ML20246M974
| ML20246M974 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/25/1989 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20246M965 | List: |
| References | |
| NUDOCS 8909080008 | |
| Download: ML20246M974 (13) | |
Text
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ATTACHMENT A
Revise the Technical Specifications as follows:
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, LIMITING SAFETY SYSTEM SETTINGS BASES (Steam /FeedwaterFlowMismatchandLowSteamGeneratorWaterLevel The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Pro-tection System.
This trip is redundant to the Steam Generat'or Water Level Low-Low trip.
The Steam /Feedwater Flow Mismatch portion of tnis trip is activated when the steam flow exceeds the feedwater flow by > 1.55 x IOS lbs/ hour in any loop.
The Steam Generator Low Water level portion of the trip is activated I
whc.n the water level drops below 25 percent, as indicated by the narrow range I
instrument.
These trip values include sufficient allowance in excess of normal operating values to preclude sputtous trips but will initiate a reactor trip before the steam generators are dry.
Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.
-COeleQ Undervoltage and Underfrequency - Reactor Luolant Pump Bust,es The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protettion against DNB as a result of loss of voltage or under-frequency to more than one reactor coolant pump.
The specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached.
Time delays are incorporated in the under. frequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transienis.
For undervoltage, the delay is set so that the time required for a signal to reach th'e reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.
For underfre-quency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is reached shall not exceed 0.6 seconds.
On decreasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10 percent of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10 percent of full power equivalent); and on increasing power, refnstated automatically by P-7.
Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9.
Each of the turbine trips provide turbine protection and reduce the severity of the ensuino transient.
No credit was taken in the accident analyses for operatien of these trips.
Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.
BEAVER VALLEY - UNIT 2 B 2-6 (Preposed We4d
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BEAVER VALLEY - UNIT 2 3/4 3-3 (Propen} WcNh
TABLE 3.3-2 g
REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES F_UNCTIONAL UNIT RESPONSE TIME 1.
Manual Reactor Trip NOT APPLICABLE 2.
Power Range, Neutron Flux 5 0.5 seconds
- 3.
Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.
Power Range, Neutron Flux, High Negative Rate 1 0.5 seconds" 5.
Intermediate Range, Neutron Flux NOT APPLICABLE 6.
Source Range, Neutron Flux NOT APPLICABLE (Below P-10) 7.
Overtemperature aT 5 4.0 seconds
- 8.
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< 4.0 seconds
- 9.
Pressurizer Pressure--Low
< 2.0 seconds (Above P-7)
- 10. Pressurizer Pressure--High 5 2.0 seconds
- 11. Pressurizer Water Level--High NOT APPLICABLE (Above P-7)
- 12. Loss of Flow - Single Loop (Above P-8) 5 1.0 seconds
- 13. Loss of Flow - Two Loop
-< 1.0 seconds (Above P-7 and below P-8)
- 14. Steam Generator Water Level--Low-Low
-< 2.0 seconds (Loop Stop Valves Open)
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- 15. Steam [Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE
- 16. Undervoltage-Reactor Coolant Famps
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- 17. Underfrequency-Reactor Coolant Pumps
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- Neutron detectors are exempt from response time testing.
Response
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electronic component in channel.
l BEAVER VALLEY - UNIT 2 3/4 3-8 (PrepcshoN
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ATTACHMENT B Safety Analysis Beaver Valley Power Station Unit No. 2 Proposed Techr ical Specification Change No. 27 Description of Amendment Reauest:
The proposed amendment would revise the Reactor Trip System specification on Beaver Valley Unit 2 by deleting.the trip function on Steam /Feedwater Flow Mismatch I
coupled with Low Steam Generator Water Level.
The criteria on line 14 on Table 2.2-1 in Technical Specification 2.2.1 and on line 15 on Table 3.3-1, 3.3-2, and 4.3-1 in Technical Specification 3.3.1.1 would be entirely removed.
The words "(Deleted)" would be inserted to maintain the current numbering format.
The bases for the Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level on Page B
2-6 along with the Note number 6 on Table 2.2-1 would be deleted since they are no longer required.
Discussio.D*
Background and Original Design Criteria The reactor trip on steam /feedwater flow mismatch coupled with low steam generator level (sometimes referred to as the low feedwater flow trip) was included in the original design of the-Reactor Trip System (RPS) on Beaver Valley Unit 2
to address potential control / protection interaction as discussed in General Design Criteria 24 in Appendix A of 10 CFR 50 and 10 CFR 50.55 a(h) which endorsen IEEE-279, 1971.
Control and protection system interaction addresses any mechanism where the protection system's ability to accomplish its safety function is adversely affected by a control system.
