ML19209B043
| ML19209B043 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/17/1979 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| NUDOCS 7910090035 | |
| Download: ML19209B043 (13) | |
Text
T'lC WISCONSIN Electnc eom convr 231 W. MICHIGAN, P.o. BOX 2046. MILWAUKEE WI 53201 September 17, 1979 Mr. James G. Keppler, Director Office of Inspection and Enforcement Region III U. S. NUCLEAR REGULATORY COMMISSION 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
DOCKET NOS. 50-266 AND 50-301 RESPONSE TO IE BULLETIN NO. 79-21 POINT BEACH HUCLEAR PLANT, UNITS 1 AND 2 Attached are the responses to IE Bulletin No. 79-21, " Temperature Effects on Level Measurements". The attached information responds, item by item, to the four action areas listed in the bulletin.
Occurrence of a high energy line break inside containment, followed immediately by the uniform heatup of the reference leg of any level measuring system to te,,eratures near the maximum containment temperature for such accidents, is a very remote possibility.
Furthermore, we do not believe that the possible effect of containment temperature on indicated water levels inside containment presents a " potential substantial safety hazard". We have, however, performed our reviews and analyses assuming the worst potential temperature e ffects.
Very truly yours, c, -
C. W. Fay, Director Nuclear Power Department Attachments Copy to: Di rector Office of Inspection and Enforcement Division of Reactor Operations Inspection 791000003 5
$f.9 N 1114 268
ATTACHMErlT RESP 0!iSE TO IE BULLETIf179-21 Item 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 syst;ms that are so employed; a description of the type of reference leg shall be included, i.e., open column or sealed reference leg.
RESP 0!1SE After a review of all liquid level measuring systems within containment, the following systems, described below, were determined to initiate safety actions or provide information required for post-accident monitoring:
1.
Steam Generator flarrow-Range Level Each steam generator (S/G) has three narrow-range level measuring channels which measure downcomer liquid level above the top ele-vation of the S/G U-tubes.
Each system consists of one Foxboro Model flo. 613 HM " force-balance" differential-pressure transmitter.
The transmitter is connected to the steam generator downcomer on the variable leg side and is connected to an open-column reference leg on the other side.
The outside of the reference leg pipe is exposed to containment atmosphere and has a condensing pot on the top which is connected to the steam generator steam space.
Signals from these transmitters are used for remote level inc" cation, level control, alarms, and safety trips.
2.
Steam Generator Wide-Range Level Each steam generator has one wide-range level measuring system.
It is identical to the S/G narrow-range system except that the larger Page 1 of 13 1114 289
t level cnan covers downcomer liquid level from just above the S/G tube sheet to above the moisture separators.
Signals are used for remote level indication only.
3.
Pressurizer Level The pressurizer has three hot-calibrated and one cold-calibrated liquid level measuring systems.
Each system consists of one Foxboro Model No. 613 HM " force-balance" differential-pressure transmitter.
The transmitter is connected to the pressurizer near the bottom on the variable-leg side.
It is connected to a sealed reference leg on the other side.
The outside of the reference leg pipe is exposed to the containment atmosphere and has a bellows-type seal near the top. A water seal is maintained on the top of the bellows by a condensing pot connected to the pressurizer steam space. The signals from the hot-calibrated transmitters are used for remote level indication, level alarms, level control, and safety trips. The cold-calibrated transmitter signals are used for remote level indication only.
4.
Containment Sump B Levc!
Containment Sump B has two magnetically-actuated float switch assemblies which energize lights in the control room at containment sump water levels of 3 and 7 feet, and 5 and 9 feet, respectively.
The indication is used to confirm sufficient water inventory in the sump prior to shifting the Emergency Core Cooling Systems to the recirculation mode following safety injection.
This type of measuring system has no reference leg.
Page 2 of 13 j j j 4 2()Q
Item 2 On those systems described in Item 1 above, evaluate the effect of post-accident ambient temperatures on the indicated water level to determine any change in indicated level relative to actual water level.
This evaluation must include other sources of error includiag the effects of varying fluid pressure and flashing of reference leg to steam on the water level measurements. The results of this evaluation should be presented in a tabular form similar to Tables 1 and 2 of Enclosure 1.
RESP 0f1SE The effect of post-accident 3mbient temperatures on the indicated water level for the systems described previously is presented in Tables 1 through 4.
ihe containment sump B level system is not affected significantly by ambient temperatures so no table was provided for it.
The possibility of reference leg flashing on indicated pressurizer and steam generator level, during high energy line break accidents, was analyzed.
Reference leg flashing on the pressurizer is not significant since a sealed reference leg is used.
During a major steam line or feedwater line break accident, reference leg flashing, on the c.Tiected steam generator only, may occur late in the blowdown phase. The affe;ted steam generator would boil dry in the normal course of the accident, so leval indication late in the transient is not required.
