ML20206T603
| ML20206T603 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/25/1986 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML19292F547 | List: |
| References | |
| NUDOCS 8607080187 | |
| Download: ML20206T603 (12) | |
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ATTACHMENT I TO JPN-86-29 i
PROPOSED CHANGE TO THE TECHNTCAL SPECIFICATIONS REGARDING WATER LEVEL SETPOINT FOR THE MAIN STEAM ISOLATION VALVES, MAIN STEAM LINE DRAIN VALVES AND REACTOR WATER SAMPLE LINE ISOLATION VALVES J
NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 8607080187 860625 PDR ADOCK 05000333 P
PDR m
l JAFNPP TABLE 3.2-1 INSTRUMENTATION THAT INITIATES PRIMART CONTAINMENT ISOLATION Total Number of Instrument Cicimum Number of Channels Provided by Design Action l
Operable Instrument Channels j
per Trip System (1)
Instrument Trio Level Settina for Both Trio Systems (2) 2 (6)
Reactor Low Water 5 12.5 in. Indicated 4 Inst. Channels A
Level Level () 177 in. above j
the. top of active fuel) 1 i
/
1 Reactor High Pressure f 75 psig 2 Inst. Channels D
Isolation) 2 Reactor Low-Low-Low D18 in, above the top of 4 Inst. Channels A
Water Level active fuel.
i j
j 2 (6)
High Drywell Pressure Q.7 psig 4 Inst. Channels A
j 2
High Radiation Main
( 3 x Moraal Rated 4 Inst. Channels B
Steam Line Tunnel Full Power Background (9) 2 Low Pressure Main 5825 psig (7) 4 Inst. Channels B
Steam Line 1
2 High Flow Main Steam
( 140% of Rated Steam 4 Inst. Channels B
Line Flow j
1 2
Main Steam Line Leak f40*F above max 4 Inst. Channels B
Detection High ambient Temperature J
)
3 Reactor Cleanup Sys-f40*F above man 6 Inst. Channels C
i I
tem Equipment Area ambient High Temperature 2
Low Condenser Vacuum
)8" Hg. Vac (8) 4 Inst. Channels B
1 Closes MSIV's Amendment No. }W" f, g, pf,)Hf f
64 i
1
JAFNPP NOTES FOR TABLE 3.7-1 ISOLATION SIGNAL CODgS Description Sinnal Reactor vessel low water level - (A scram occurs at this l
A*
level also. This is the highest of the three low water i
level signals.
(
B*
Reactor vessel low-low-low water level - (This is the lowest of the three low water level signals.
i C*
High radiation - main steam line Line break - main steam line (steam line high steam flow)
D*
2 Line break - main steam line (steam line high temperature)
E*
j F8 High drywell pressure
)
Reactor vessel low water level or high drywell pressure G
(Energency Core Cooling Systems are started) 1 3
3*
Line break in Reactor Water cleanup System - high space temperature i
Line break in RCIC System steam line to turbine (high K*
steam line space temperature, high steam flow, low steen j
j line pressure, or high turbine exhaust pressure)
Line break in HPCI System steam line to turbine (high L*
steam line space temperature, high steam flow, low steam i
line pressure, or high turbine exhaust pressure)
M Low main steam line pressure at inlet to main turbine P*
(RUN mode only)
S Low drywell pressure l
T Low reactor pressure permissive to open core spray and i
RHR-LPCI valves
- These are the isolation functions of the Primary Containment and Reactor Yessel Isolation Control System; other functions are given for information l
only.
1 of 4 Amendment No.)MI I
- _... ~.,...
i
~
i JAFEPp
}
3.2 BASES In addition to reactor protection instrumentation has a direct bearing on safety, are chosen at a level which initiates a reactor scram, protective instru-away from the normal operating range to prevent ined-j montation has been provided which initiated action to vertent actuation of the safety system involved and citigate the consequences of accidents which are exposure to abnormal situations.
beyond the operator's ability to control, or termi-i nates operator errors before they result in serious Actuation of primary containment valves is initiated consequences.
