therefore, no new analysis is required and the conclusions presented in the UFSAR remain valid.
4.2.1.1.9 EartiaLLoss_oLEorcedEeactoLGoolanLE!awlptesented lo_UES AB .Section 1511)
For this ANS Condition 11 event, the transient is terminated by a low RCS loop flow reactor trip. Since the steam generator low-low water level reactor tria and high-high water level turbine trip /feedwater isolation functions are not chal enged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification.
Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
/scl:1019:11
u O
4.2.1.1,10 Complele 1.oss_oLEorced3eactoLCoolantfjow_(presentedlnllESAB SectionJ5,3,2)
- For this ANS Condition ll event, the transient is terminated by an undervoltage or underfrequency reactor trip. Since the steam generator low-low water level reactor tnp and high-high water level turbine inp/feedwater isolation functions are not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by the steam generator modification. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
4.2.1.1.11_ Beactot Coolantf umplocked3olor/ShafLBlaaklpresentedJo_UESAB S_ectionl5A3)
For this ANS Condition IV event, the criteria include showing that peak design pressures are not exceeded and that the cladding at the " Hot Spot" in the core remains intact.- Since the steam generator low-low water level reactor trip and high-high water level turbine trip /feedwater isolation functions are not challenged -
in this event and the steam generator mass remains relatively constant throughout the event,- it is not affected by the steam generator modification.
Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
4.2.1.1.12 Loche.dRotor - Bods-inMB
_ This svent is analyzed to determine the percentage of fuel rods that experience DNB. Since the steam generator low-low water level reactor trip and high high water level turbine trip /feedwater isolation functions are not challenged In this event and the steam generator mass remains relatively constant throughout the event, it la not affected by the steam generator modification. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
L 4.2.1.1.13 Unconttotled RCCA Bank _ Withdrawal.1tonta_SubcriticaLoLLoxEowar Slattup_ Condition (presante_djn_UFSAR Section_15A1)
For this ANS Condition ll event, the analysis is performed at zero power conditions. A _ rapid reactivity addiilon resuits from the withdrawal of a bank of-rodsi Because of the fast nature of this event, the secondary side is not
-modeled. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
l L / sci:1019:12
4.2.1.1.14 Uncontrolled.BCDA.BanlLWilbdtawaLat.Eowet(presentedjn_UESAB -
- Sectionl5A 2)
For this ANS Condition ll event, various power levels and reactivity insertion raws for both minimum and maximum reactivity feedback are analyzed. The transients are terminated by an Overtemperature AT or High Neutron Flux reactor trip.
Since the stearn generator low-low water level reactor trip and high high water
- level turbine trip /feodwater isolation functions are not challenged in this event and the steam generator mass remains relatively corntant througlout the event, it is not affected by the steam generator modification. Therefore,- no analysis is i required and the conclusions of the UFSAR remain valid.
l 4.2.1.1.15 RCCA.M!sonetation.1pte.sentedJo_UESAB_SectionJ5A3) i i
For the events presented In the section of the UFSAR, the DNBR criterion is i applied. Since the steam generator low-low water level reactor trip and high-high .I' water level turbine trip /feedwater Isolation functions are not challenged in t ais event and the steam generator mass remains relatively constant throughout the event (or the secondary side is not modeled), they are not affected by the steam generator modification. Therefore, no analysis is required and the conclusions of ,
the UFSAR remain valid.
4.2.1.1.16 Startup of an Inactive ReactotCoolant Pump at anJacorrect Temperature (ptesentedla_UESAR.Sectionl5AA)
For this ANS Condition ll event, a calculation is performed to determine the time
- at which the appropriate shutdown margin is lost and the core becomes critical.
This calculation does not model the secondary side, and thus, is not affected by the steam generator modification. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
- 4.2.1.1.171 CYCS_MalfunctionthatBesults.la a.DecreaseJn.the. Boron Concentration in.tbe_HeactoLCoolanL(pLe.sented in UFSAFLSection 15A6)
This ANS Condition ll event is analyzed to show that adequate time exists for operator action to terminate a dilution event prior to a loss of shutdown margin.