Steam generator pressure during a loss of coolant accident is expected to remain high enough to prevent reference leg flashing during that accident.
In conclusion, reference leg flashing is not a concern on any high energy line break.
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Page 3 of 13
souie i Correction Required to Indicated Steam Generator Narrow-Range Level for Reference Leg Heatup Effects
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Due to Post-Accident Containment Temperature i
Correction Required Correction Required Correction Required to Low-Low Level Trip to Minimum Allowed Post-to Maximum Allowed Post-Reference Leg Temperature Setpoint (Before Reactor Trip)
Accident Indicated Level
- Accident Indicated Level **
( F)
(% of span)
(% of span)
(% of span)
S/G Pressure 800 500 800 500 200 800 500 200 (psig) 68
-1
-2
-1
-1 0
-1 3
8 100 0
-1 150 2
0 200 4
3 250 7
6 300 10 9
10 10 11 10 14 19 350 14 13 400 18 17 Basis:
Level Calibration Steam Generator Pressure = 800 psig (saturated water)
Reference Leg Calibration Temperature = 100 F 20% of level span
- 70% of level span
Table 2 Correction Required to Indicated Steam Gererator Wide-Range Level for Reference leg Heatup Effects Due to Post-Accident Containment Temperature Correction Required to Cerrection Required to Minimum Allowed Post-Maximum Allowed Post-Reference Leg Tem, erature Accident Indicated Level
- Accident Indicated Level **
( F)
(% of span)
(% of span)
S/G Pressure 800 500 l200 800 500 200 (psig) 68
-18
-15
-10
-21
-17
-12 300
-10
-7
-3
-12
-9
-4
?
l r
8 asis:
Level Calibration Steam Generator Pressure = 0 psig (subcooled water at 68 F)
Reference leg Calibration Temperature = 68 F 80% of level span
- 90% of level span M
emuse Nc u
Table 3 Correction Required to Indicated Pressurizer Hot-Calibrated Level for Reference Leg Heatup Due to Post-Accident Containment Temperature Correction Required to Correction Required to Minimum Allowed Post-Maximum Allowed Post-Pressurizer Pressure Accident Indicated Level
- Accident Indicated Level **
(psia)
(% of span)
(% of span)
Reference Leg Temperature 68 300 a
300 (OF)
^
2250
-1 15
-1 15 2000
-2 13 2
18 1750
-3 13 6
23 1500
-4 13 10 26 O
l 1000
-6 11 17 32(I) 500
-5 12 26 38(I)
Basis:
Level Calibration Pressurizer Pressure = 2250 psia (saturated water)
Reference Leg Calibration Temperature = 100 F 20% of level span
- 70% of level span (1) Indicated level cannot exceed 100% of level span N<
4
Table 4 Correction Required to Indicated Pressurizer Cold-Calibrated Level for Reference Leg Heatup Due to Post-Accident Containment Temperature Correction Required to Correction Required to Minimum Allowed Post-Maximum Allowed Post-Pressurizer Pressure Accident Indicated Level
- Accident Indicated Level **
(psia)
(% of span)
(%ofspan)
Reference Leg Temperature ( F) 68 300 68 300 2250 0
7
-26
-18 2000
-2 7
-24
-16 1750
-2 6
-22
-15 1500
-3 6
-21
-13 1000
-3 5
-17
-9 500
-2 6
-13
-5 a
?,
Basis:
Level Calibration Pressurizer P/ essure = 14.7 psia (subcooled water at 68 F)
Reference Leg Calibration Temperature = 680F 20% of level span
- 70% of level span M
M M
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Item 3 Review all safety and control setpoints derived from level signals to verify that the setpoints wil' initiate the action required by the plant safety analyses throughout ti.
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 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 timetable for implementation.
RESPOUSE A review of all safety and control setpoints derived from the level measurement systems described earlier showed that the only setpoint which could be adjusted for conservatism was the steam generator "10-10 level reactor trip" and
" auxiliary feedwater system startup" setpoint.
This setpoint provides primary protection for loss ~of feedwater accidents (including pipe breaks) and is used as a backup signal for auxiliary feedwater system startup on other high energy line breaks inside containment. No.other safety or control setpoints initiate any action required by the safety analyses for high energy line break accidents.
Heatup of the reference leg for steam generator level indication following a high energy pipe break in containment requires some elaboration.
If a loss of feedwater due to a small pipe brea'K inside containment were to occur, the plant operators would be alerted by increased containment humidity, increased sump level alanns, and increased containment pressure.
The main feer
_+.er pumps would continue supplying water to the steam generators, albeit the indicated level in one steam generator could be slightly in error. A "high feedwater flow" alarm could be annunciated, depending upon the size of the leak.