This set of specifications provides by protective instrumentation shown in Table 3.2-1
}
the limiting conditions of operation for the primary which senses the conditions for which isolation is I
cystem isolation function, initiation of the Core required.
Such instrumentation must be available cooling Systems, control Rod Block and Standby Gas whenever primary containment integrity is required.
Treatment Systems.
The objectives of the specifica-l tions are to assure the effectiveness of the protec-The instrumentation Which initiates primary system tive instrumentation when
- required, even during isolation is connected in a dual bus arrangement.
periods when portions of such systems are out of service for maintenance, and to prescribe the trip The low water level instrumentation set to trip at settings required to assure adequate performance.
177 in, above the top of the active fuel closes all l
When necessary, one channel may be made inoperable isolation valves except those in Group 1.
Details of 1
for brief intervals to conduct required functional valve grouping and required closing times are given
{
tests and calibrations, in specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent some of the settings on the instrumentation that uncovering the core in the case of a break in the t
initiate or control core and containment cooling have largest line assuming a 60 see valve closing time.
l tolerances. explicitly stated where the high and low Required closing times are less than this.
values are both critical and may have a substantial l
effect on safety. The set points of other instrumen-The low-low reactor water level instrument: tion is i
tation, where only the high or low end of the setting set to trip When reactor water level is 126.5 in.
above the top of active fuel
(-38 in.
on the Lastrument). This trip e
9 l
Amendment No.
55
9 JAFNPP initiates the HPCI and RCIC and trips the the breaks discussed above, this iststrumestatics will Irecirculation pumps.
The low-low-low reactor water generally initiate ECCS operation before the low-low-level instrumentation is set to trip when the water low water level instrumentation; thus the results level is 18 in, above the top of active fuel.
This given above are applicable here also.
See Specif1-l trip activates the remainder of the ECCS subsystems, cation 3.7 for isolation valve closure group.
The closes the main steam isolation valves, main steam water level instrumentation initiates protection for line drain valves and reactor water sample line the full spectrum of loss-of-coolant.2ccidents.
Isolation valves, and starts the emergency diesel generators.
These trip level settings were chosen to Ventuels are provided in the main steam lines as a means of measuring steam flow and also limiting the be high enough to prevent spurious actuation but low enough to initiate ECCS operation and primary system loss of mass inventory from the vessel during a steam isolation so that post-accident cooling can be line break accident.
The primary function of the cccomplished and the guidelines of 10CFR100 will not instrumentation is to detect a break in the main
~
be exceeded.
For large breaks up to the complete steam line.
For the worst case accident, aain steam
~
circumferential break of a 24 in, recirculation line line break outside the drywell, a trip setting of 140 cod with the trip setting given above.
ECCS percent of rated steam flow in conjunction with the initiation and primary system isolation are initiated flow limiters and main steam 11ae valve closure, in time to meet the above criteria.
Reference limits the mass inventory loss such that fuel is not paragraph 6.5.3.1 FSAR.
uncovered, fuel temperature peak at approximately 1,000*F and release of radioactivity to the environs The high drywell pressure instrumentation is a diverse is below 10CFR100 guidelines.
Reference Section signal for malfunctions to the water level instrumen-14.6.5 FSAR.
tation and in addition to initiating ECCS, it causes Isolation of Groups B and 3 isolation valves. For Amendment No. )d, ps, MI 56
e ATTACHMENT II TO JPN-86 29 SAFETY EVALUATION FOR THE PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS REGARDING WATER LEVEL SETPOINT FOR THE MAIN STEAM ISOLATION VALVES. MAIN STEAM LINE DRAIN VALVES AND REACTOR WATER SAMPLE LINE ISOLATION VALVES NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59
I.
