ML20039D578
| ML20039D578 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 12/23/1981 |
| From: | Linder F DAIRYLAND POWER COOPERATIVE |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.1, TASK-2.K.3.28, TASK-TM LAC-7990, NUDOCS 8201050228 | |
| Download: ML20039D578 (41) | |
Text
T 8
D DA/RYLAND h
[k COOPERAT/VE
- PO BOX 817 2615 EAST AV SOUlH = LA CROSSE WISCONSIN 54601 (608) 788-4000 December 23, 1981 In reply, please refer to LAC-7990 l
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DOCKET.NO. ca-409 0)
- 4) 0 U.
S.
Nuclear Regulatory Commission
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ATTN:
Mr. Darrell G.
Eisenhut, Director g
g' Division of Licensing RECEIVED
((
Of fice of Nuclear Reactor Regulation Division of Operating Reactors it JAN 4 1982m-42 Washington, D.
C.
20555
- unsmaurmusesse M G sanaG G M I
SUBJECT:
DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR (LACBWR).
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PROVISIONAL OPERATING LICENSE NO. DPR-45 b)
ADDITIONAL INFORMATION - NUREG-0737
REFERENCES:
(1)
NRC Letter, Eisenhut to All Licensees of Operating Plants, Dated October 31, 1980.
(2)
DPC Letter, Linder to Crutchfield, LAC-7983, Dated December 17, 1981.
(3)
DPC Letter, Linder to Crutchfield, LAC-7834, Dated September 29, 1981.
(4)
DPC Letter, Linder to Crutchfield, LAC-7790, Dated September 14, 1981.
(5)
DPC Letter, Linder to Denton, LAC-6680, Dated December 6, 1979.
Gentlemen:
Your letter (Reference 1) transmitted NUREG-0737, " Clarification of TMI Action Plan Requirements",
which established the requirements to submit additional information to the Nuclear Regulatory Commission by January 1, 1982.
i Enclosed with this letter are the following attachments: - Item II.F.1.3
- Containment High Range Radiation Monitor - Item II.F.1.4
- Containment Pressure Monitor i - Item II.F.1.5
- Containment Water Level Monitor - Item II.F.1.6
- Containment Hydrogen Monitor - Item II.K.3.28 - Verification of Qualification of Accumulators on Automatic Depressurization Systems 8201050228 811223 f
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s Mr. Darrell G.
Eisehnut, Director Decembe r 23, 1981 U.
S.
Nuclear Regulatory Commission LAC-7990 Item II.D.1, Performance Testing of BWR Safety Valves, was transmitted in Reference 2.
Regarding Item II.K.3.30, Revised Small Break Loss-of-Coolant Accident-Methods, the need for model revision was discussed on' February 11, 1981 by three representatives.of DPC's consultant Nuclear Energy' Services with the NRC staf f personnel at Bethesda,, Maryland.
The NRC will review previous analysis to determine if any model revisions are required.
The material required for review was submitted to the NRC by Reference 3.
If there are any questions' regarding this submittal, please let us know.
Very truly yours, DAIRYLAND ER COOPERATIVE
~
A Frank Linder, General Manager FL:JDP:af cc:
J. G.
Keppler, Regional Director, NRC-DRO III
. NRC Resident Inspector STATE OF WISCONSIN )
)
COUNTY OF LA CROSSE) d !-*Y day of December, 1981, the Personally came before me this above named Frank Linder, to me known to be the person who executed the foregoing instrument and acknowledged the same.
/'
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Notary Publiq/ La Crosse County, Wisconsin.
i My Commission Expires February 26, 1984.
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ATTACHMENT 1 CONTAINMENT HIGH RANGE RADIATION MONITOR I
NUREG-0737, Item II.F.1.3
T
. o CONTAINMENT HIGH-RANGE. RADIATION MONITORS NUREG 0737 ITEM II.F.1.3 NRC POSITION & CLARIFICA TION Position In containment radiation-levet monitore vith a maximum range of 108 rad /hr shall be instatted.
A minimum of two such monitore that are physically esparated shall be provided.
Monitore shall be developed and qualified to function in an accident environment.
Clarification (1)
Provide tuo radiation monitor eyeteme in containment which are documented to meet the requiremente of Table II.F.1-3.
(2)
The specification of 100 rad /hr in the above position vae based on a calculation of post-accident containment radiation levels that included'both particulate (beta) and photon (gamma) radiation.
A radiation detector that responde to both beta and gamma radiation cannot be qualified to post-LOCA (lose-of-coolant accident) containment'environmente but gamma-sensitive instrumente can be so qualified.
In order to foltou the course of an accident, a containment monitor that measures only gamma radiation is adequate.
The requirement uae revised in the October 30, 1979 letter to provide for a ' photon-only measurement with an upper range of 107 R/hr.
(3)
The monitore shalt be tocated in containmnetle) in a manner, to provide a reasonable aseeeement of area radiation as conditione incide containment.
The monitore shalt be videly separated so as to provide independent measuremente and shalt "vieu" a large fraction of the containment volume.
Monitore should not be placed in areas chich are protected by massive shielding and should be reasonably accessible for replacement, maintenance, or calibration.
Placement high in a reactor building dome is not recommended because of potential maintenance difficulties.
(4)
For BVR Mark III containmente, tuo euch monitoring systems should be inside both the primary containment (dryvett) and the secondary containment.
(S)
The monitore are required to respond to gamma photone with energies as too as 60 kev and to provide an essentially flat response for gamma energies between 100 kev and 3 MeV, as
'specified in Table II.F.1-3.
Monitore that use. thick shielding to increase the upper range vitt under-catimate poet-accident radiation levete'in containment by several WP1 1-1 m
m
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. orders of magnitude because of their insenettivity to too energy gammas and are not acceptable.
TABLE II. F.1-3 CONTAINMENT HIGH-RANGE RADIA TION MONITOR The capability to detect and measure the REQUIREMENT radiation levet within the reactor contain-ment during and following an accident 1 rad /hr to 108 rade/hr (beta and gamma) or RANGE alternatively 1 R/hr to 107 R/hr (gamma only).
60 kev to 3 MeV photone, with linear energy
RESPONSE
response + 20% for photone of 0.1 MeV to 3 MeV.
Instrumente must be accurate enough to provide usable information..
A minimum of two physically separated monitore REDUNDA NT (i.e, monitoring videly separated spaces within containment).