The mechanism may be physical, electrical or functional.
Each steam generator has three independent water level instrument channels which provide input to the RPS for a reactor trip on two out of three low-low levels.
(See Beaver Valley Unit 2 UFSAR Section 7.2.1.1.2 and Figure 7.2-1).
The Feedwater Control System, which controls the Feedwater Regulating Valve and hence feedwater flow into the steam generator also uses one of the three steam generator water level channels for input.
Therefore common instrument channels are used for both RPS and feedwater control, separated electrically by isolation devices.
This design assures that the plant will be controlled by the same measurements with which it is protected.
Such a
scheme precludes any sensor deviation between control and protection functions thereby reducing the likelihood of spurious reactor trips and reduces the task of channel calibration and maintenance.
, Attachment B (Continued)
Such a
design would present a difficulty in the control / protection interaction criteria assuming the steam /feedwater Iaismatch trip were not present.
Certain failures may be postulated which could negate a
particular water level channel and simultaneously cause undesirable control system action that requires subsequent protective action in order to prevent exceeding design safety limits.
For example, a postulated failure (high) of one steam generator water level channel would send a one out of three high level signal to RPS and, if being also used -for feedwater
- control, send a
high level signal to it thereby causing the -feedwater regulating valve to shut and reducing water level in the. steam generator.
For such a scenario, IEEE-279, Section 4.7.3 imposes the requirement for degradation by a second random failure.
The underlying logic is that the initial protection system failure is considered the initiating event for the transient,
- and, therefore, does not constitute the " single failure" IEEE-279 imposes on the protection system.
As such, an additional protection failure must be postulated to occur, and the protection system must continue to be capable of initiating the appropriate protective action.
The limiting single failure in this instance would be a failure (fail as-is) of one of the remaining two steam generator level channels.
This leaves only one operating channel which is insufficient to satisfy the 2/3 logic needed for a low-low steam generator water level reactor trip.
With the diverse steam /feedwater mismatch trip' function included in the
- design, the necessary protective action (i.e.,
reactor trip needed to meet safety limits) would still function since the mismatch trip occurs upon a one of two channel steam /feedwater mismatch coupled with a one of two channel low steam generator water level.
Hence adequate control / protection system interaction protection is afforded.
Addition of the Median Signal Selector (MSS)
A control / protection system interaction issue arose during the final licensing review of Beaver Valley Unit No. 2 (See Section 7.3.3.12 in NRC Safety Evaluation Report NUREG-1057 and Supplement SER No. 2 and No.
5).
The concern addressed the reverse of the postulated scenario described above.
A postulated failure low of one steam generator water level channel would cause the feedwater regulating valve to fully 'open and increase steam generator water level.
A second postulated failure (as-is) precludes the 2/3 turbine trip on high steam generator water level.
.; Attachment B (Continued) l l.
A median signal selector was added to the feedwater level control q
cystem design before Beaver Valley Unit 2 cormenced initial operation
-in 1987 (See UFSAR Section 7.2.2.2.3 and 7.2.2.3.5) to address the control / protection system interaction issue.
The prime objective of the signal selector is to prevent a failed instrument channel from causing a
disturbance in the controlled system which will initiate a plant. transient.
- Formerly, the Feedwater Control System received only.
a. single input channel of ' steam generator water level measurement; failure of that channel could cause adverse control system action.
With a signal selector, all three level measurement channels are input to the control system and compared by the signal selector.
The device selects the median signal for use by the control
- system, and control system action is then based on this signal.
By rejecting the high and low signals,.the control system is prevented from acting on any
- single, failed protection system instrument channel.
Since no adverse control system action may now result from a single, failed protection instrument channel, a second random protection system failure (as would otherwise be required by IEEE-279) need not be considered.
Signals resulting from a single failed high or low steam generator level measurement will be rejected for control purposes and, therefore, will not affect the system; the control and protection system interaction mechanism has been eliminated.
This design was approved by the NRC Staff in supplement 5
of NUREG-1057 as conforming to the requirements of Paragraph 4.7 of IEEE-279 and hence 10 CFR 50.55 a(h).
WCAP-11484, "Feedwater Control System Median Signal Selector" also confirmed the acceptability of i
the MSS.
The installation and use of the MSS also effectively precludes the control / protection system interaction addressed for a
decreasing steam generator water level scenario by the same cationale.
Thus the MSS can be substituted for the reactor trip on steam /feedwater flow mismatch to meet the control / protection system interaction criteria of IEEE-279.
The MSS design already approved and in use throughout the operation of Beaver Valley Unit 2
has demonstrated reliable operational readiness.