Any changes to the secondary side setpoints have no impact on the determination of the time available for operator action. With respect to the DNBR criterion, the at-power cases, Modes 1 and 2, are bounded by the " Uncontrolled RCCA Bank Withdrawal at Power" (15.4.2).. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
/scl:1019:13
4.2.1.1.18 Specitum_QLilCCAEjectionAccidenta_(presented'in UFSAR S9ction
. 16 4 Bl
- This is a Condition IV event'and is analyzed to show that the cladding _ _
- tem 3erature at the " Hot Spot" in the core remains below 2700*F and that the fuel-melt is less than 10%._ A rapid reactivity addition results from the ejection of an .
RCCA. Because of the fast nature of this event, the secondary side is not -
modeled. Therefore, no ana_ lysis is required and the conclusions of the UFSAR -
remain valid.
-- 4.2.1.1.19 . ' InadvertenLOperation of Emergency _ Core._ Cooling _Syslent(ECCS) Ruring --
Eoner_OperatiorL(plelentedjnllFSMLSection_15,5.1) m For this ANS Condition ll event, the transient is initiated by a spurious safety injection signal. The injection of borated water drives the nuclear power and RCS temperature down. Since the steam generator low-low water level reactor trip .
and high-high water level turbine trip /feedwater isolation functions are not
' challenged in this event and the steam generator mass remains relatively ."
- constant throughout the event, it is not affected by the steam generator-modification. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
4.2.1.1.20 loadvertent.Opealog.otAEtessurizeLSately_orEelleLValy_e (ptesentedJn
. UESABEectionl511)
For this ANS Condition ll event, the transient is terminated by an
- Overtemperature AT reactor trip. Since the steam generator low-low water level raactor trip and high high water level turbine trip /feedwater isolation functions are
.._not challenged in this event and the steam generator mass remains relatively constant throughout the event, it is not affected by tne steam generator.-
modification. Therefore, no analysis is required and the conclusions of the UFSAR remain valid.
4.2.1.1.21 SuperheaterLSieamline BreakMassjkEnergyEeleaseDutskte
- Containment For the steamline breaks occurring outside containment, the analyses were performed at 100% and 70% power conditions, consistent with the approach-presented in Reference 1. The results are pr'marily dependent upon secondary side parameters such as the break size, initia. steam flow, initial steam generator inventory, steam pressure, protection system setpoints, and auxiliary feedwater -
flowrates.
L
/ sci:1019:14
. . - _ _ _ _ _ . _ _ ~ _ . . . - _ _ . _ . _ . _ _ ...._ - . _ . _ __ _ . _._ _ .,-. _
l 4.2,1.1.21 Superheated Steamline. Break Mass & Energy. Release Outside Containment (con't)
The tap modification to the D-5 steam generators will yield initial masses slightly lower than those considered in the current analyses. Given that the higher original masses will act to delay low low steam generator level reactor trip and the ESF actuation required for mitigation, the new initial masses will result in releases no worse than those in the current calculations. There will be no change in the equivalent mass used in the analyses to represent the lowslow steam generator water level setpoint as a result of the tap modification. Therefore, the results of the Superheated Steamline Break Mass and Energy Releases presented in WCAP-10961-P remain valid and no new analyses are required.
4.2.1.2- TransientsBequirJngBeanalysis The following transients were analyzed since they are adversely affected by changes in the low-low steam generator water level reactor trip, high-high steam generator water level turbine /feedwater isolation function, and Initial water level.
4.2.1.2.1 EeedwatetSystentMalfunctiolCausing anlacreaseJn.Eeedwater_Elow (presented.irtVESAR Section 15,L2)
This ANS Condition ll event is analyzed to demonstrate that the DNB design basis is satisfied. This is ensured by showing that the minimum DNBR value never decreases below the appropriate safety analysis limit value. Changes to the initial steam generator water level as well as the high-high steam generator water level turbine trip /feedwater isolation function setpoint result in potentially more limiting conditions, and thus, a new set of analyses must be performed to demonstrate that the acceptance criteria are met and the conclusions presented in the UFSAR are still valid.
l Cases analyzed are both full power and zero power conditions. The analyses have been performed consistent with the cases presented in the UFSAR with the y exceptions identified below.
l l a. For the feedwater control valve accident at full power, one feedwater valvo is assumed to malfunction resulting in a step increase to 129.5% of nominal feedwater flow to one steam generator. Feedwater flow to the remaining steam generators is assumed to be constant. Feedwater temperature decreases from 440 F to 402.8 F.
