ML19347C579
| ML19347C579 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 12/09/1980 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Abel J COMMONWEALTH EDISON CO. |
| References | |
| NUDOCS 8012310170 | |
| Download: ML19347C579 (10) | |
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UNITED STATES f
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fj NUCLEAR REGULATORY COMMISSION
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WASHING TCN. O C. 20555
=, es.aj j DEC 9 1980 Docket Nos. : 50-373 O
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Mr. J. S. Abel JE'~
N Director of Nuclear Licensing
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Conronwealth Edison Company mgj k
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Dear Mr. Abel:
Subject:
Requests For Additional Information - LaSalle County Station As a result of our review of your application for operating licenses for LaSalle, we find that we need additional information in the areas of Mechanical Engineering and Instrumentation and Control.
The specific information required is provided in the Enclosure.
If you desire any discussion or clarification of the information requested, please contact A. Bournia, Licensing Project Manager, (301) 492-7200.
Sincerely, e4 Robert L. Tedesco, Assistant Director for Licensing Division of Licensing l
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Enclosure:
l Requests For Additional Information ces w/ enclosures:
1 See next page
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I Mr. J. S. Abel Director of Nuclear Licensing Commonwealth Edison Company Post Office Sox 767 Chicago, Illinois 60690 cc: Philip P. Steptoe, Esq.
Suite 4200 One First National Plaza Chicago, Illinois 60603 Dean dansell, Esquire Assistant Attorney General 188 West Randolph Street Suite 2315 Chicago, Illinois 60601 Mr. Roger Walker, Resident Inspector U. S. Nuclear Regulatory Commission Post Office Box 224 Mar;.illes, Illinois 61364 S
Enclosure Request For Additional Information LaSalle County Station, Unit Nos.1 & 2 Docket Nos. 50-373 and 50-374 030.0 Instrumant & Control'Systen Branch 031.285 Several instances have been reported where automatic closure of the containment ventilation / purge valves would not have occurred because the safety actuation signals were either manually overriden or bypassed (blocked) during normal plant operations.
In addition, a related design deficiency with regard to the resetting of engineered safety feature actuation signals has been found at several operating facilities where, upon the reset of an ESF signal, certain safety related equipment would return to its non-safety mode.
Specifically, on June 25, 1978, Northeast Energy Company discovered that intermittent containment purge operations had been conducted at Millstone Unit No. 2 with the safety actuation signals to redundant containment purge isolation valves (48 inch butterfly valves) manually over.-iden and inoperable. The isola-tion signals which are required to automatically close the purge valves to assure containment integrity were manually overriden to allow purging of containment with a high radiation signal present. The manual override circuitry designed by the plant's architect / engineer defeated not only the high radiatien signal but also all other isolation signals to these valves. To manually override a safety actuation signal, the enerator cycles the. valve control switch to the closed position and then to the ooen position. This action energized a relay which u
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3 5 blocked the safety signal and allowed manual operation independent of any safety actuation signal. This circuitry was designed to oermit reopening of certain valves after an accident to allow manual operation of required safety equipment.
On September 8,1978, the staff was advised that, as a matter of routine, Salem Unit flo. I had been venting the containment through the containment ventilation system valves to reduce pressure.
In certain instances this venting has occurred with the containment high particulate radiation monitor isolation signal to the purge and oressure-vacuum relief valves overridden.
The override of this containment isolation signal was accomolished by re-setting the train A and B reset buttons. Under these circumstances, six valves in the containment vent a[id purge systems could be opened with the radiation isolation signal present. This override was cerforced after verify-ing that the actual containment particulate levels were acceptable for vent-ing. The licensee, after further investigation of this practice, determined that the reset of the particulate radiation monitor alarm also overrides the containment isolation signal to the purge valves such that the purge valves would not have automatically closed on an emergency core cooling system (ECCS) safety injection signal.
A related design deficiency was discovered during a review of system operation-following a recent unit trip and subsequent safety injection at North Anna fio.1.
Specifically, it was found that certain equipment important to safety (for example, control room habitability system damoers) would return to its.non-safety mode following the-reset of an ESF signal.
i 3-In addition, many utilities do not have safety grade radiation monitors to 1
i initiate containment. isolation.
SAFETY SIGNIFICANCE The overriding of certain containmer,~ ventilation isolation signals could also c
bypass other safety actuation signals and thus prevent valve closure when the other isolation signals are present. Although such designs may be acceptable, and even necessary, to accomplish certain reactor functions, they are generally I
unacceptable where they result in the unnecessary bypassing of safety actuation signals. Where such hypassing is also inadvertent, a more serious situation i
is created especially where there is no bypass indication system to alert. the operator.
- l Where the resetting ESF actuation signals, such as safety injection, directly causes ecuipment important to safety to return to its non-safety mode, orotec-l tive actions of the affected systems could be prematurely necated when the l
associated actuation signal is reset.
Prompt operator action would be required to assure that the necessary equipment is returned to its emergency mode.
l The use of non-safety grade monitor to initiate cc.
inment isolation could seriously degrade the reliability of the isolation system.
STAFF POSITION i
It is our position that, in addition to other aoplicable criteria,.the follow-ing should be satisfied for all operating license aoplications currently under review:
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- 1) The overriding of one tyce of safety actuation signal (e.g., carti-culate radiation) should not cause the blocking of any other tyoe of safety actuation signal (e.g., iodine radiation, reactor oressure) i for those valves that have no function other than containnent isolation.
- 2) Physical features (e.g., key lock switches) should be provided to en-sure adequate administrative controls.
- 3) A system level annunciation of the overridden status should be provided for every safety system imoacted when any override is active.