Category 1 instrumente as described in Appendix DESIGN AND Q UA LIFICATION A,
except as listed below.
In situ calibration by electronic signal SPECIA L CALIBRATION eubstitution is acceptable for att range decades above 10 R/hr.
In situ calibration for at least one decade betov 10 R/hr ehatt be by means of calibrated radiation source.
The original Laboratory calibration'is not an acceptable position due to the possible differences after in situ instattation For high-range calibration, no adequate sources exist, so an alternate vae provided.
Calibrate and type-test representative specimens SPECIA L ENVIRONMENTAL of detectore at sufficient pointe to demonstrate t
Q UA LIFICATIONS linearity through att scales up to 100 R/hr.
^
Prion to initial use, certify calibration of each detector for at least one point per decade of range between 1 R/hr and 103 R/hr.
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r DPC RESPONSE LACBWR has installed two General Atomic High Range Radiation Monitors.
The units each consist of a Model RD23 gamma radiation detector, a Model 2C readout module, a Model RP23 power supply and share a Model RP20-01 Nim Bin.
Documentation on environmental qualification of the high range radia-tion monitors was submitted in Reference 4.
The in situ calibration for one decade below 10 R/hr has not yet been completed, but will be completed by January 1, 1982.
A 10CFR21 report has been filed with the U.
S.
Nuclear Regulatory Commission by the equipment vendor.
This report discussed environmental qualification dif ficulties with the cable.
Replacement cable has been requested by DPC and once received will be installed at the earliest opportunity af forded by plant availability.
WP1 1-3
ATTACHMENT 2 CONTAINMENT PRESSURE MONITOR NUREG-0737, ITEM II.F.1.4 h -
CONTAINMENT PRESSURE MONITOR NUREG 0737, Item II.F.1.4 NRC POSITION AND _CL A RTFICA TION POSITION A continuous indication of containment pressure shall be provided in the control room of each operating reactor.
Measurement and indication capbility shalt include three times the design pressure of the containment for concrete, four times the design pressure for steet, and -5 peig for all containmente.
CL A RIFICA TION (1)
Design and qualification criteria are outlined in Appendix A.
(2)
Measurement and indication capability shall extend to 5 peia for sub-atmospheric containment.
(3)
Tuo or more instrumente may be used to meet requiremente.
However, instrumente that need to be evitched from one scale to another scale to meet the range requiremente are not acceptable.
(4)
Continuous display and recording of the containment pressure over the specified range in the control room is required.
(5)
The accuracy and response time specifications of the pressure monitor shall be provided and justified to be adequate for their intended function.
DPC RESPONSE l
A reactor is required to provide containment pressure indication to cover the range of atmospheric pressure to four times design j
pressure (if steel).
LACBWR does not nave a sub-atmosphere containment and does not require indication to 5 psia.
l Additionally, LACBWR is equipped with two vacuum breakers to l
insure a sustained vacuum cannot exist.
Continuous display of l
each containment pressure channel is provided in the control room, however, recording is not.
The emergency operations f acility/ technical support center data monitoring modification will provide a printed record of containment pressure using one minute averages.
The reactor is fueled with stainless steel clad l
fuel which reduces the threat of pressure spikes due to hydrogen recombination (no zircaloy clad fuel is used at LACBWR; the re fo re,
no hydrogen is generated from this type of source).
The l
containment pressure indicator at LACBWR is advisory only, it does not actuate safety systems.
Automatic actuation of emergency core cooling systems and isolation of containment are not accomplished l
using this containment pressure indication.
Separate pressure WP1 2-1 l
o switches are provided for these safety system actuation functions.
The range of pressure indication provided is approximately two times containment design (100 psig full scale) on one channel and 70 psig full scale on the other channel.
This instrument range is adequate for both integrated leak rate testing and post-accident monito ring.
The design system accuracy is approximately + 3%.
Time response testing is not performed on this equipment, as it performs only a monitoring function.
This direct indication is adequate without such testing.
NRC CRITERTA_
Applicability To the extent feasible and practical (in conformance with the stipulations of Appendix A and ancillary requiremente), equipment is to be installed by the specified implementation dates.
Where equipment is unavailable, precluding conformance uith equipment qualification and schedular raquiremente, the implementation dates are to be met by installation of best available equipment.
In euch caeae, deviatione a=e to be described and a schedule for the feasible instattation of equipment in conformance uith the stipulatione of Regulatory Guide 1.97 (vhen the guide is used) is to be provided.
Appendix A is consistent vith our current draft version of Regulatory Guide 1.9 7.
We expect no further revisione to our requiremente.
Criteria (1)
The instrumentation should be environmentally qualified in accordance with Regulatory Guide 1.89 (NUREG-0588).
l Qualification applies to the complete instrumentation channet from censor to display uhere the display is a direct indicating meter or recording device.
Where the instrumentation channet signal is to be used in a computer-i based display, recording and/or diagnostic program, qualification applies to and includes the channet isolation device.
The location of the isolation-device should be auch that it vould be accessible for maintenance during accident conditions.
The seismic portion of environmental l
qualification should be in accordance uith Regulatory Guide 1.100.
The instrumentation should continue to read uithin the required accuracy foltoving, but not necessarity during, a safe shutdoun earthquake.
Instrumentation, uhoec ranges are required to e= tend beyond those ranges calculated in the most severe design basis accident event for a given variable, l
ehould be qualified using the foltooing guidance.
T WP1 2-2
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4.
The qualification environment shalt be based on the design basis accident evente, except the assumed maximum of the value of the monitored variable shall be the value equal to the maximum of the variable.
The monitored variable shall be assumed to approach this peak by extrapolating the moet severe initial ramp associated with the design basie accident evente.
The decay for this variable shall be considered proportional to the decay for this var ~~able associated with the design basie accident evente.
No additional qualificaiton margin neede to be added to the extended range variable.
All environmental envelopea except that pertaining to the variable measured by the information display channet shall be those associated with the design basic accident evente.
The above environmental qualification requirement does not account for eteady-etate elevated levels that may occur in other environmental parameters associated with the extended range variables.
For example, a sensor measuring containment pressure must be qualified for the measured procese variable range, but the corresponding ambient temperature is not mechanisticatty linked to that pressure.
Rather, the ambient temperature value is the bounding value for design basis accident evente analyzed in Chapter 15 of the final safety analysis report (PSAR).
The extended range requirement is to ensure that the equipment vitt continue to provide information should conditione degrade beyond those postulated in the safety analysis.