Description of the Proposed Chance The proposed amendment to the FitzPatrick Technical Specifications changes the water level set point for MSIV, MSLDV and RWSV closure from reactor low-low water level (Level 2) [3t126.5 inches from top of active fuel (TAF)] to reactor low-low-low water level (Level 1) (3r 18 inches from TAF).
This change is reflected in Table 3.2-1 on page 64, in Notes for Table 3.7-1 on page 206 and in the Bases on pages 55 and 56 of the Technical Specifications.
II.
Purpose of t*~
Proposed Chance The proposed change for the MSIV water level setpoint would reduce the pocsibility of spurious valve closure due to variation of water level, reduce challenges to the Safety Relief Valves and minimize suppression pool heatup.
The set of water level instruments controlling the MSIVs also control the MSLDVs and the RWSVs and the setpoint for these valves is also lowered.
III.
Impact of the Proposed Chance The proposed change is based on the results of the analyses performed by General Electric for the FitzPatrick plant (Reference II).
In these analyses, G.E.
evaluated the safety impact of lowering the water level setpoint for MSIVs, MSLDVs and RWSVs.
i Change in the setpoint for the MSIVs may affect several operating parameters; e.g.,
the minimum critical power ratio (MCPR), peak vessel pressure, radiation release and shutdown capability during normal operational transients.
It may also affect the fuel cladding integrity during a LOCA and the reactor response during an ATWS event.
The results of these analyses are summarized below:
A. Transient Events The potential impact of the setpoint change for transient events was determined by reviewing all the abnormal operational transients evaluated in the FitzPatrick Final Safety Analysis Report (PSAR).
For all the transient events analysed in the FSAR, e.g.
loss of feedwater flow, loss of auxiliary power, generator trip, control rod withdrawal during power operation, feedwater controller failure - maximum demand etc., it can be shown (Reference II), that only two events can be affected by changing the water level setpoint at which the MSIVs isolate.
These events are feedwater controller failure - maximum demand and loss of feedwater flow.
In the loss of feedwater flow event, reactor core isolation cooling (RCIC) is capable of providing adequate core cooling even if the MSIV water level trip is lowered to reactor low-low-low water level (Level 1).
The RCIC flow is sufficient to compensate for the steam flow through the turbine bypass valves to the main condenser.
It also maintains the reactor water level 11-1
above Level 1.
Consequently, throughout the event, the I
water level remains above TAF and the turbine bypass valves I
maintain the reactor pressure of 950 psig.
This precludes any SRV operation and also eliminates the suppression pool heatup.
Thus the MSIV setpoint change will not compromise core cooling capability for the loss of feedwater flow event, and pool heatup will be reduced because the main condenser is available for a longer time.
The feedwater controller failure - maximum demand event causes the feedwater pumps to trip at high water level.
The subsequent core cooling is similar to the loss of feedwater flow event.
Therefore. the MSIV setpoint change will not have any. adverse effect on the feedwater controller failure - maximum demand event.
Thus, the impact of the setpoint change on abnormal operational transients will not cause, a reduction in MCPR, an increase in peak reactor vessel pressure, an increase in radiation release, equipment damage, a reductior in plant i
shutdown capability or a decrease in core cooling capability.
B. Loss of Coolant Accidents i
Large and intermediate LOCA events will not be affected by l
the setpoint change because the rapid depressurization and l
inventory loss will cause the MSIVs to close almost immediately after the accident, before any fuel failure will occur.
Consequently, the lower MSIV trip will not increase the inventory loss from the reactor core or the radiation release to the environment and the limiting maximum average planar linear heat generation rate (MAPLHGR) will not be changed.
Therefore, the setpoint change will not affect the design basis accident (DBA)~LOCA event.
Main steam line (MSL) breaks will not be affected by the setpoint change because other MSIV' isolation signals such as high flow, high tunnel temperature or low pressure at the turbine stop valve would occur well before a low reactor water level isolation signal for a MSL break.