(See R.G.
'i 1.47).
- 4) The following diverse signals snould be provided to initiate isolation of the containment purge / ventilation system: containment high radiation, safety injection actuation, and containment high oressure (where con-tainment high pressure is not a cortion of safety injection actuation).
1 5)
The instrumentation systems orovided to initiate containment purge ventila-tion isolation should be designed and oualified to Class IE criteria.
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- 6) 'The overriding or resetting of the ESF actuation signal should 'not cause any equipment to change position.
Accordingly, you are requestad to review your orotection system design to deter-mine its degree of conformance to these criteria. You should report the results of your review to us describing any departures from the criteria and the corrective actions to be implemented. Design departuresofor which no correcti action is planned should be justified.
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The following definitions are given for clarity.
a0verride:
The signal is still oresent, and it is blocked in order to perform a function contrary to the sianal.
bReset:
The signal has come and gone, and the circuit is being cleared in order to return it to the normal condition.
Lil 286 Provide a response to DIE Bulletin 79-27, dated November 11, 1979, for LaSalle.
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LEVEL MEASUREMENT ERRORS DUE TO ENVIP,0NMENTAL TEMPERATURE EFFECTS ON LEVEL INSTRUMENT REFERENCE LEGS 031.287 On June 22, 1979, Westinghouse' Electric Carporation reported to NRC,5 potential safety hazard under 10 CFR 21. This report addresses errors generated in the steam generator level indicaticn sensors following high energy pipe break accidents inside containcent. - further; the report implies that previous analyses of peak contain ent te per.attfrc:
and pressure may.have been non onservative.
Srcats of this type can result in hcatup of the steam generator icvel esasurerent reference leg resulting in a decrease of the water coluca density with a consequent increase in the indicated steam generator water icvel (i.e., indicated level excceding actual level)1 IE Bulletin 79-21 includes further information on this problem and addresses appropriate actions which are to be taken by licensees of operating plants.
Applicants 'for an~
era ting itense are requested to submit a response to the following questions and to revise their safety analysis report consistent with this response.-
1.
Describe the liquid level rea'suring' systems within containment that are used to initiate safety actions or are used to provide port-accident ronitoring infon.ation., Provide a description of the type of reference leg used i.e., open column or sealed reference leg.
Provide an evaluation of the effect.of post. accident arbient tetperatures 2.
on the indicated water level to, determine the change in indicated level relative to actual water level This evaluatien must inciude other sources of ermr including the effects of varying fluid pressure and flashing of, reference leg to, steam on the water level reasurements.
Trovide an analysis of tite idpact th'at the leverreasuremant errors in 3.
control and pmtection systems (2 above) have on the assu ptions used in the plant transient and accident analysis, This should include a l
review of all safety and control setpoints derived fec= level signals to verify that the setpoints will initiate the action required by the plant safety, analyses t!)roughout the range of ambient te peratures encountered by the instrumentation, including accident terreratures.
If this analysis demonstrates that level measurement errors are greater than assumed in the safety analysis, address the corrective action to l
be taken. The corrective actions consider.ed should include design chances that could be made to ensure that containment temoerature effects l
are automatically accounted for. These measures ray include setroint changes as an acceptable corrective action for the short term. Howeser,
- some fonn of temperature compensation or modification to eliminate or reduce temperature ermrs should be investigated as a long term solution.
4.
Review and indicate the required revisions, as necessary, of c.crgency procedures to include specific information obtained from the review and evaluation of Items 1, 2, and 3 to ensure that the operators are instructe:
on the potential for and ragnitude of erroneous level sicnals. Trovide 9 copy of tables, curves, or correction factors that wouId be applied to post, accident monitoring systems that will be used by plant operators.
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110.0 Mechanical Engineerina Branch 111.82 In order that we can determine if LaSalle meets our oosition concerning pressure isolation valves (See Attachment A), the following additional information is required:
1.
Provide a list of all pressure valves which perform a pressure isolation function between the RCS and low oressure system identified by valve number and system.
2.
Provide information as to the methods used for extrapolating the low pressure leak rate (Appendix J, Type C test with water) to the RCS operating pressure leak rate.
3.
Provide the leak rate acceptance criteria for each valve. ' One GPM at full RSC pressure is the present leak rate criteria accepted by us.
1 4.
Provide the mode and at what point during the shutdown the leak rate testing will be performed. Our position is that testing should l
be performed at the end of the shutdown prior to entering Mode 2.
j 5.
For those pressure isolation barriers which consi_st of check valves in -
series with motor operated valves (MOV). provide ini9r. nation as to the provisions available for determining the cressure on the high pressure-side of the MOV prior to performing the quarterly Section XI operability' l
tests.
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ATTACHMENT A MECHANICAL ENGINEERING BRANCH POSITION There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure. There are also some sys tems which are rated at full reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systens from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems. The leak tight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the appropriate requirements of IWV-3420 of Section XI of the ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action, i.e., shutdcwn or system isolation when the final approved leakage limits ara not met. Also surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical specifications.
Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, af ter valve maintenance prior to return to service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justification is given.
The testing interval should average to be approximately one year. Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance and etc.
The staff's present position on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time. Significant increases over this limiting valve would be an indication of valve degradation from one test to another.
Leak rates higher than 1 GFM will be considered if the leak rate changes are below 1 CPM above the previous test leak rate or system design precludes measurir.g 1 GPM with sufficient accuracy. These items will be reviewed on a case-by-case basis.
The Class 1 to Class 2 boundary will be considered the isolation point which must be p otected by redundant isolation valves.
In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide isolation, only two of the velves need to be leak tested.
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