Since variable ranges are nonmechanisticatty determined, extension of associated parameter levels is not justifiable and has, therefore, not been required.
DPC RESPONSE The existing containment pressure channels were installed during facility construction and are currently being reviewed in the initial Systematic Evaluation Program.
Originally, facility design did not address the seismic issue as the plant was included in Zone 0 of the Uniform Building Code, 1958 Edition. The Systematic Evaluation Program seismic evaluation will be complete in the evaluation phase by August 1982 and any required modification by mid-1985.
This evaluation is a review of equipment anchorages for transmitters, cable trays and ind icato rs.
A review of environmental qualifications of safety related electrical equipment is being performed under the Systematic Evaluation Program.
The containment pressure indicators are located outside of the containment building and are not subjected to a harsh environment in the accident they are designed for except for radiation exposure.
The Main Steam Line Break incident
~
which would subject the containment pressure instrumentation transmitter'to high temperature and humidity does I/
WPl 2-3 r--
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N not require utilization of this equipment..'The equipment was
-intended to function in~a post loss-of-coolant: accident environment.
Current review indicates that shielding or
' relocation may be required to fully comply with the required environmental radiation exposure.
The installation of this equipment predated the issuance of 1
f~
Regulatory Guides 1. 89 and - 1.100 and NUREG-0588. ~It is our conclusion'that the instrumentation is adequate now.
Should the SEP seismic and environmental qualification programs indicate the need -for modification, it.will be accomplished.
i-NRC CRITERIA (2)
No single failurs uithin either the accident-monitoring instrumentation, its aumiliary supporting features or its pover sources concurrent with the-failure that'are a condition or reeutt of a specific accident should prevent the operator from being presented the information necessary for him to determine the safety status of the plant and to bring the plant to a safe-condition and maintain it in a safe condition foltooing that accident.
Where failure of one i
accident-monitoring channet reeutte in ambiguity (that is, the redundant dieptage disagree) which could lead the operator to defeat or fait to accomplish a required safety function, additional information should be provided to attou the operator to deduce the actual conditions in the plant.
.This may be accomplished by:
(a) providing additional independent channets of information of the same" variable (addition of an identical channet), or (b) providing an independent channet'which monitore a different variable bearing a knoun retationship to the multiple channets (addition of a diverse channet), or (c) providing the capability, if sufficient time is available, for the^ operator i
to perturb the measured variable and' determine which channet has failed by observation of the response on each instrumentation channet.
Redundant or diverse channete should.be'electricatty independent, energized from station Class 15 power source, and physically separated in'accordance uith Regulatory Guide 1.75 up to' and including any isolation device.
At least one channel should be displayed on a direct indicating or recording device.
(NOTE:
Within each i
redundant division of a safety eyetem, redundant monitoring channels are not required'
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.The instrument' channels.are supplied by separate' essential buses
- which were constructed prior to :the establishment of Class lE
- power source requirements.
There exist 2 separate' transmitters,
~
- cabling runs and indicators.
Cable separation does not comply with Regulatory Guide: 1. 75.
No isolation devices are' included as the only output of.this signal is indication, no automatic actions occur.. The redundant channels are electrically-independent.
NRC CRITERA
. (3)~ The instrument should be energized from station Class 1E-
.pover sources..
- DPC RESPONSE The instrumentation is energized by two separate essential buses, one ' energized through inverter 1A and one.through inverter 1B.
The " Class lE Power Source" requirementswas not part of the original plant design.
A seismic evaluation program is currently in progress to determine what, if any, modifications are required.
The power supplies-as-they exist'have been accepted by the NRC.to-meet the Interim ~ Acceptance Criteria for Emergency Core Cooling Systems.
NRC CRITERIA (4)
An: instrumentation channel should be available prior to an
-accident except as provided in Paragraph 4.11,
" Exemption",
as defined in IEEE Std. 279 or as specified in technical specifications.
DPC RESPONSE Both pressure channels are maintained energized during operation.
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r NRC CRITERTA
.(5)
The recommendations of_the follouing regulatory guides pertaining to quality assurance should be followed:
1.28
" Quality Accurance Program Requiremente (Design and Contruction)"
1.30
" Quality Assurance Requiremente for the Installation, Inspection, and Testing of Instrumentation and Electrical Equipment" 1.38
" Quality Assurance Requiremente for Packaging, Shipping, Receiving, Storage,.and Handling of Items for Vater-Cooled Nuclear Pouer Plante" 1.58
" Qualification of Nuclear Pouer Plant inspection, Examination, and Testing Personnel" 1.64
" Quality Assurance Requiremente for the Dceign of Nuclear Pouer Plante" 1.74
" Quality Assurance Terme and Definitions" 1.88
" Collection, Storage, and Maintenance of Nuclear Pouer Plant Quality Assurance Records" 1.123
" Quality Assurance - Requiremente for Control of Procurement of Items and Services for Nuclear; Pouer Plante" 1.144
" Auditing of Quality Assurance Programs for Nuclear Power Plante" Task
" Qualification of Quality Assurance Program Audit i
RS 810-5 Personnel for Nuclear Pouer Plante" (Guide number to be inserted).
Reference to the above regulatory guides (except Regulatory Guides 1.30 and 1.38) are being made pending issuance of a l
regulatory guide endorsing NQA-1 (Task RS 002-5), nov in progrees.
l DPC RESPONSE l
The containment pressure indicators are maintained under the l
provisions of the LACBWR Quality. Assurance' Program which has been l
approved by the NRC.
The LACBWR QA Program is based on the requirements of 10CFR50, Appendix B, Regulatory _ Guide 1.33, which addresses the requirements of ANSI N18.7-1972 and ANSI N45.2-1971.
No further modification of the Quality Assurance Program is anticipated at this time.
i
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2-6
NRC CRITERIA (6)
Continuous indication (it may be by ri:co rding) display should be provided at att times.
Where tuo or more instrumente are needed to cover a particular range, overlapping of instrument span should be provided.
DPC RESPONSE As referenced in response to Criteria 2, continuous indicators are available for each channel.
l l
l NRC CRITERIA (7)
Recording of instrumentation readout information should be provided.
Where trend or transient information is essential for operator information or action, the recording should be analog strip-chart or stored and displayed continuously on demand.
Intermittent displays, such as data loggers and scanning recorders, may be used if no significant transient responec information is likely to be lost by such devices.
DPC RESPONSE No recording of containment pressure is provided.
