ML19269F495
| ML19269F495 | |
| Person / Time | |
|---|---|
| Site: | San Onofre, Trojan  |
| Issue date: | 09/14/1979 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| Shared Package | |
| ML13330A826 | List: |
| References | |
| IEB-79-21, NUDOCS 7912210162 | |
| Download: ML19269F495 (15) | |
Text
.
Portland General ElectricCcmpany C
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"t September 14, 1979 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. R. H. Engelken, Director Nuclear Regulatory Commission Region V Suite 202, Walnut Creek Plaza 1990 N. California Blvd.
Walnut Creek, CA 94596
Dear Sir:
Attached is the PGE response to IE Bulletion 79-21 that was transmitted in your letter dated August 13, 1979. This Bulle-tin required a PGE evaluation of the effects of increased Containment temperature on liquid level measuring systems inside Containment.
Sincerely, Yf f
c/ W :h
/
C. Goodwin, Jr.
Assistant Vice President Thermal Plant Operation and Maintenance CG/CJP/4kk6A25
/.ttachment c:
Mr. Lynn Frank, Director w/ attach State of Oregon Department of Energy
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.c PGE RESPONSE TO IE BULLETIN 79-21 1.
Review the liquid level measuring systems within Containment to determine if the signals are used to initiate safety actions or are used to provide post-accident monitoring information.
Provide a description of systems that are so employed; a description of the type of reference leg shall be included, i.e.,
open column or sealed reference leg.
PCE Response The only liquid level measuring systems inside Containment that are used for initiating safety actions are the steam generator narrow-ronge water level and the pressurizer water level.
However, as indicated in attached Table 1, the only safety actions for which credit is taken in the Trojan safety analyses following a high-energy line rupture inside Containment are a reactor trip snd auxiliary feedwater initiation on low-low steam generator water level; therefore, only the modification of this setpoint must be considered to accommodate the effect of refer-ence leg heatup. The level indications listed in Table 1 all provide post-accident monitoring information to some degree, and the effect of increased ambient temperatures has been evaluated to assure that improper operator actions would not be taken as the result of possibly erroneous level indications.
Steam Generator Narrow-Range Water Level The steam generator narrow-range level instrument is an open re fe rence leg, condensing pot design that provides a level signal proportional to the differential pressure developed between the water column hel,.its in the steam generator down-comer and in the reference leg.
A low-low steam generator level initiates reactor trip and Auxiliary Feedwater System operation as the primary protection against a feedwater line rup tu re.
2167 078
Additional protective actions based on steam generator water level are a turbine trip and feedwater isolation on a high-high level and a reactor trip on low level in coincidence with steam flow / feed flow mismatch. However, no credit for these functions is taken following a high-energy line break inside Containment.
Steam gene r r water level is used as an indicator of adequate secondary inventory for post-accident heat removal.
Pressurizer Water Level The pressurizer water level instrument is a sealed reference leg, condensing pot design that provides a level signal pro-portional to the differential pressure developed between the water column heights in the pressurizer anc
'n the reference leg.
The only protective action, based on pressurizer level, is a reactor trip on high water level; however, no credit is taken for this function following a high-energy line break inside Containment. The pressurizer water level is used as an indicator of post-accident reactor coolant inventory.
Accumulators The accumulator water level instrument is a sealed reference leg design that provides a level signal proportional to the differ-ential pressure developed between the sater column heights in the accumulator and the reference leg.
The accumulator levels are used as indicators that the accumulators have discharged into the Reactor Coolant System following a large LOCA in which the system dep;essurizes to below accumulator pressure.
Pressurizer Relief Tank The pressurizer relief tank level instrumer t is an open refer-ence leg design that provides a level signal proportional to the differential pressure developed betwee., the water column heights in the pressurizer relief tank and in the reference leg.
The 2167 079
pressurizer relief tark level is used following an accident as an indication of discharge from the pressurizer power-operated relief valves and safety valves.