- b. For the feedwater control valve accident at zero power, one feedwater I valve is assumed to malfunction resulting in a step increase from zero to 58.4% of nominal feedwater flow to one steam generator. The flow to the remaining steam generators is zero. The feedwater temperature assumed was 35 F.
l l
I
/scl:1019:15
I For the zero power analysis,it has been shown that the maximum reactivity insertion rate is less than the insertion rate analyzed in the " Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Condition" (UFSAR 15.4.1).
For the full power analysis, the steam generator level in the f aulted loop reaches the high high level setpoint and all the feedwater isolation valves are automatically closed and the main feedwater pumps are tripped. A turbine trip and a reactor trip are also generated at that time. The transient results show that the minimum DNBR value is above the safety analysis limit value at all times, thus DNB does not occur at any time during the event. Therefore, the change in the D-5 tap location is acceptable.
4.2.1.2.2 Eeedwater_Syste m_ Malf unctions _C auting a B eductionJrLEeedw ate r Temperature.(presented in UESAR SectiorL15.L1)
This ANS Condition 11 event is boundad by "Feodwater System Malfunctions that Result in an increase in Feedwater Flow" (15.1.2). The DNB design basis is met and the conclusions of the UFSAR remain valid.
4.2.1.2.3 Loss _of Non-EmergencyAC Power to_theflant Auxillarles/ Loss _of Normal EaedwaterflawJpresentedJJLUESAB_ Sections _1510L151Z)
These ANS Condition ll events are analyzed to demonstrate that adequate heat removal capability exists to remove core decay heat and stored energy following reactor trip. This is ensured by showing that there is no overpressurization of the primary or secondary side and that pressurizer filling does not occur. A reduction in the steam generator low-low level NRS setpoint, due to the D-5 tap relocation, could potentially minimize the amount of mass available following reactor trip to remove the core decay heat and stored energy, resulting in a more severe transient. Thus, a new analysis was performed to demonstrate that all of the acceptance criteria are satisfied.
An auxiliary feedwater flow of 612 gpm from one auxiliary feedwater pump delivered to four steam generators at a temperature of 125 F was assumed. To maximize pressurizer filling, the 3ressurizer power-operated relief valves and oressurizer spray were assumec operable. The steam generator low-low water evel was assumed to be at 400 inches above the tube sheet in addition, a 10%
steam generator tube plugging level was assumed. The method of analysis and assumptions used were otherwise in accordance with those presented in the UFSAR.
/scl:1019:16
1 o
4.2.l.2.3 Loss.otNon EmergencyAGEower toJhe.ElantAuxillaries/ Loss of Normal EeodwaterfowlpresentedlfLUESAR.Sectionsl516/15,Pm7) (con't) :
The translent results show that the capacity of the auxiliary feedwater system is adequate to provide sufficient heat removal from the RCS following reactor trip.
The criterion that the pressurizer does r'ot fillis met, thus assuring that the integrity of the core is not adversely affected. For the case without power, the
- results verify the natural circulation capacity of the RCS to provide sufficient heat removal capability to prevent fuel or clad damage following reactor coolant pump coastdown. Thus, ihe change in the D 5 tap location is acceptable.
4.2.1.3 Non LQCA.Gorduslan" Based upon the analyses and evaluations presented, the Byron /Braidwood Unit 2 D 5 steam generator modification resulting 10 the relocation of the lower narrow range tap can be made without violatlag any 4 'he conclusions of the UFSAR.
This includes a tull power indicated steam y m ator water level of 482 inches, a bl0h high water level of 522 inches and , low.sw water level of 418 inches.
4.2.2 Steam.GeneratorlubsBupturelvaluation The steam generator tube rupture (SGTR) analysis presented in the Byron /Braidwood UFSAR was perfo ned to evaluate the radiological consequences resulting from a SGTR accident. The major factors that affect the resultant offsite radiation doses are the amount of fuel defects (level of reactor coolant contamination ), the primary to secondary mass transfer through the ruotured tube, and the steam released from the ruptured steam generator to the atmosphere. An evaluation has been parformed to determine the effect on these faewrs due to the modifications of the Model D 5 steam generator level moasurement system, 4.2.2.1 ChangejfL61eam_Gener.atoLLo.w_LeyeLandBlotLLeyeLT11pEetpolrls The steam enerator low level and high level trip setaoints are not modeled in the SGTR anal sis. Therefore, a change to these setpo!nts will not affect the SGTR analysis in hs Byron /Braidwood UFSAR.