Therefore, lowering the setpoint would not have any impact on the MSL break results.
Small break accidents are not limiting events for the FitzPatrick plant.
However, since there is potential for MS1V and SRV action during small break LOCA, the highest peak cladding temp (PCT) calculated for the limiting small break and the worst case single failure (i.e. failure of the HPC1 System) is substantially less than the 22000F PCT limit.
Therefore the setpoint change will have no effect on the limiting MAPLHGR.
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C. Anticipated Transient Without Scram (ATWS)
Lowering the MSIV water level trip to Level 1 will not cause any safety concerns during an ATWS event for the following reasons:
(1)
The recirculation pump trip at Level 2 reduces the reactor power significantly because of increased void generation.
(2)
The initiation of HPCI and RCIC at Level 2 will make up the reactor inventory loss through the steam lines.
(3)
By maintaining the main condencer availability for as long as possible, the delayed isolation will prevent excessive suppression pool heatup.
The MSLDVs are considered part of the MSL isolation system and the amount of radiation release and inventory loss though the drain valves will be insignificant when compared to the MSIVs.
The drain line valves are normally closed during plant operation and the probability of the valves being opened is not a function of the reactor water level.
Therefore, changing the isolation set point for these valves along with the MSIVs will have no effect on the analysis discussed for the MSIV change.
The isolation of RWSVs is currently obtained through the signal from Level 2.
The amount of inventory loss through the RWSV line for a break in the line outside the containment would be insignificant.
The RWSVs will isolate on the same signal as the MSIVs and since it has been demonstrated in the analysis above that lowering the water level trip for the MSIVs will not increase the radioactive materials released during a DBA, isolation of the RWSVs at Level 1 will not affect the calculated radiation doses.
The results of the above evaluation show that a change in the water level isolation set point from Level 2 to Level 1 for the MSIVs, MSLDVs and RWSVs will produce acceptable safety margins for any abnormal operational transients, LOCA or ATWS events.
The safety margin of the plant as defined in the FitzPatrick FSAR is not reduced.
IV.
Sionificant Hazards Considerations As defined in 10 CFR 50.92, this proposed amendment to the FitzPatrick Technical Specifications can be classified as not likely to involve a significant hazards consideration since it does not:
(1) involve a significant increase in the probability or consequences of an accident previously evaluated because all the abnormal operational transient, LOCA and Anticipated Transients Without Scram (ATWS) events which II-3
are analyzed in the FitzPatrick Final Safety Analysis Report (FSAR) have been evaluated and the proposed change will have no adverse effect on the safety of the plant as discussed below:
a) Abnormal Operational Transients:
-For all of the abnormal operational transient events analyzed in the FSAR, only two events can be affected by changing the water level setpoint at which the MSIVs isolate.
These events are loss of feedwater and feedwater controller failure - maximum demand.
For the loss of feedwater event, Reactor Core Isolation Cooling (RCIC) is capable of providing adequate core cooling even if the MSIV water level setpoint is lowered to Level 1.
RCIC flow is sufficient to compensate for the steam flow through the turbine bypass valves to the main condenser.
RCIC also maintains the reactor water level above Level 1.
Consequently, throughout the event, the water level remains above the top of active fuel (TAF) and the turbine bypass valves maintain the reactor pressure at 950 psig.
This precludes any Safety Relief Valve (SRV) operation and also eliminates suppression pool heat up.
Thus the proposed change will not compromise core cooling capability for the loss of feedwater event.
The feedwater controller failure - maximum demand event causes the feedwater pumps to supply excess feedwater.
This would cause a simultaneous turbine trip and reactor scram at high water level.
The subsequent core cooling is similar to the loss of feedwater event and therefore MSIV set point change will have no adverse effect on the feedwater controller failure - maximum demand event.
b) Loss of Coolant Accidents:
Loss of Coolant Accident (LOCA) events will not be affected by the proposed setpoint change because for large and intermediate LOCA events, and inventory loss through the break will rapidly depressurize the reactor vessel.