One channel is being added to the post-accident Technical Support Center /Energency Ope rations Facility printout.
Continous scanning will provide one minute averages of containment pressure which will in post-accident situations be printed out each minute at both the Technical Support Center and the Emergency Operations Facility.
NRC CRITERIA (8)
The instrumente should be specifically identified on the contret panete so that the operator can easily discern that they are intended for use under accident conditions.
DPC RESPONSE The instrument readout is labeled.
WP1 2-7
r NRC CRITERIk.
(9).The transmission of signate-from the~ instrument or associated sensore for'other use should be through-isolation devices'that are designated as part of monitoring instrumentation and that meet the provisions of the document.
DPC RESPONSE There currently are no output uses other than indication so isolation devices are not used.
NRC CRITERTL (10)
Means should be provided for checking, with a high degree of confidence, the operationat' availability of each monitoring channel, including its input eensor, during reactor operation.
.This may be accomplished in various vays; for example:
(a)
By perturbing the monitored variable.
(b)
By introducing and varying, as appropriate, a substitute input to the sensor of the same nature as the measured.
variable.
(c)
By cross-checking betveen channets that bear a known relationship to each other and that have readoute available.
DPC RESPONSE The method of cross-checking between channels is aiailable at LACBWR to verify operational availability.
NRC CRITERIA (11)
Servicing, testing, and calibrating programe should be.
epecified to maintain the capability of the' monitoring instrumentation.
For thoac instrumente where the required interval between testing uitt be tees than the normat time interval between generating station shutdowne, a capability for testing during power operation should be provided.
DPC RESPONSE The containment pressure indicators are calibrated each refueling shutdown.
WP1 2-8
-- a
r-NRC CRITERIA.
(12)
Whenever means for removing channele from;eervice are included' in the design, the design ehould facilitate;adninistrative controtl of:the ' access to euch removat means.
DPC RESPONSE The containment' pressure -channels can be removed from service.undsr administrative controls if required.
As they-do not activate any auxiliaries or equipment, no bypass features are required.
NRC CRITERIA (13)
The design should facilitate administrative. control of the accese to att setpoint adjustmente, module calibration adjust-mente, and test pointe.
DPC RESPONSE The design includes the transmitter in the turbine hall piping tunnel and the indicators in the' control room.- All equipment is accessible.
The Duty Shif t Supervisor exercises administrative control. by Maintenance Request Forms over this equipment.
NRC' CRITERIA (14); The monitoring instrumentation design should minimize the' development of conditione that vould cause metere, annunciatore, recordere, alarme, etc., to give'anomatous indicatione potentially ' confusing - to the operator..
DPC RESPONSE The instrumentation is designed to minimize anomalous indication.
-The indicator is simple and requires no conversion to interpret.-
l NRC CRITERIA e-(15) 'The instrumentation should be designed'to facilitate the recogn'ition, Location, replacement, repair, or adjustment'of j
. malfunctioning componente or modulee.
j DPC RESPONSE The' instrumentation is well designed for repair -and ' adjustment with
.a - minimum. number of components.
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NRC CRITERIA (16)
To the extent practical, monitoring instrumentation inpute should be from sensore that directly measure the desired variables.
DPC RESPONSE The sensors directly measure containment pressure.
NRC CRITERIA, (17)
To the extent practical, the same instrumente should be used for accident monitoring as are used for the normat operatione of the plant to enable the operator to use, during accident situatione, instrumente Uith which the operator is most familiar.
However, where the required range of monitoring instrumentation results in a lose of instrumentation sensitivity in the normat operating range, separate instrumente should be used.
DPC RESPONSE Not applicable.
No significant pressure above atmospheric is normally present in the reactor containment during normal ope ra t ion.
NRC CRITERIA (18)
Periodic testing should be in accordance with the applicable portions of Regulatory Guide 1.118 pertaining to testing of instrumente channete.
DPC RESPONSE The calibration is performed each refueling outage.
WP1 2-10 L
..m m.
r ATTACHMENT 3 CONTAINMENT WATER LEVEL MONITOR NUREG-0737, ITEM II.F.1.5
CONTAINMENT WATER LEVEL MONITOR N UREG-0737, ITEM, II,F.1.5 NRC POSITION AND CLA RIFICA TION.
Position A continuous. indication of containment water levet shall be provided in the control room for att plante.
A narrou range instrument shall be provided for PVRs and cover the range from the bottom to the top of the containment sump.
A vide range instrument shalt ateo be provided for PWRe and shall acuer the range from the bottom of the containment ~to the elevation equivalent to a 600,000 gallon capacity.
For BWRe, ~ a vide ' range instrument shall be provided and cover the range from the bottom to S feet above the normat vater levet of the suppression pool.
' Clarification (1)
The containment' uide-range water levet indication channels.
chall meet the design and qualification criteria as outlined in Appendix A.
The narrou-range channet shall meet the requiremente of Regulatory Guide 1.89.
(2)
The measurement capability of 600,000 gattone is based on recent plant designs.
For older plants with smaller water capacities, licensees may propose deviations from this requirement based on the available uater supply capability at their plant.
'(3). Narrou-range water levet monitore are required for att sizee of eumps but are not' required in those plants that do.
not contain cumps inside the~ containment.
(4)
For BWR pressure-suppression containmente, the emergency core cooling system (ECCS) suction line intete may be used as a starting reference point for the narrou-range and vide-range water levet monitors, instead of the bottom of the suppression pool.
(5)
The accuracy requiremente of the vater levet monitore chall be provided and justified to be adequte for their intended function.
DPC RESPONSE A' boiling water reactor is required to provide a wide-range instrument which' covers the range from the bottom of the suppression pool to 5 feet above the normal water level.
WP1 3-1
LACBWR is an Allis-Chalmers boiling water reactor which does not have a suppression pool.
Clarification Items 2 and 3 apply to pressurized water reactors and Item 4 applies to pressure suppression containments on BWRs.
LACBWR, in Reference 5, specified the range of instrumentation and indicated that it is adequate.
The design channel accuracy is approximately + 3%.
The two containment water level transmitters each cover 41 feet of the reactor containment building height including the full height of the reactor vessel.
The equipment as installed is compared to the requirements of Appendix B of NUREG-0737 as follows:
NRC CRITERIA Applicability To the extent feasible and practical (in conformance with the stipulations of Appendix A and ancittary requiremente), equipment is to be instatted by the specified implementation dates.