4.2.2.2 BelocatiortotSteam Gened.or Lower _Levellap The SGTR analysis in the Byron /Braldwood Ui:SAR is based on the consem.tlu assumption of 1% 'uel defects, and this a3sumption will not be affected by the modificatica to the steam generator level measurement system.
l l
/scl:1019:17
4.2.2.2 Relocation.olSteamR wtotlowetteveLTap (con't)
To calculate the change in pr' mary to secondary break flow and steam release to
. tho atmos)here, an evaluation for the D 5 modification was performed using the
- tesults of uhe SGTR analysis performed to support o)eration with VANTAGE 5 fuel with a T-hot range of 600'F to 618.4'F. The VA9TAGE 5 analysis concluded that the consequences of a SGTR as reported in the UFSAR will not be increased. The evaluation of the D 5 steam generator level measurement system modification showed the break flow and steam releases from the Model D 5 steam generators are bounded by the SGTR results performed for the VANTAGE 5 analyses. Therefore, the SGTR consequences as reported in the UFSAR will not be increased by the D 5 modifications.
4.2.2.3 SGIRRonclusion The ellect of modification of the Model D 5 steam generatorlevel measurement system has been evaluated for the SGTR analysis. It has been concluded that ,
the results of the SGTR analysis in the UFSAR remain valid. j 4.2.3 LQ_CAAccideota Small aerturbations in steady state steam generator level do not affect large break OCA, long term core cooling, hot leg switchover, or LOCA hydraulic forces. However, level changes can affect the small break LOCA analysis. In the small break evaluation model, steam generator level is initially assumed to be at eight feet above the top of the tubes. A steam generator water mass value is input based on steam generator specific performance data (as opposed to plant specific data). Before the LOCA transient begins, the small break code adjusts the levelin the steam generator until more important aarameters approach their desired 100% power values and a steady state condiulon is reached.
The nominal steam generator level and water mass at 100% power will not be altered by a change in the narrow band level tap location. This being the case, and since only a nominal mass value is analyzed, the small break LOCA analysis is not adversely affected by the change in level tap location. >
p l
l l
l l
p /scl:1019:18-
- . - _ ~_ _ _ _ . _ -
4.3 AdditionalConsiderations !
4.3.1 AT.WS MitigaticitSystem lmpact The changes were also evaluated for their indirect effect on the Anticipated Transient Without Scram (ATWS) Mitigation System Setpoints. The Byron /Braidwood ATWS mitigation system will automatically trip the main turbine and initiate Auxiliary Feedwater when level falls below a prescribed selpoint in 3/4 Steam Generators. The steam generator level setpoint for ATWS Initiation is specified as 3% of narrow range span below the nominal Low Low Steam '
- Gererator Level Reactor Trip / Auxillary Feedwater Initiation Setpoint. Therefore, when the Reactor Trip Setpoint is changed the ATWS Setpoint must be changed accordingly.
The guidance provided by Wostinghouse WCAP 11436 "AMSAC Generic Design Package" for selection of the ATWS Setpoint places restrictions on setpoint.
- - selection. The first is that it occur below the reactor 3rotection system Low Low I - Steam Generator level trip setpoint. The second is flat the reduced ATWS :
setpoint be no lower relative to the reactor trip setpoint, than the environmental and reference leg heatup allowances. For Byron and Braidwood stations the additive values 01 these two allowances is 12.6% of span. The third restrictiors is that the setpoint be no lowe than 5% level. It can be seen that the first two restrictions are met by virtue of the way the ATWS setpoint is specified ,
(36.353%=33.3%). The third restriction, a minimum value ci 5% lsvel,is easily met by the new setpoint value of 33.3%. Therefore, the ATWS mitigation system initiation assumptions are still valid.