The lower setpoint will not increase the inventory loss from the vessel.
The new setpoint will still cause the MSIVs to close almost immediately after the accident, before any fuel failure will occur.
Main Steam Line (MSL) breaks will not be affected by the setpoint change because other MSIV isolation signals such as high flow, high tunnel temperature or low pressure at the turbine stop valve would close the MSIVs well before a low reactor water level isolation signal.
II-4
T Small break LOCAs are not limiting events for the FitzPatrick plant.
For these events, the reactor will remain pressurized until the initiation of the Automatic Depressurization System (ADS), assuming the single failure of the HPCI system.
With the MSIV water level setpoint lowered to Level 1, the highest peak cladding temperature (PCT) calculated for the limiting small break and worst case single failure (i.e. failure of HPCI system) is substantially less than 2200*F PCT limit.
c) Anticipated Transients Without Scram:
Lowering the MSIV water level setpoint to Level 1 will not cause any safety concerns during an ATWS event.
This is because the recirculation pump trip at Level 2 reduces the reactor power significantly as a result of increased void generation.
The initiation of HPCI and RCIC at Level 2 will make up the reactor inventory loss which occurs through the steam lines.
By maintaining the main condenser availability as long as possible, the delayed isolation will prevent excessive suppression pool heat up.
The water level instruments which control the MSIVs also control the Main Steam Line Drain Valves (MSLDVs) and Reactor Water Sample Line Isolation Valves (RWSVs).
Changing the setpoint for these valves does not alter the conclusions stated above.
(2) create the possibility of a new or different kind of accident from any accident previously evaluated because the MSIV, MSLDV and RWSV water level setpoints are safety features designed to mitigate the effects of previously evaluated events.
Failure of this safety feature whose only design function is to mitigate events, cannot by itself cause an accident.
(3) involve a significant reduction in margin of safety.
The affect of the proposed change on critical operating parameters; e.g.,
the minimum critical power ratio (MCPR), peak vessel pressure, radiation release and shutdown capability during abnormal operational transients, fuel cladding integrity during a LOCA and the reactor response during an Anticipated Transient Without Scram (ATWS) event, has been fully analyzed.
The analyses show that the affect of the setpoint change, on these parameters, would not result in a reduction in the safety margin of the plant.
Furthermore, this change reduces the possibility of spurious valve closures due to water level variations, reduces challenges to the SRVs and minimizes suppression pool heat-up, thereby increasing the operational safety of the plant.
II-5
Based on the above considerations, the proposed amendment meets the Commission's standards in 10 CFR 50.92(c) that the Application involves no significant hazards consideration.
V.
Implementation of The Proposed Chance The proposed change does not adversely impact the ALARA, Security or Fire Protection programs at the FitzPatrick plant, nor does it impact the environment.
VI.
Conclusion The change, as proposed, does not constitute an unreviewed L
safety question as defined in 10 CFR 50.59, that is, it (a) will not increase the probability or the consequences of an accident or malfunction of equipment important to safety as evaluated previously in the Safety Analysis Report (SAR): (b) will not increase the possibility of an accident or malfunction of a type other than that evaluated previously in the SAR: (c) will not reduce the margin of safety as defined in the basis for any Technical Specification: (d) does not constitute an unreviewed safety question; and (e) involves no significant hazards considerations, as defined in 10 CFR 50.92.
VII.
References
- 1. James A. FitzPatrick FSAR Section 14.
- 2. General Electric Report NEDC-30838-1 Rev 1 dated June 1985,
" Safety Review of Safety Relief Valve Simmer Margin Analysis and Water Level Setpoint Change for the James A.
l FitzPatrick Nuclear Power Plant."
(Proprietary)
Sub-sections 1.2 and 2.2 Sections 4, 5 and 6 l
II-6