Where equipment is unavailable, precluding conformance uith equipment qualification and schedular requiremente, the implementation dates are to be met by instattation of best available equipment.
In auch cases, deviatione are to be described and a schedule for the feasible instattation of equipment in conformance with the s tipulations of Regulatory Guide 1.9 7 (when the guide is used) is to be provided.
Appendix A is concietent with our current draft version of Regulatory Guide 1.97.
We expect no further revisione to our requiremente.
Criteria (1)
The instrumentation should be environmentally qualified in accordance with Regulatory Guide 1.89 (NUREG-0588).
Qualification applies to the complete instrumentation channet from eensor to display where the display is a direct indicating meter or recording device.
Where the instrumentation channet signal is to be used in a computer-based display, recording and/or diagnostic program, qualification applies to and includes the channet isolation device.
The location of the isolation device should be euch that it would be accessible for maintenance during accident conditione.
The seismic portion of environmental qualification should be in accordance uith Regulatory Guide 1.100.
The instrumentation should continue to read within the required accuracy foltouing, but not necessarity during, a safe shutdoun earthquake.
Instrumentation, whose ranges are required to extend beyond those ranges calculated in the most severe design basis accident event for a given variable, should be qualified using the fctioving guidance.
WP1 3-2
The qualification environment shall be based on the design basie accident evente, except the assumed maximum of the.
value of the monitored variable shall be the value equal to the maximum of the variable.
The monitored variable shall be assumed to approach this peak by extrapolating the most severe initiat ramp associated with the design basis accident evente.
The decay for this variable chait be considered proportional to the decay for this variable associated with the design basis accident evente.
No additional qualificaiton margin needs to be added to the extended range variable.
Att environmental envelopes except that pertaining to the variable measured by the information display channet shall be those associated uith the design basis accident evente.
The above environmental qualification requirement does not account for steady-etate elevated levele that may occur in other environmental parametere associated uith the extended range variables.
For e= ample, a sensor measuring containment pressure must be qualified for the measured process variable range, but the corresponding ambient temperature is not mechanisticatty linked to that pressure.
Rather, the ambient temperature value is the bounding value for design basie accident evente analyzed in Chapter 15 of the final safety analysis report (PSAR).
The extended range requirement is to ensure that the equipment vitt continue to provide information should conditione degrade beyond those postulated in the safety analysis.
Since variable ranges are nonnechanisticatty determined, extension of associated paraneter tevete is not justifiable and hae, therefore, not been required.
DPC RESPONSE The installed water level channels were installed during initial f acility construction and are currently being reviewed during the Systematic Evaluation Program.
Originally, facility design did not address the seismic issue as the plant was included in Zone 0 of the Uniform Building Code, 1958 Edition.
The Systematic Evaluation Program seismic evaluation will be complete in the evaluation phase by August 1982 and any required modification (s) by mid-1985.
This evaluation is a review of equipment anchorages for transmitters, cable trays and indica to rs.
A review of environmental qualifications of safety related electrical equipment is being performed under the Systematic Evaluation Program.
The ccntainment water level indicators are located outside of the containment building and are not subjected to a harsh environment (in the accident they are designed for) except for radiation exposure.
The Main Steam Line Break incident which would subject the containment water level instrumentation transmitter to high temperature and humidity does WP1 3-3
not require utilization of this equipment.
The equipmtent'was
- intended to function in a post loss-of-coolant accident-environment.
Current review indicates that shielding or relocation may be required to fully comply with the required environmental radiation exposure.
The installation of this equipment predated the issuance of Regulatory Guides 1.89 and 1.100 and NUREG-0588.
It is our conclusion that the instrumentation is adequate now.
Should the SEP seismic and environmental qualification programs indicate the need for modification, it will be accomplished.
NRC CRITERIA (2)
No single failure within either the accident-monitoring instrumentation, its auxiliary supporting features or its pouer sources concurrent with the failure that are a condition or result of a specific accident should prevent the operator from being presented the information necessary for
-him to determine the safety status of the plant and to bring the plant to a safe condition and maintain it in a safe condition foltooing that accident.
Where failure of-one accident-monitoring channel reeutta in ambiguity (that is, the redundant displays disagree) ohich could lead the operator to defeat or fail-to accomplish 1 required safety funciton, additional infomvation should be provided to attou the operator to deduce the actual' conditione in the plant.
This may be accomplished by:
(a) providing additional independent channets of information of the same variable (addition of an identical channet), or (b) providing an independent channet which monitore a different variable bearing a knoun relationship to_the multiple channels (addition of a diverse channet), or (c) providing the capability, if sufficient time is available, for the operator to perturb the measured variable and determine which channet has failed by observatien of the response on each instrumentation channet.
Redundant or diverse channete should be electricatty independent, energized from station Class 1E pouer source, and physically separated in accordance uith Regulatory Guide 1.75 up to and including any isolationt device.
At least one channel should be displayed on a direc indicating or recording device.
(NOTE:
Within aach redundant division of a safety system, redundant monitoring channele are not required).
WP1 3-4 m -.
. m.
m...
.__________-.__.___,,__._________..___._._._______._____..___.____._._.__._.a
r DPC RESPONSE
'The instrument channels are supplied by separate essential buses which were constructed prior to the establishment of Class lE power source requirements.
There exists 2 separate transmitters and cabling runs to a single indicator with a selector switch.
Cable separation does not comply with F.egulatory Guide 1.75.
No isolation devices are included as the only output of this signal is indication, no. automatic actions occur.
The redundant
. channels are electrically independent and at least one channel is displayed on a direct-indicating device.
NRC CRITERA (3)
The instrumentation should be energized from station Class 1E pover sources.
DPC RESPONSE-The instrumentation is energized by two separate essential buses, one energized through inverter lA and one through inverter 1B.
The " Class lE Power Source" requirement was not part of the original plant design.
The seismic evaluation program is currently in progress to determine what, if any, modifications are required.
The power supplies as they exist have been accepted by
. the NRC to meet the Interim Acceptance Criteria for Emergency.
Core Cooling Systems.
NRC CRITERIA (4)
An instrumentation channel should be available prior to an accident except as provided in Paragraph 4.11, "E=emption",
as defined in IEEE Std. 279 or as specified in technical specifications.
