ML20028H337
| ML20028H337 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/19/1990 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20028H336 | List: |
| References | |
| 1885, NUDOCS 9012210109 | |
| Download: ML20028H337 (8) | |
Text
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Dockst Nunbar 50-346 i
Lic;nsa Nunbar NPP-3 Sirial Numb 2r 1885 l
' Enclosure Page 1 APPLICATION FOR AMENDHENT TO FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UNIT NUMBER 1 Attached is a requested change to the Davis-Besse Nuclear Power Station, Unit Number 1 Facility Operating License Number NPF-3 Appendix A, Technical Specifications.
Also included is the Safety Assessment and Significant Hazards Consideration.
The proposed change (submitted under cover letter Serial Number 1885) concerns:
Technical Specification 3/4.3.2, Instrumentation, Steam and Feedwater Rupture Control System Instrumentation.
By:
s% 'N m D. C. Shelton, Vice President - Nuclear Sworn and subscribed before me this 19th day of December, 1990.
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Notary P)(blic, State of Ohio EVELYN L DRESS NOTMi'i PUSLC. iiTAT5 CFCH10 MyCrd,ss.1Ep;9;iyge,1cta 9012210109 901219 PDR ADOCK 05000346 p
Docket Numbar-50-346:
I
- Lictnss Numbsr NPF-3 Sarial Numb 2r 1885 Enclosure Page 2 The following information'is provided'to support issuance of the requested 3
change to the. Davis-Besse Nuclear Power Station,~ Unit Number 1 Operating j
License Number NPF-3, Appendix A,1 Technical Specifications,--Technical.
i Specification'3.3.2.2,_ Table: 3.3-ll.
- A. Time required tofimplement: This; change-is-to be: implemented within_45-days-after.NRC issuance of.-the License Amendment;by_the NRC.'
B. Reason for-change.(License: Amendment Request Number 90-0046)t> _ Thisichange-1 vill minimize the possibility of an inadvertentemain steam low pressure:
j trip occurring during plant cooldown and heatup by~ modifying the'lov pressure Steam and FeedvaterzRupture' Control System trip; block permit' setpoint--from 700 psig;to 750 psig and; increasing the steam pressure where
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the. block permitcis automatically removed-to 800 psig.-
C. Safety Assessment andLSignificant HazardsfConsiderations See Attachment.
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Dockst Nunbar 50-346 Licensa Numb 2r NPF-3 S2ricl Numbar 1885-Attachment Page 1 of 11 SAFETY ASSESSMENT AND SIGNIFTGNT HAZARDS CONSIDERATION FOR LICENSE AMENDHENT REQUEST NUMBER 90-0046:
~ TITLE Proposed change to the Davis-Besse Nuclear Power Station,' Unit 1 Operating License, Appendix A, Technical Specification 3/4.3.2, Safety System Instrumentation,-Steam and Feedvater Rupture Control System Instrumentation,-
Table 3.3-11 Steam and Feedvater Rupture Control System Instrumentation, t
DESCRIPTION The_purpou of this Safety Assessment and Significant Hazards Consideration:
is to review the proposed change to the Davis-Besse Nuclear Power Station Unit No. 1 Technical. Specifications (TS) to ensure that the: change does not-constitute a significant hazards consideration._ The proposed TS change is.
to increase the Steam and Feedvater Rupture. Control System (SFRCS) Main-Steam (HS) low pressure block permit setpoint specified in_TS Table ~3.3-11 from 700_psig to 750 psig and to increase.the steam pressure where the-block is Suiomatically removed from 750-psig to 800 psig. :This-change-vill <
berease the pressure margin between the SFRCS block permit;and the SFRCS HS low pressure trip setpoints, thereby minimizing the possibility of an-inadvertent MS low-pressure trip during plant cooldown_ operations. =This change vill also increase-the margin between the automatic block reset and-the low pressure trip' reset to ensure:that;the low pressure trips have
-cleared prior to the automatic resetiof the SFRCS=during plant startup i
operations.