4.3.2 N uclear_EueLSeMee s _(N ESh.SW Evaluation Although Westinghouse currently holds the record of analysis for Steam Generator Tube Rupture, NFS has submitted a new To alcal Report for NRC review. NFS periormed an evaluation of the proposed 735 S/G Tap Relocation Modification to determine the "cact on the NFS Byron and Braidwood Steam Generator Tube rupture Ana ,.,,s (Reference 2). Neither high high level trip nor low low level trip was credited in the NFS SGTR Analysis. Therefore, changes in level setpoints other than nominal level setpoint, do not impact the initial secondary water mass asumptions made in tho SGTR submittal. The norninal level setpoint mass assumption has been determined to remain bounding. In addition, the velocity head errors introduced by this modification will not delay the reactor operator's re.sponses while mitigating a SGTR accident in the SGTR analysis, NFS assumed reactor operator's responses while mitigating a SGTR accident from radiation monitors and from differentiallevels between the faulted L and non f aulted steam generators. Velocity head errors will be uniform across all i: four steam generators and not influence differential levels any more than the D4 steam generators.
In conclusion, the effect of modificallon of the D 6 steam generator level
. measurement system has been evaluated and the NFS SGTR analysis remains valid.
l
/se' 1019:19 wwe.m., ...,-,...-.....-m,,wm -- ---- ..,.-,y.- e n-,--y pr,,,,,,_,,,.. , ,., ,. om_,7,..p , . . - , . _ - _ , , , , _ ., . _ , , . . . . , , ,,.,-
~
~
4.4 - Determination _oLUnteyJeweiSafety_ Question
- 1. Will the probab!!ity of an accident previously evaluated in the SAR be :
-inc eased?
The modification to the D 5 steam generator level tap location will not increase the probability of an accident previously evaluated in the l UFSAR. Evaluations and reanalysis have determined that the modification of the various steam generator level setpoints will not affect :
any of the equipment or circumstances assumed for initiation of the ;
UFSAR accidents.
- 2. Will the consequences of an accident previously evaluated in the SAR be increased?
No, per the discussion presented in Section 4.2, all the applicable non LOCA acceptance criteria are still met for both the transients !
evaluated and the two events analyzed. The evaluation of the D 5 steam ;
generator level measurement system modification showed the break flow !
and steam releases from the Model D 5 steam generators are bounded by the steam generator tube rupture results performed for the Vanta e 5 analyses. Additionally, no new limiting single failure is introduced b the '
proposed change. Therefore, based on the above, there is no pote llal ',
"or an increase in the dose releases.
- 3. May the possibility of an accident which is different than any already evaluated in the SAR be created?
No, changing the tap location and the steam generator level setpoints for !
the D 5 steam generators does not introduce a new accident initiator mechanism or any new features into the plant such that a new or different type of accident could be created,
- 4. Will the probability of a malfunction of equipment important to safety previously evaluated in the SAR be increased? ,
The modification to the steam generator level measurement system will i not introduce any new features into the plant such that the probability of a i malfunction of equipment important to safety previously evaluated in the UFSAR will be increased.- Although the proposed change will affect several reactor protection system setpoints, changing the D 5 steam generator narrow range tap locations will not adversely affect the operation of the remainder of the reactor protection system, any of the other protection setpoints, or any device required for accident mitigation.
- 5. Will the consequences of a malfunction of equipment important to safety E previously evaluated in the SAR be increased?
u l
L
- /scl:1019:20.
1
s The modification to the steam generator lovel measuromont system will not challenge the integrity of the equipment assumed to be operable.
Therefo'e, the consequencoo of the malfunction already considered in the licensing basis safety analys will not be increased.
G. May the possibility of a malfunction of equipment important to safety different than any already evaluated in the SAR be created?
Since the modification of the steam generator level measurement system does not challongo the Integrity of the reactor coolant system and the change is not expected to directly or Indirectl affect any safety equipment rolled upon for safety, the malfunction of equ ment important to safety different than already evaluated in the UFSA is not expected to result.
- 7. Will the margin of safety as defined in the Bases to any Technical Specification be reduced?
No, as discussed in the attached safety evaluation, the proposed chan0e in the location of the narrow range taps will not invalidato any of the conclusions presented in the UFSAR accident analyses. The new loss of normal feodwater/ loss of all AC power and feedwater malfunction analyses concluded that all the applicable acceptance criteria were still ratisflod. For all the UFSAR non LOCA transients, the DNB design basis, orimary and secondary pressure limits, and dose release limits continue lo be met. The evaluation of the SGTR accident analysis ccasidered in the Byron /Braidwood Unit 2 licensing basis determined that the analysis acceptance criterla will continue to be met with the modification to the steam generator level measuremont system. Furthermore, the modification will improve safety since the modified steam generators will be less susceptible to foedwater transients, thus reducing the potential for reactor and turbine trips and avoiding unnecessary transients on the primary and secondary systems. Therefore, the margin to safety determined for the Byron /Braidwood Unit 2 licensing basis safety analyses remains unchanged.