DPC RESPONSE i
l Both channels are maintained energized during operation.
l i
l l-t i
WP1 3-5
.NRC CRITERIA (S)
The recommendations of the following regulatory guides pertaining to quality' assurance should be foltoued:
1.28
" Quality Assurance Program Requiremente (Design and Contruction)"
1.30
" Quality Assurance Requiremente for the Installation, Inspection, and Testing-of Instrumentation and Electrical Equipment" 1.38
" Quality Assurance Requiremente for Packaging, Shipping, Receiving, Storage, and Randling of Items for Water-Cooled Nuclear Pouer Plante" 1.58
" Qualification of Nuclear Power Plant Inspection, Examination, and Testing, Personnel" 1.64
" Quality Assurance Requirements for the Design of; Nuclear Power Plante" 1.74
" Quality A ssurance Terms and Definitione" 1.88
" Collection, Storage, and Maintenance of Nuclear Pouer Plant Quality Assurance Recordo" 1.123
" Quality Assurance Requiremente for Control of Procurement of Items and Services for Nuclear Pouer Plante" 1.144
" Auditing of Quality Assurance Programe for Nuclear Power Plante" Task
" Qualification of Quality Assurance Program Audit RS 8:9-5 Personnel for Nuclear Pouer Plante" (Guide number i
to be inserted).
j Reference to the above regulatory guides (except Regulatory Guides 1.30 and 1.38) are being made pending issuance of a regulatory guide endorsing NQA-1 (Taek'RS 002-5), now in j __
progrees.
DPC RESPONSE The containment water level indicators are maintained-under the provisions of the LACBWR Quality Assurance Program which has been approved by the NRC.
The LACBWR QA Program is based on the requirements of 10CFR50, Appendix B, Regulatory Guide 1.33, which addresses the' requirements of ANSI N18.7-1972 and ANSI N45.2-1971.
No further modification of the Quality Assurance Program is anticipated at this time.
WPl.
3-6 l
NRC CRITERZL (6)
Continuous indication (it may be by recording) display-chould be-provided at att times.
Where tuo or more instrumente are-needed to cover a particular-range, overlapping of instrument:
span should be provided.
DPC RESPONSE As referenced in response to Criteria 2, a single continuous indicator is available for the two channels, either of which may be selected for display.
NRC CRITERIA (7)
Recording of instrumentation readout information should be provided.- Where trend or transient information is essential for operator information or action, the recording should be analog strip-chart or stored and displayed continuously on demand.
Intermittent dieptage, such as data loggers.and scanning recorders, may be used if no significant transient response information is likely to be lost by such devices.
DPC RES'ONSE No recording of containment water level is provided.
One channel is being added to the post-accident Technical Support Center / Emergency Operations Facility printout.
Continous scanning will provide one minute averages of containment water level which will in post-accident situations be printed out.each niinute at both the Technical Support Center and the Emergency Operations Facility.
NRC CRZTQf[L (8)
The inetrumente should be specifically identified on the controt panete so that the operator can eaeity discern that they are intended for use under accident conditione.
DPC RESPONSE The instrument readout is labeled.
?
WP1 3-7 3
m m
.m
NRC CRITEPTA (9)
The transmission of signate from the instrument or associated sensors for other use should be through isolation devices that are designated as part of monitoring instrumentation and that meet the provisions of the document.
DPC RESPONSE There currently are no output uses other than indication so isolation devices are not used.
LRC CRITERLL (10)
Means should be provided for checking, uith a high degree of confidence, the operational availability of each monitoring channel, including its input sensor, during reactor operation.
This may be accomplished in various ways; for example:
(a)
By perturbing the monitored variable.
a substitute (b)
By introducing and varying, as appropriate, input to the sensor of the same nature as the measured variable.
(c)
By cross-checking betueen channels that bear a knoun relationship to each other and that have readoute available.
DPC RESPONSE The method of cross-checking between channels is available at LACBWR to verify operational availability.
NRC CRITERTA (11)
Servicing, testing, and calibrating programe should be specified to maintain the capability of the monitoring instrumentation.
For those instrumente uhere the required interval between testing uitt be less than the normat time interval between generating station shutdouns, a capability for testing during pouer operation should be provided.
DPC RESPONSE The containment water level indicator is calibrated each refueling shutdown.
WP1 3-8
.1 NRC CRITERZL
. (12 )
Whenever means for: removing channets from' service are included in the design, the design ehould facilitate administrative
. control of the access to such removat means.
DPC~ RESPONSE The containment water level channels can be removed from service under a5..iinistrative control if required. As they do not activate auxiliaries or equipment, no bypass features are. required..
MRC CRITERIA The~ design should facilitate administrative controtLof the (13). accese to all~eetpoint adjustmente, module calibration adjust-
^
mente, and test pointe.
DPC RESPONSE The design-includes the transmitter in the turbine hall piping tunnel and the indicators in the control room.
A11' equipment is accessible.
The Duty Shif t Supervisor exercises' administrative control by Maintenance Request forms over this equipment.
L NRC CRITERZL (14)
The monitoring' instrumentation design should minimize the development of conditions that could cause metere, annunciatore, recorders, alarme, etc., to give anomatous indicatione potentially confusing'to'the operator.
DPC RESPONSE The instrumentation is designed to minimize anomalous' indication.
The ' indicator is simple and requires no conversion to interrupt.
.NRC CRI,T,ERJJ E
(15 )
The instrumentation should be designed to facilitate the recognition, loca : 4.o n, replacement, repair, or adjustment of malfunctioning comp,nente or modules.
DPC RESPONSE TheEinstrumentation is well designed for repair and adjustment ~ with a minimum number of components.
WP1
.3-9
~
m
. ~.
j-i WRC' CRITERIA.
- (16 )) To the entent-practical, monitoring instrumentation inputs should be from sensors that'directly measure.the-desired
~
variables.
DPC RESPONSE The sensors directly measure' containment water level.
NRC CRITERIA
' (17 )
To the entent. practical, the same instruments should be used l'
for accident monitoring as'are used for the.normat operations of the plant to enable _the operator to use, during accident
. situations, instruments oith which the operator is most
. familiar.
However, where the required range of monitoring instrumentation resulte in a loss of instrumentation sensitivity in the normat operating range, separate instruments should be used.
-DPC RESPONSE Not. applicable.
A' water level is not normally present in the
. reactor containment during normal operation.
l NRC CRITERIA i
(18)
Periodic testing should be in accordance with the applicable:
portions of Regulatory Guide 1.118 pertaining to testing of
' instruments channels.
DPC RESPONSE The calibration is performed each refueling outage.