SYSTEMS, COMPONENTS AND ACTIVITIES AFFECTED j
SteamandFeedvaterRupture.ControlSystem[(SFRCS)-
~ SAFETY FUNCTIONS OF THE AFFhCTED SYSTEMS, COMPONENTS [ AND ACTIVITIES The SFRCS is'an automatic system designed _to_detectiand mitigate the effects-
- of -major upsets in the.HS and Main Feedvater- (HFV) -systems,.-including MS' and HFV line ruptures, loss.of HFV events, Steam. Generator _(SG) overfeed,Eand a
. loss of Reactor Coolant System-(RCS) forced circulation cooling. The SFRCS
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detects.these events through. sensing and logicLchannels-and mitigates-their-consequences by automatically positioning valves!in the-MS, MFW, and Auxiliary Feedvater -(AFV): systems vith: appropriate? actuation signals -
denendent upon the-initiating _ event..
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The SFRCS consists of,four identical sensing and logic channels housed in two electrically separate cabinets.
Each cabinet consists of two redundant =
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- sensing and logic channels.
Logic ChannelsJ1-and 3 are locatedlin Cabinet =1 and form Actuation Channel 1 (predominantly SG 1).
Logic Channels-2 and 4 are-lccated in Cabinet 2 and form Actuation Channel 2 (predominantly SG 2).
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Dockst Nu2bar 50-346 Lic:nsa Numb:r NPF-3 Sorial Numb 2r 1885 Attachment Page 2 of 11 The sensing channels consist of the instrumentation used to monitor the various parameters which provide inputs to the logic channels.
These inputs include:
- SG high and lov vater level
- MS lov pressure
- SG to HPV differential pressure RCP high and lov motor current The only SFRCS sensing instrumentation relevant to this proposed change is the MS low pressure signal and high SG vater level signal.
The High Vater Level trip is not required for mitigation of any design basis accident. Therefore, this function vill not be discussed.
A HS low pressure trip of an Actuation Channel of SFRCS during plant operation would be indicative of a main. steam line break (HSLB). The MS low pressure trip instrumentation includes pressure switches for all four SFRCS logic channels on each MS header.
A trip of single pressure switches in both Logic Channels 1 and 3 on a MS header would cause a trip of SFRCS Actuation Channel 1 whereas a-trip of single pressure switches in both Logic Channels 2 and 4 on a HS header would cause a trip of SFRCS Actuation Channel 2.
An SFRCS Actuation Channel trip would cause the complete isolation of the SG connected to the MS header experiencing the trip signal, the re-alignment of the affected SG's AFV pump to the opposite SG, and the initiation of AFV to.the unaffected SG.
It would also close the main steam line isolation valve and selected main feedvater valves on the unaffected SG.
The SFRCS also includes a manual low pressure block permissive feature that allows the operator to block the SFRCS MS low pressure and SG High Vater Level trip signals during plant cooldovn. This manual operator action is intended to prevent inadvertent actuation and unnecessary'chal. ans to SFRCS and associated systems during plant cooldown..Each MS header contains four pressure switches for the block permissive signal. Two of these pressure switches are associated with a single SFRCS Logic Channel and the remaining two pressure switches on that MS header are associated with the complimentary logic channel in that Actuation Channel. :This results in the four pressure switches for the block permit on the MS header for SG 1'being used to block Actuation Channel 1 and the four pressure switches on the MS header for SG 2 being used to block Actuation Channel 2.
Once a block permit for a channel'is received, manual action is required to actually block that channel from tripping.
As required to comply with IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," for a protection system and to satisfy the Technical Specification requirements, during plant heatup the SFRCS low pressure block signal is automatically removed. This action ensures that the safety function of the SFRCS MS low pressure trip signal is activated automatically during heat-up operations and remains activated when the plant is critical in either Modes 1 or 2.
Docket Nunbar 50-346' l
License Numbar NPF-3l l
Sarial Numbar 1885_
Attachment i
Page 3 of 11 i
EFFECTS ON SAFETY
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DBNPS Technical Specifications, Limiting condition for Operation 3.3.2.2 requires that:
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"The Steam and Feedvater Rupture Control System (SFRCS) instrumentatiori channels-shown in Table 3.3-11 shall be OPERABLE vith their trip setpoints set consistent with the values shown infthe Trip Setpoint column of Table 3.3-12 and with-RESPONSE TIMES as-shown in Table 3.3-13."