Safety injection and reactor trip is initiated at a containment pressure of 6 psig or less and would terminate power operation if operating personnel had not done so. The saturation temperature for a homogeneous air-steam Page 8 of 13
] ] ] 4 [J6
mixture at this pressure is aoout 170 F, well below the 300 F temperature assumed in our analysis.
The physical configuration of the steam generators at Point Beach Nuclear Plant tends to reduce the impact of a major feedwater line break.
The feedwater lina penetrates the steam generator above the biological shield, eliminating any
" compartmental" or " chimney" effect of hot steam rising up to affect the reference legs.
In addition, the level taps are on the opposite side of the steam generator from the feedwater line preventing direct impingement of steam on the reference legs. Therefore, the rapid water level decrease resulting from this accident would trip the reactor on lo-lo level before the steam could significantly bias the indicated level.
The effects of a steam line break upon the reference legs have also been evaluated. The results oi in assumed small leak are similar to that of a feedwater line leak. The primary protection for a large steam line break is a high containment pressure signal backed up by a low steam line pressure signal.
The steam generator lo-lo level reactor trip and auxiliary feedwater systen startup setpoint may be raised to account for possible reference leg heatup effects following a feedwater line break inside containment.
The maximum possible containment bulk temperature which could occur, prior to receiving a safety injection signal from high containment pressure, is 170 F.
A temperature of 300 F corresponds to the temperature of superheated steam in containment resulting from the release of steam generator secondary-side saturated steam at the maximum expected enthalpy. This situation is not plausible because the containmeni and equipment in it serve as a heat sink to
(.,
reduce the superheat. The new setpoint for lo-lo steam generator level is conservatively calculated as shown below:
1114 297 Page 9 of 13
Narrow-Range BASIS Level Span Safety Limit Used in Accident Analyses 0%
Process Measurement Accuracy 2%
Instrument and Bistable Error 1%
Reference Leg Heatup Maximum Error 10%
(for a reference leg temperature of 300 F)
New Setpoint 13%
The new setpoint will increase the plant's vulnerebility to inadvertent reactor trips and auxiliary feedwater system startups. Although we believe there is minimal positive benefit and there are negative aspects resulting from the set-point change, the setpoints will be changed.
The setpoint adjustment to a new value greater than or equal to 13% of narrow-range level span will be accomplished on Unit 2 by September 30, 1979, and on Unit 1 prior to startup following the refueling outage scheduled to begin on October 5,1979.
The safety and control setpoints derived from steam generator narrow-range level and pressurizer level are shown in Tables 5 and 6, respectively.
The other level measurement systems described earlier do not have safety and control functions other than post-accident monitoring.
Page 10 of 13
TABLE 5 STEAM GEllERATOR f1 ARROW-RAllGE LEVEL C0llTROL Ar4D SAFETY SETP0lllTS Function Level Span Low-Low Level Reactor Trip and Auxiliary Feed 10%
System Startup Setpoint Low Level Reactor Trip Setpoint (in coincidence 20%
with feedwater flow / steam flow mismatch)
Level Program (low limit for 0% power) 39%
Level Deviation Alarm Setpoint
+3% from level program Level Program (high limit for >20% power) 52%
High-Level Override for Feedwater Valve Closure, 70%
Turbine Trip, and Main Feed Pump Trip Page 11 of 13 ll}4
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Pressurizer Hot-Calibrated Level Control and Safety Setpoints Function Level Span Low-Low Level Heater Cutout, Letdown 12%
Isolation Setpoint Level Program (low limit for Tavg.1 5470F) 20%
Level Deviation Alann Setpoint
+ 5% from Tevel program Level Controller Gain 1% charging pump speed /1%
level deviation from level program Level Program (high limit for Tavg 3.570 F) 45.8%
High-Level Deviation Heaters-On Setpoint 5% above level program High-Level Alarm Setpoint 70%
High-Level Reactor Trip Setpoint 90%
1 Page 12 of 13 1114 300
Item 4 Review and revise, as necessary, emergency procedures to include specific infor-mation obtained from the review and evaluation of Itens 1, 2 and 3 to ensure that the operators are instructed on the potential for and magnitude 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 revi-sions and a completion date for operator training on the revisions.
RESPONSE
The operators at Point Beach Nuclear Plant have been instructed on the general effect of post-accident ambient temperatures on level instrument readings following a high-energy line break inside containment; however, slight modi-fications to emergency procedures and/or standing orders would provide more specific information to the operator.
These revisions will be made on a schedule compatible with the schedule of procedure modifications. required in response to IE Bulletin No.79-06C, " Nuclear Incident at Three Mile Island -
Supplement".
The revisions will be completed by December 31, 1979.
Implemen-tation of the revised procedures and completion of operator training will require an additional one to two months.
If NRC review of the procedure modifications is required, the implementation and operator training will not begin until the NRC review is completed.
1114 301 Page 13 of 13
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