Reactor Coolant Drain Tank The reactor coolant drain tank (RCDT) level instrument is an open reference leg design that provides a level signal propor-tional to the differential pressure developed between the water column heights in the tank and in the reference leg.
The RCDT level instrument provides starting and stopping signals to pump the contents of the RCDT to the clean waste receiver tanks in the Auxiliary Building.
Contcinment Sumps The Containment sump level instruments are magnetic float switch mechanisms.
Sump level is monitored during normal operations to provide indication of water leakage inside Containment.
2.
On those systems described in Item 1 above, evaluate the effect of post-accident ambient temperatures on the indicated water leve] to determine any change in indicated level relative to actual watet level. This evaluation must include other sources of error including the effects of varying fluid pressure and flashing of reference leg to steam on the water level measure-ments.
The results of this evaluation should be presented in a tabular form similar to Tables 1 and 2 of Enclosure 1.
PCE Response For level measuring systems with reference legs exposed to the Containment atmosphere, high-energy line breaks inside Contain-ment can result in heatup of the reference legs.
Increased reference leg water column temperature will result in a decrease of the water column density with a consequent apparent increase 2167 080
a:
in the indicated water level.
In addition to the reference leg density change under subcooled conditions, boiling could con-ceivably occur in the reference leg folicwing depressurization of a steam generator with high Containment temperature. This combination of conditions could caly occur follcwing a steam line or feedwater line rupture inside Containment.
If such boiling were to occur, it could cause a major bias in indicated level for a short time period, in the extreme case indicating 100 percent level when the vessel is actually empty. A bias in indicated water level may also be introduced by changes in pressurizer or steam generator pressure, due to changes in the densities of the saturated water and steam within these vessels.
While prediction of the effects of rapid depressurization requires complex calculations for each specific case, the bias which would exist at low power under quiescent conditions has been estimated by Wet -inghouse using a conservative simplified analysis model and assumptions that are appropriate for Trojan.
Steam Generator Narrow-Range Water Level Credit is taken in the Trojan feed line break accident analysis for a reactor trip and auxiliary feedwater initiation on low-low steam generator water level. As far as the trip setpoint is concerned, because large steam generator pressure changes are not expected before the trip, only the reference leg heatup effects need to be considered and not the effects of system pressure changes.
The calculated error in the indicated steam generator level as a function of reference leg temperature, based on calibration conditions of 1000 psia steam generator pressure and 90*F Containment ambient temperature, is summarized in Tabl 2.
Calculations of the combined effects of the longer-term density changes in the steam gene.cator and its reference leg during depressurization and of reference leg heatup are summarized in attached Tabic 3.
This table provides correction values for 2167 081
+
minimum and maximum allowable indicated levels to assure the actual steam generator water level remains in the allowable range during the post-accident recovery period.
Reference leg boiling is a remote possibility that could occur following a secondary high-energy line rupture. Thi; could cause large errors in indicated level for short periods of time.
Pressurizer Water Level No credit is taken for a pressurizer level-derived reactor trip following a high-energy line rupture inside Containment. Thus, the trip setpoint need not be revised to compensate for environ-mental effects.
Effects of pressurizer reference leg heatup and density changes during depressurization are summarized in attached Figure 1 and 2, illustrating the impact on post-accident monitoring. These curves are based on calibration conditions of 90*F Containment temperature and 2250 psia system pressure.
Reference leg boiling due to a combination of rapid system depressurization and reference leg heatup is a small effect for the pressurizer due to the sealed reference leg design.
This effect has been previously considered in the Trojan FSAR (Sec-tion 7.2.2.3.4) and would lead to a level error of at most 1 ft, or about 2 percent of span, which is negligible for post-accident monitoring purposes.