Although not part of the protection system, it has been designed to reduce the frequency of system failures through utilization of highly reliable components in a design that relies on a
minimum of additional equipment.
It should be noted that failure of the signal selector does not directly compromise the ability of the protection system to perform its safety related functions (i.e.,
failure of the MSS will not disable any protection channel).
Attachment B (Continued)
Furthermore, the design provides the capabi31ty for complete on-line MSS testing that provides unambiguous determination of credible system failures.
Monthly maintenance surveillance procedures (during Modes 1,
2,
& 3) test each MSS on each steam generator.
During the monthly surveillance for each steam generator level channel, the MSS is verified to pass the median level signal (1) with the channel under test operating
- normally, (2) with the channel under test adjusted artificially
- high, and (3) with the channel under test adjusted artificially low.
This test directly verifies that the MSS will not follow a
failed steam generator level signal.
Over the course of each month, all nine steam generator level channels and all three median signal selectors are tested.
The steam Generator Water Level Control
- Loops, which includes the MSS, are re-calibrated at least every 36 months.
The Main Feedwater Regulating Valve and Feedwater Bypass Control Valve on the affected steam generator are both placed in manual control during test eliminating the possibility of any adverse automatic feedwater control system interaction while testing.
The first operational application of the MSS in the nuclear industry was the Beaver Valley Unit 2 use of the MSS in its steam generator water level instrumentation in 1987.
No failures on the MSS have been experienced to date.
The combination of demonstrated performance coupled with minimizing the probability of failure and the likelihood of any future failure remaining undetected for an extended period of time through continued frequent testing provide the necessary degree of confidence relative to MSS operational readiness and reliability.
In addition to eliminating control / protection system interaction
- concerns, the reliability of the feedwater control system is increased with the use of the MSS since it eliminates potential transients which could result from water level channel instrument channel single failures.
Impact On Safety Analyses No reference to the steam /feedwater mismatch trip is indicated in Table 15.0-4 and Table 15.0-6 which lists all reactor trip functions assumed in the accident analyses in Chapter 15 of the Beaver Valley Unit 2
The Loss of Heat Sink Analyses, Loss of Normal Feedwater Flow analyses in Section 15.2.7 and Feedwater System Pipe Break analyses in Section 15.2.7, both state that protection is provided by a
reactor trip on low-low steam generator water level.
No credit was taken or needed for a steam /feedwater flow match trip in these two UFSAR sections.
It is conservative to use the low-low water level trip since the steam /feedwater mismatch trip would, in many
- cases, be received earlier than the low-low water level trip.
Table 15.2-1 which provides the calculated sequence of events for the Loss of Normal Feedwater Flow and Feedwater System Pipe Break safety analyses also does not make any reference to the steam /feedwater mismatch trip.
No other Chapter 15 analyses take credit for a reactor trip function on low secondary water level for protection.
. Attcchm:nt B (Continued)
Chapter 7
(Section 7.2.2.2.3) of the UFSAR describes safety analyses for control / protection system interaction.
As explained earlier, the initial design used the steam /feedwater mismatch trip if the initiating transient of an event were postulated to be a failure of one steam generator water level channel with a single failure being the second steam generator water level channel, rendering the 2/3 low-low water level trip inoperable.
However, the Median Signal Selector now provides sufficient substitute protection for this
-event.
Table 7.2-4, No. 16, in the UFSAR indicates that the steam /feedwater flow-mismatch reactor trip is pertinent to the Loss of Normal Feedwater Flow analyses (15.2.7) and to the Feedwater System Pipe Break analyses (15.2.8).
As discussed, it is non-conservative to use or take credit for this reactor trip in Section 15.2.7 or 15.2.8 safety analyses.
Thun this reference to Section 15.2.7 and 15.2.8 in Table 7.2-4 only indicates analyses where it could potentially be used.
Table 7.2-4 goes on to indicate that a Technical Specification is not required because the steam /feedwater mismatch trip is not assumed to function in any accident analyses addressed in Chapter 15.
Thus since no credit is taken for the steam /feedwater flow mismatch reactor trip in Chapter 15 and the Median Signal Selector provides a sufficient substitute against an adverse control / protection system interaction, there would be no change nor any adverse effect to any safety analyses addressed in the UFSAR upon removal of the steam /feedwater flow mismatch trip function for the Beaver Valley Unit 2 design.
Supplemental Benefits Removal of the steam /feedwater flow mismatch reactor trip provides many additional significant benefits.
The most significant contribution would be.the reduced challenges to the overall plant safety systems by eliminating potential spurious and inadvertent reactor trips from this trip function.