I f'
f 2
t i
I' l
WP1 3-10
ATTACHMENT 4 CONTAINMENT HYDROGEN MONITOR NUREG-0737, ITEM II.F.1.6
CONTAINMENT HYDROGEN MONITOR.
NUREG 0737,. ITEM II.F.1.6 NRC ~ POSITION AND CLARIPICA TION Position 4
A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room.
Measurement capability shall be provided'over the range of 0 to
. 10% hydrogen concentration under both positive and negative ambient pressure.
Clarification (1)
Design and qualification criteria are outlined in Appendix A.
(2)
The continuous indication of hydrogen concentration is not required during normat operation.
If an indication is not available at att times, continuous indication and recording shall be functioning uithin 30 minutes of the initiation of safety injection.
4 (3)
The accuracy and placement of the hydrogen monitore shalt be provided and justified to be adequate for-their intended function.
DPC RESPONSE LACBWR has installed two hydrogen analyzers Beckman Model 7C with 0-25%- hydrogen concentration ranges.
The hydrogen monitors will
~
receive their input of containment atmosphere from the containment post-accident sampling system which is to be operable by January.
1, 1982.
The instrument accuracy and calibration accuracy gives a combined accuracy of + 1 1/2%.
LACBWR utilizes stainless steel fuel cladding and therefore, has no potential for a Zircaloy fuel clad water reaction. 'Any hydrogen generated in the primary-system would normally be vented to the reactor containment building through the pipe break causing the LOCA or through the Manual Depressurization System Valves.
The LACBWR design -does not contain a suppression pool so the primary safety valves (3) and the Manual Depressurization System Valves (2) relieve directly into the reactor containment building.
The suction for the containment sampling system is at the 714 foot level near.the top of the reactor containment building which is at level 756'6".
-As any noncondensible (including hydrogen) would be vented directly into the open area of the reactor containment through either the safety valves or the Manual Depressurization valves, any hydrogen present would collect at the high point of
- WP1 4-1 4
y
...m.
,..,_s~.,,,_x,___c--m_,..-.-,,-7.r%-,
-.. ~. -,.., -. -..
.__,,.,..--.m_
, -..~
\\
containment and thus be drawn through the hydrogen analyzer.
The analyzers will normally be ~ energized.
However, they will not be connected to the reactor containment unless. an accident occurs, they can be functioning within 30 minutes of the event.
NRC CRITERIA Applicability To the extent feasible and practical (in conformance with the s tipulations of Appendix A and ancillary requiremente), equipment is to be installed by the specified implementation dates.
Where equipment is unavailable, precluding conformance uith equipment qualification and scheduled requiremente, the implementation dates are to be met by instattation of best available equipment.
In euch cases, deviationa are to be described and a schedule for the feasible instat tation of equipment in conformance with the stipulations of Regulatory Guide 1.97 (when the guide is used) is to be provided.
Appendix A is concietent uith our current draft version of Regulatory Guide 1.9 7.
We expect no further revisions to our requiremente.
Criteria (1).The instrumentation should be environmentatty qualified in accordance uith Regulatory Guide 1.89 (NUREG-0588).
Qualification applies to the complete instrumentation channet from sensor to display where the display is a direct
-indicating meter or recording device.
Where the instrumentation channet signal is to be used in a computer-based display, recording and/or diagnostic program, qualification applies to and includes the channet isolation device.
The tocation of the isolation device should be such that it vould be accessible for maintenance during accident conditions.
The seismic portion of environmental qualification should be in accordance'vith Regulatory Guide 1.100.
The instrumentation should continue to read uithin the required accuracy foltoving, but not necessarity during, safe shutdown earthquake.
Instrumentation, uhose ranges.
a a re required to extend beyond those ranges calculated in the moet severe design basis accident event for a given variable, should be qualified using the foltooing guidance.
The qualification environment shalt be based on the design basie accident events, except the aceumed maximum of the value of the monitored variable shatt'be the value equal to the maximum of the variable.
The monitored variable shall be assumed to approach chie peak by extrapolating the most severe initial ramp associated uith the design bacia accident evente.
The decay for this variable shall be considered WP1 4-2
s proportional to the decay for this variable ' associated with The qualification environment shall be based on the design baele accident evente, except the assumed maximum of the
-value of the monitored variable shalt be the value equal to the maximum of the variable.
The monitored variable shall be assumed to approach this peak by extrapolating the most severe initial ramp associated uith the design basie accident events.
The decay for this variable shall be' considered proportional to the decay for this variable associated with the design basis accident evente.
No additional qualificaiton margin needs to be added to the extended range.
variable.
Att environmental envelopes except that pertaining to the variable measured by the information display channet shall be those associated with the design basis accident evente.
The above environmental qualification requirement does not account for steady-etate elevated levele that may occur in other environmental parametere associated vith.the extended range variables.
For example, a sensor measuring containment pressure must be qualified for the measured procese variable range, but the corresponding ambient temperature is not mechanisticatty linked to that pressure.
Rather, the ambient temperature value is the bounding value for design basis accident evente anatyaed in Chapter 15 of the final safety analysis report (PSAR).
The extended range requirement is to ensure that the equipment oilt continue to provide information should conditione degrade'beyond those postulated in the safety analysis.
Since variable ranges are.
nonmechanisticatty determined, extension of associated parameter levels is not justifiable and hae, therefore, not been required.
DPC RESPONSE The reactor containment atmosphere sampling system including the hydrogren analyzer was constructed using the mounting techniques required to have the system withstand seismic events.
The environment where the hydrogen analyzers are mounted is mild in a post-accident situation resulting from a loss of coolant accident.
Originally facility design did not address the seismic issue as the plant was included in the Uniform Building Code, 1958 Edition, Zone 0.
The Systematic Evaluation Program seismic evaluation will be complete in the evaluation phase by August 1982 and any required modification by mid-1985.
This evaluation includes a review of equipment anchorages.
A review of environmental qualifications of safety related electrical equipment is being performed under the Systematic Evaluation Program.
WP1 4-3
e s
..=
It'is our conclusion that-the instrumentation is adequate now..
Should the SEP seismic and environmental' qualification programs
~
indicate the need for_ modification, it will be accomplished.
NRC CRITERxf (2)
No single failure within either the accident-monitoring.
instrumentation, ite aumillary ' aupporting features or -ite pouer sources concurrent with the. failure that are a condition or result of a specific accident should prevent the operator from being presented the information necessary for him to determine the safety statue of the plant and to bring the plant to a safe condition and maintain it in a cafe condition follouing that accident.