Accumulators The accumulators are passive devices that are discharged into the RCS if the system de: ressurizes below the accumulator nitrogen pressure of 600 psig. The major function of the level instrumentation is to provide assurance during operation that 2167 082 the design volume of water is available for injection into the RCS in the event of a LOCA. As far as post-accident monitoring is concerned, the accumulator levels do not provide essential information to the operator during the accident recovery and, therefor it is judged that no further analysis of the effect of reference leg heatup is necessary for these ir struments.
Pressurizer Relief Tank The pressurizer relief tank receives discharge from the pres-surizer safety valves and power-operated relief valves. The water level is maintained and monitored in this tank during normal operation to provide quenching of the steam discharge.
During the post-accident period, the level provides indication of the amount of relief or safety valve discharge that has occurrad. Other indications such as charging system inflow and relief valve outlet piping temperature are available that provide the same information, and no inappropriate operator actions would result from an indicated level higher than the actual level in the PRT.
Therefore, it is judged that no further analysis is necessary of the effect of reference leg heatup on this instrument.
Reactor Coolant Drain Tank The RCDT receives discharge of reactor coolant from a number of sources. The level instrument contrels automatic pumping of the RCDT contents to the clean waste receiver tanks outside of the Containment. The potential consequences of an erroneously high indicated RCDT level would be to pump the RCDT before the intended setpoint was reached or to continue pumping the RCDT after the setpoint intended for pump shutoff.
Neither of these consequences presents a significant safety hazard, although the automatic transfer of potentially radioactive RCDT fluid out of the Containment could occur sooner than intended.
A high-energy line break inside Containment that leads to the adverse environ-2167 (183
ment of concern would automatically isolate the discharge line from the RCDT.
Therefore, it is judged that no further analysis is necessary of the ef fects of reference leg heatup on this Instrumt,t.
Containment Sumps The sump level instruments are magnetic float switch mechanisms that have not been qualified for service in the post-LOCA Containment environment. These instruments provide indication of water leakage inside Containment during normal operation.
The Containment sumps (but not the recirculation sump) are automatically pumped to the dirty wa..e drain tank outside Containment on a high level signal.
The consequences of an erroneous level indication do not pose a significant safety hazard, since a high-energy line break inside Containmenc that creates the adverse environment of concern would lead to auto-matic isolation of the sump discharge line. Furthermore, as a followup of our review of the TMI-2 accident, we are presently develop'ng a design change to the sump discharge system that would provide additional protecticn against inadvertent trans-fers of the sump contents. This change, to be completed prior to the Cycle 3 startup, involves installation of a radiation monitor in the sump discharge line.
Therefore, it is judged that no further analysis is neccesary of the effects of post-accident environment on the sump level switches.
3.
Keview all safety and control setpoints derived from level signals to verify that the setpoints will initiate the action required by the Plant safety analyses throughout the range of ambient temperatures encountered by the instrumentation, including accident temperatures.
Provide a listing of these setpoints.
If the above reviews and evaluations require a revision of setpoints to ensure safe operation, provide a description of 2t6/ 084
the corrective action and the date the action was completed.
If any corrective action is temporary, submit a description of the proposed final corrective action and a time table for implementation.
PCE Response The only safety action derived from a level signal, for which credit is taken following a high-energy line break inside Containment is the reactor trip and auxiliary feedwater initia-tion on low-low steam generator water level following a feed line break.
During a high-energy line break inside Containment, both the Containment ambient temperature and ambient pressure will increase. Within a few seconds, depending on the break size, the high Concainment pressure trip setpoint (5 psig) would be reached providing acceptable backup protection to the low-low steam generator Icvel trip.
In the analyses performed for the 1971 Westinghouse instrumentation environmental qualification program, in which Trojan is a participant, the high Containment pressure setpoint was reached prior to the Containment tempe ra-ture reaching 280 F.
A change in reference Icg temperature from the calibration condition of 90*F to 280*F would result in an indicated level which is 10 percent higher than actual, as shown in the respense to Question 2.