4.5 COLclusl0Da Based upon the information presented above, it can be concluded that the modification of the steam generator level m9asurement system does not constitute an unroviewed safety question.
4.6 fieferences
- 1. WCAP-10961 P Rev.1, "Steamlino Break Mass / Energy Releases for Equipment Environmental Qualification Outside Containment", dated October 1985.
- 2. " Steam Generator Tube Rupture Analysis for Byron and Braidwood Plants, Revision 1," Nuclear Fuel Services, March 1990.
l l
i /scl:1019:21 l
j t
~
TADLE1 Current and Modified Technical Specification Setpoints for Byron /Braidwood Unit 2 Current Setpoints(1) Modified _Setpolats_(j)(2)(3)
Narrow Range Span 438 to 566 inches 333 to 566 inches (distance between taps)
Tap to Tap Distance 128 inches 233 inches Nominal Level 502 inches 482 inches (4)
(at full power) (50% NRS) (63.7% NRS)
Low Low Level 460 inches 418 inches (17% NRS) (36.3% NRS)
High High Level 538 inches 522 inches (78.1% NRS) (80.8% NRS)
Margin to High-High 36 inches 40 inches Margin to Low Low 42 inches 64 inches Total Operating Region 76 inches 104 inches NOTE: (1)- Values represented in the Table are in terms of Inches above the tubesheet, i.e. O inches = elevation of tubesheet.
(2) Modified setpoint levels are indicated levels.
(3) Setpoints are calculated as follows:
In.:hes = % NRS
- 233.778 in/NRS + 333 in.
Where: 233.778 in, accounts for growth of the S/G and 333 in represents the lower tap elevation.
(4) Actual level at 100% power is 502 inches.
l
/ sci:1019:22'
. IABLE_2 Relationship Between Safety Analysis Limit and Technical Specification Trip Value Non LOCA Safety Tech Spec Total Anajysisilmit(D Irlp_Value(1) Allowance SG level High High 99.7% NRS 80.8% NRS 18.9% I4RS (566 inches) (522 inches) (44 inches) ,
SG level Low-Low -18.6% NRS 36.3% NRS 17.7% NRS (Feodline Break) L(376.5 inches) (418 inches) (41.5 inches) l SG level Low Low 28.6% NRS 36.3% NRS 7.7% NRS (Loss of Normal FW) (400 inches) (418 inches) (18 inches) _
NOTE: The Total Allowance consists of a statistical combination of uncertainties associated with the sensor, rack, and process measurement, as well as ,
available margin. Specific components of the statistical combination of uncertainties .nclude such items cs sensor calibration, sensor drift, sensor temperature and pressure effects, arocess measurement, environmental allowances (when appropriate), ve ocity head effects (when appropriate), etc. '
This methodology is consistent with that used in WCAP 12583 and WCAP-12523, which are Westinghouse Proprietary Class 2 documents. The calculated Total Allowance must equal the cifference between the Safety Analysis Limit and the Technical Specification Trip Sctpoint.
l.
(1) inches as measured from top of the tubesheet.
j ._
l L
L i
- /scl:1019:23
-- - - . - . - . - - - . .-.- - .-..-_a ..-_...;--.___-...-
s 5.0 Scledule.Bequitements Scheduling of review of the proposed amendment is requested to support the next Unit 2 Byron refueling outage. The projected start date of this outage is February 29,1992, therefore review completion is requested by January 29,1992 to allow ;
sufficient time for planning, l
L l
/scl:1019:24
ATIACHMENT D PROMOSED. CHANGES.TO APPENDIX A TEC H NIC AL SP ECIFIC ATJON S.OEEACILITY OPERAllNG LICENSES NPE 06.AND. NEE-77. -
ByroRStation Braidwood. Station Revisedfages: 2-5 Boylsedfaces: 2-5 3/4325 3/4325 3/4 3 20 3/4326 l
l I
l l
/scl:1019:25
_ _ _ _ _ - _