Any unwarranted use of the Reactor Trip System challenges numerous safety systems and puts the plant through additional, unnecessary transients.
Beaver Valley Unit 1,
which has been operating since 1976 and has a similar Reactor Protection System
- design, has already experienced one unnecessary reactor trip and four inadvertent reactor trips during calibration / surveillance activities on the steam /feedwater mismatch trip function.
Beaver Valley Unit 1 and Unit 2 have had eleven and three other steam /feedwater mismatch trips, respectively; a couple of which conceivably could have avoided the low-low steam generator water level trip if the steam /feedwater mismatch trip had not been operable because of an improved steam generator level operating margin (low level trip setpoint is higher than the low-low setpoint),
l thereby eliminating another major contributor to plant trips.
Deletion of this now unnecessary trip function will actually increase the safety level of Beaver Valley Unit 2
by removing these unnecessary challenges to the safety systems.
_____.--__.m._._m-.__--_.___.-m__.____
o
, httachm:nt B (Continund) i Elimination of the steam /feedwater mismatch trip function also reduces the work level and system knowledge / interaction complexity level of the plant.
Removal of this trip function will lower the required surveillance / maintenance work level needed for the Reactor Protection System and will reduce the active systems which can potentially interact with all other plant activities and training.
Preceding Events Beaver Valley Unit 2
has already had the application of the Median Signal Selector on the steam generator water level channels approved i
for use as adequate criteria to address any adverse control / protection system interaction as per NUREG-1057, Safety Evaluation Report Supplement No.
5.
Recently, the Prairie Island nuclear power station also incorporated the MSS into their steam generator water level instrumentation system similar to the Beaver Valley Unit 2
design.
Prairie Island, recognizing the increased level of safety obtained by the elimination of the steam /feedwater flow mismatch reactor trip
- function, requested and received staff approval to delete this reactor trip function.
This trip function deletion has also been identified by the Westinghouse owners Group Trip Reduction Assessment Program as one upgrade which would help address feedwater related trips (the leading cause of reactor trips in Westinghouse NSSS plants) and hence help reduce the reactor trip rate.
Conclusion i
The Duquesne Light Company staff at the Beaver Valley Power Station are committed to constantly improving the reliability, safety and i
availability of the plant.
Elimination of the steam /feedwater flow I
mismatch reactor trip coincident with low steam generator water level is acceptable since the Median Signal Selector (already installed at Beaver Valley Unit 2) provides sufficient substitute protection i
against adverse control / protection system interaction, which was the basis for including this trip function in the original design bases,
)
and no credit is taken for the steam /feedwater mismatch trip in any Chapter 15 safety analyses.
Elimination of this now unnecessary j
reactor trip function will actually increase the level of plant safety due to the reduction in potential plant safety system l
challenges, reduction in plant surveillance / calibration activity, and I
the human factor benefit from the reduction in plant complexity.
i
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?.y o ATTACHMENT C No Significant Hazard Evaluation Proposed Technical Specification Change Unit No. 2 - Chance No. 27 Basis for Proposed No Significant Hazards Consideration Determination:
The Commission has provided standards for determining whether a
significant hazards consideration exists (10 CFR 50.92(c)).
A proposed amendment to an operating license for a
facility involves no significant hazards consideration if a
operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated:
(2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
The proposed changes does not involve a
significant hazards consideration because:
1.
This change will not involve a
significant increase in the probability or consequences of an accident previously evaluated because
- credit, for the steam /feedwater mismatch reactor trip coincident with low steam generator water level, was not taken in any Chapter 15 safety analyses and the Median Signal Selector provides sufficient substitute protection against any adverse control / protection system interaction.
The probability for some previously evaluated accidents such as inadvertent trips is reduced due to the elimination of this reactor trip function.
Therefore no significant increase in the consequences of any accident previously evaluated results from this change.
2.
This change would not create the possibility of a
new or different kind of accident from any accident previously evaluated.
There is no new or different kind of accident because there will be no new equipment or systems provided.
This change removes an existing plant function which has been shown to be unnecessary as well as not being credited for in the safety analyses.
Thus there will be no changes to any previously evaluated accident.
3.
This change would not involve a significant reduction in a margin of safety.
The proposed Technical Specification will actually I
enhance safe operation eince the continued use of the steam /feedwater mismatch trip function has been shown to be unnecessary and its removal decreases the challenges to the plant safety
- systems, decreases the plant surveillance / maintenance
- activity, and reduces the plant complexity; all resulting in a l'
reduction in the potential of unnecessary safety system challenges.
Based on the above considerations, it is proposed to characterize the changes as involving no significant hazards considerations.
1 1
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