Therefore, the previous low-low steam generator level trip setpoint has been raised f rom 5 per-cent to 15 percent as an interim measure to accommodate the possible ef fect of reference leg heatup. This change has previously been identified to you in letters dated June 26 and July 5, 1979. A combination of the revised low-low level trip setpoint and the backup protection afforded by the high Con-tainment pressure trip ensures that the feed line break criteria stated in the Trojan FSAR continue to be met.
Westinghouse is currently evaluating possible steam generator level instrument design changes to compensate for reference leg heatup.
It is possible that these revised designs would permit returning the setpoint to its previous value.
2167 085
_8_
4.
Review and revise. 2s necessary, emergency procedures to include specific information obtained from the review evaluation of Items 1, 2 and 3 to ensure that the operators are instructed on the potential for and magnitule of erroneous level signals. All tables, curves, or correction factors that would be applied to post-accident monitors should be readily available to the operator.
If revisions to procedures are required, provide a completion date for the revisions and a completion date for operator training on the revisions.
PCE Response Based on the Westinghouse and PGE reviews described above, curves showing the range of correction factors for steam genera-tor water 1cvel indication at various Containment temperatures, up to 280*F, were developed. These curves were placed in the Trojan Control Room Operating Curves and Tables Reference Manual.
In addition, the appropriate control room indicators have been marked to caution the operators regarding accuracy of the indicator at elevated Containment temperatures.
Operators have been made aware of the necessity for, and the proper use of, these correction factors.
b 00b
_9_
CJP/4kk6A26
TABLE 1 LIQUID LEVEL MEASURING SYSTEMS INSIDE CONTAINMENT THAT INITIATE SAFETY ACTIONS OR PAh[b] INFORMATION Liquid Level Reference Initiates Provides Measuring System Leg Design Safety Actions, PAM Information 1.
Steam Generator Open Yes Yes Narrow-Range Water Level I
2.
Pressurizer Water Scaled No Yes Level 3.
Acc umulator s Scaled No Yes 4.
Pressurizer Open No Yes Relief Tank 5
Reactor Coolant Open No Yes Drain Tank II II 6.
Containment Sumps N/A No Yes
[a] Safety actions for which credit is taken following a HELB inside Containment.
[b] Post-accident monitoring.
[c] Two Containment sumps and one Containment recirculation sump.
[d] Containment sump level is measured by a float switch mechanism.
2167 087 CJP/4kk6BS
TABLE 2 CORRECTION TO INDICATED STEAM GENERATOR WATER LEVEL FOR REFERENCE LEG llEATUP EFFECTS DUE TO POST-ACCIDENT CONTAINMENT TEMPERATURE (BEFORE REACTOR TRIP)
Maximum Containment Correction to Temperature Reached Steam Generator Before Reactor Trip Level
(*F)
(% of Span)
- 90 0
200 4
280 10 320 13 400 20 Basis:
Level calibration pressure f 1000 psia Reference leg calibration temperature > 90*F Height of reference leg i 1.1 x level span.
2167 088 CJP/4kk6B6
TABLE 3 CORRECTIONS TO ALLOWABLE INDICATED STEAM GENERATOR FATER LEVEL FOR REFERENCE LEG IIEATUP AND PRESSURE CllANGES FOLLOWING A lilGil-ENERGY LINE BREAK, 10 ASSURE TilAT TRUE LEVEL IS BETWEEN Tile LEVEL TAPS Corrections to Corrections to Containment Minimum Allowed Maximum Allowed Tempe ra ture Indicated Level Indicated Level
(*F)
(% of Span)
(% of Span) 90
+1
-4 200
+6
-4 280
+11
-4 320
+14
-4 400
+21
-4 Basis:
Level calibration pressure j[ 1000 psia Reference leg calibration temperature 2; 90*F lieight of reference leg < l.1 x level span Pressure }; 50 psia Pressure j[ 200 psi + calibration pressure Boiling in the reference leg is not assumed.
2167 089 CJP/4kk6B7
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