One of the types of SFRCS sensing-instrumentation listed in Table 3.3-11 is 3
the MS low pressure l instrumentation. -Table 3.3-12' lists a-trip:setpoint;for the SFRCS MS low pressure trip signal las greater 1than or equal to-591.6 psig. A footnote in Table 3.3-11 for the low pressure inctrumentation j) i channels states that this instrumentation ~
"May be bypassed when steam pressure is below 700 psig.. Bypass shall i
be automatically removed when the steam-pressure exceeds 750 psig."-
This proposed TS change vill increase the steam pressure in the above i
footnote to 750 psig belov which the SFRCS MS low pressure instrumentation.
- I can be manually bypassed.
Additionally, the-steam pressure above which the block permit is automatically removed vill be increased.to'800 psig.
-l As noted above, there are two-block permit; pressure scitchestfor each SFRCS MS low pressure logic channel. During pressure increase,Ethe block in both logic channels associated'vith an Actuation Channel needs to reset to automatically-remove the block'in theiassociated Actuation Channel. However, only one.of the-two pressure-switches in each, logic-channelLneeds to automatically reset to remove thatilogic channel'siblock. The-MS lov-pressure trip signal is provided by separate pressureisvitches which'hav' much vider reset dead bands than the: Block "ermit pressure switches. This s
Ecould result in the block _being automaticallyiremovediprioreto theftrip _
signal having cleared causing an-SFRCS activation.' Since the:same switches are used for the block permit and the automatic-reset,-both vill be addressed.
1 The-reason for the setpoints.potentially; overlapping as' described above is that the reset-dead band associated with1the only available environmentally qualified replacement low pressure trip-svitches is larger than that of the previously installed switches.
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i The' primary purpose of-the MS lovipressure-instrumentation ofithe SFRCS'is steam line~ break detection and mitigation'in= Modes 1.and(2.- During-plan't cooldown operations, the RCS temperature' closely approaches the saturation temperature associated with the SG secondary side pressure.
Consequently,
- the present' block permit,value
- of 700 psig correspondsfto an'RCSJtemperature
-of-approximately 506'F.
Since the nominal-RCS'temperatureLin-Mode 3 immediately following a reactor trip 11s approximately 545'P,-which-corresponds to the saturation temperaturecfor-the_ post-trip turbine bypass-valve:setpoint of 995.psig, SFRCS main steam line break: protection isL available only for a duration associated with approximately a 40'F cooldown in Mode 3.
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-Dockst Nurbar 50-346 Licensa Nuaber NPF-3 1
S2riol Number 1885-Attachment i
Page 4 of 11 Thus, the present block permit setpoint results in the SFRCS MS lov pressure instrumentation channels being blocked over most of the RCS temperature
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range associated with Mode 3.
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By raising the block permit setpoint to 750 psig, the RCS temperature at l
vhich the MS low pressure instrumentation can be blocked is increased from-l cpproximately.506'F to approximately-514'F..This represents an increase of
-l 8'F in the range of RCS temperatures in Mode 3 vhere the MS lov pressure ~
instrumentation would be unavailable _during cooldown operations. The RCS temperatures from 506'F.to 514'F-represent a range of transient plant 1
operations in Mode-3 and do mot reptesent temperatures _in Mode-3 where the-L RCS vould be stabilized for any long periods of' time.._Using a nominal
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cooldovn rate of 15'F/hr, the raising of the block permit'value to 750 psig vould increase by approximately 30 minutes the period of time ~in Mode-3 where the MS low pressure instrumentation'would be unavailable during a-normal plant cooldown operation.
The probability that a MSLB vould occur in this short time period is extremely lov.
By raising the automatic reset pressure of the block to 800.psig from 750 psig, the dead band associated vith the resetting of the low pressure trip switches vill not overlap the automatic reset of the block switches._ Based on past surveillance tests of the block permit pressure switches, the switches typically reset within 20-30 psi above the block permit setpoint.
Consequently, the 50 psi difference between the block permit setpoint and 2
automatic reset point specified-by Technical Specifications is considered to be an appropriate pressure range for the equipment.in use.
i By raising the automatic reset value to 800 psig, the RCS temperature at which the block automatically resets is increased from approximately 514'F to approximately 520*F. This increases'by 6'F the range of.RCS temperature in Mode 3 where the MS low pressure instrumentation would be unavailable during heat-up operations. However, it still ensures that;the automatic reset occurs before the plant enters Mode 2, since per TS 3.1.1.4 the plant is not allowed to go critical until-RCS T is greateri han or equal-to t
525'F.
ave Using a nominal heat-up rate of 15'F/hr, the raising of the automatic reset setpoint to 800 psig would increase the time period _during heat-up where the MS low pressure-instrumentation is blockedby approximatelyL30 minutes. As with the increased time period associated with plant cooldown operations, this time period is so short that the probability of a MSLB during-this interval-is extremely lov.
Since during normal plant operation in Modes 1 and 2 the MS line pressure is typically 870 psig, the raising of the block permit pressure to 750 psig and the automatic reset to 800 psig has no impact upon_ plant operation in Modes 1 and 2.
Use of the block permit in Modes 1 and.2 is_not possible due to the large difference in pressure between its setpoint and the normal:HS operating pressure.
Consequently, the protection against MSLBs during power operation providad by the MS low pressure instrumentation is unaffected by the proposed change.
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Docket Nunbar 50-346-1)
Liesnso Numb 2r NPF-3 Sarial Number 1885
.j Attachment i
Page 5 of 11-i SIGNIFICANT HAZARDS CONSIDERATION The Nuclear. Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining "hether a significant hazard exists due to a proposed amendment to an Operating License for a facility.
A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes would
,. 1) Not involve a
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significant increase in the probability or-consequences of an accident previously evaluatedt (2) Not create _the possibility of a nev-or different kind of accident from any accident previously. evaluated; or (3) Not involve a significant reduction in a margin of safety.- Toledo Edison has reviewed the proposed change and determined that a significant hazards consideration-does not exist because operation of the Davis-Besse Nuclear Power Station Unit 1 in accordance with these changes _vould la.)
Not involve a significant increase in the probability of an accident-previously evaluated because-the pressure switches associated with the change do not initiate any accident previously analyzed. The pressure svitches only allow a manual bypass function-for the MS low pressure trip switches to be activated by the operators.
j Additionally, the potential for an inadvertent SFRCS MS lov pressure i
trip during plant heatup or cooldown operations vill be reduced.
Ib.)
- t involve a significant increase in the consequences of an accident previously evaluated because the pressure switches, associated with-the change do not play any mitigating. role in any accident previously analyzed. The pressure switches only allow a manual bypass function for the MS low pressure trip switches to'be performed by the operators during controlled evolutions during Mode 3. Additionally, the potential for an inadvertent SFRCS MS low pressure trip during plant heatup or cooldown operations vill-be redticed. This change _ does not alter the radiological consequwces of the_ bounding' main steam line break accident evaluated in the USAR.
2a.)
Not' create the possibility-of a new kind.of accident from any.
accident previously evaluated because the setpoint change does not alter the safety function of'SFRCS or any associated systems. The 4
revised setpoints provide the same function as before and do not introduce failure modes:that are not bounded _by. existing analyzed events.
2b.)
Not create the possibility of a different kind of' accident from any accident previously evaluated because the setpoint change does not-alter the safety _ function of SFRCS or any associated systems. The revised setpoints provide the same function'as before and do'not introduce failure modes that are not bounded by existing analyzed:
events.
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= Docket Nuhbar 50-346:
d License Numbar.NPF-3 Sarial.Numbar 1885 L-
Attachment:
Page 6 of II 3.)
.Not involve a significant reduction in.a. margin of safety because the c!
change' minimizes the possibility of an unnecessary actuation.of-thet 1!
AFV system during-plant cooldown.and-heat-up operations.
The: change _
.'t in the setpoints has no impact upon-the_ availability of SFRCS duringi plant power. operations-and does not appreciably. increase the time _
l period-in: Mode 3-where:thejSFRCS main steam lov pressure-trip signal is blocked.-
iq CONCLUSION-On the basis of the above, ToledoLEdisonJhas-' determined-that the License 1 Amendment Request does not: involve a.significant-hazards consideration. As'-
'this. License Amendment. Request concerns a proposed change 3toLthe Technical-
-Specifications that must be reviewed byithe Nuclear; Regulatory Commission.
e this License Amendment Pequest does noticonstitutetan unreviewed safety 1 question.
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ATTACHMENT-Attached are the proposed marked-up'changesi o the dperating License'.ti j
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