Where failure of one accident-monitoring channet results in ambiguity (that is, the redundant displaya disagree) which could lead the operator to defeat or fait to accomplish a required safety funciton,-additional infomvation should be provided to attou the operator to deduce the actual conditione in the plant.
This may be accomplished by:
(a) providing additional independent channets of information of the'eame variable
-(addition of an identical.channet), or (b) providing an independent channet uhich monitore a different variable bearing a knoun. relationship to the multiple channels (addition of a diverse channet), or (c) providing the capability, if sufficient time is available, for the operator to perturb the measured variable and determine which channet has failed by observation of the response on each instrumentation channet.
Redundant or diverse channete
.chould be electricatty independent, energized from station Class 15 power source, and physicatty separated in accordance with Regulatory Guide 1.75 up to and inotuding any isolation device.
At least one channet should be displayed on a direct indicating or recording device.
(NOTE:
Within each redundant division of a safety eyetem, redundant monitoring channete are not required).
DPC RESPONSE The instrument channels are supplied by essential power.
No requirement for separate power supplies was established.
The essential buses were constructed prior to the establishment of Class -lE power source requirements.
Cable separation does not comply with ' Regulatory Guide 1.75.
Isolation devices are not included as the only output of this signal is indication, no automatic actions occur.
WP1 4-4
r 4
.s WRC'CRTTERIA (3) 'The instrumentation should be energized from etacion Class 1E power sources.
DPC RESPONSE The -instrumentation is supplied power from an essential. power supply.
The." Class lE Power Source" - requirement was not part of
- the. original. plant design.
The seismic evaluation program is-
~
' currently in progress to determine what, if any, modifications are
~
required.
The power suppliesf as they exist have been accepted by
.the NRC to meet the Interim Acceptance Criteria for Emergency Core Cooling Systems.
NRC CRITERIA (4)
An instrumentation channel should be available prior.to an accident'except as provided in Paragraph 4.11, "Exemp tion, "
as defined in IEEE Std. 2 79 or as specified in technical specificatione.
DPC RESPONSE Both hydrogen analyzers are energized during operation but the system is not functional except during post-accident. operation.
NRC CRITERIA (S)
The recommendations of the following regulatory guides pertainir.g to quality assurance thould be followed.
1.28
" Quality Assurance Program Requiremente (Design and Contruction)"
1.30
" Quality Assurance Requiremente for the Installation, Inspection, and Testing of Instrumentation and Electrical Equipment" 1.38
" Quality Assurance Requiremente for Packaging, i
Shipping, Receiving, Storage, and-Handling _of Items for Watar-Cooled Nuclear Power Plante" 1.58
" Qualification of Nuclear Pover Plant Inspection, Examination, and Testing Personnel" 1.64
" Quality Assurance Requirements for the Design 'of Nuclear Power Plante" 4
~ WPl.
4-5
r 1.74
" Quality Assurance Terme and Definitions" 1.88
" Collection, Storage, and Maintenance of Tuclezr Pover Plant Quality Assurance Records" 1.123
" Quality Assurance Requiremente for Control of Procurement of Items and Services for Nuclear Power Plante" 1.144
" Auditing of Quality Assurance Programs for Nuclear Pouer Plante" Task
" Qualification of Quality Assurance Program Audit
.RS'810-5 Personnel for Nuclear Power Plante" (Guide number to be-inserted).
Reference to'the above regulatory guides (except Regulatory Guides 1.30 and 1.38) are being made pending issuance of a regulatory guide endorsing NQA-1 (Task RS 002-5), nov in progress.
DPC RESPONSE The containment hydrogen concentration indicators will be maintained under the provisions of the LACBWR Quality Assurance Program which has been approved by the'NRC.
The LACBWR QA Program is based on the requirements of 10CFR50, Appendix E, Regulatory Guide 1.33, which addresses the requirements of ANSI N18.7-1972' and ANSI N45.2-1971. No further modification of the
. Quality Assurance Program is anticipated at this time.
NRC CRITERTA (6)
Continuous indication (it may be by recording) display should-be provided at all times.
Where tuo or more instrumente are needed to cover a particular range, overlapping of instrument-span should be provided.
DPC RESPONSE Each hydrogen analyzer will indicate in the control room once the
-modification is complete.
Initially, each hydrogen analyzer will be read locally in the feed pump area in the Turbine Building.
The display of hydrogen concentration will not be activated unless a post-accident situation develops.
A single instrument range is sufficient.
- WP1 4-6 V
1 NRC CRITERIA (7)
Recording of instrumentation readout information ekould be provided.
Where trend or transient information is essential for operator information or action, the recording should be analog strip-chart or stored and displayed continuoucty on demand.
Intermittent dieptage, such as data toggere and scanning recordere, may be used if no significant transient response information is likely to be lost by such devices.
DPC RESPONSE No recording of containment hydrogen concentration is provided.
One channel is being added to the post-accident Technical Support Center / Emergency Operations Facility printout.
Continuous scanning will provide one minute averages of containment hydrogen concentration which will in post-accident situations be printed out each minute at both the Technical Support Center and the Emergency Operations Facility.
NRC CRITERIA (8)
The instrumente should be specificatty identified on the control panete so that the operator can easily discern that they are intended for use under accident conditions.
DPC RESPONSE The instruments will be identified on the control panel.
NRC CRITERIA (9)
The transmission of signate from the instrument or associated sensore for other use should be through isolation devices that are designated as part of monitoring instrumentation and that meet the provisions of the document.
DPC RESPONSE There currently are no output uses other than indication so isolation devices are not used.
WP1 4-7
~
~~
NRC CRITBRJk (10)- Meane ehould'be provided for checking, with a high degree of
~
confidence, the operational availability of each-monitoring n
channet, including its input sensor, during reactor operation.
This may be accomplished in various _vays; for example:,,
l-
~
(a).By pertubbing the monitored variable.
~
/- -
.c
' ~
c
/
(b)
By introducing and varying, as appropriate, a e u'b'e t'i tu t e a
input to the sensor of the same nature asIthe measured variable.
r (c)
By cross-checking betueen channets that bear'a-knoun relationship to each other and that have readoute-available.'
DPC RESPONSE The method of cross-checking between channeis is availabli at-LACBWR to verify operational availability.
~
i I
.J.
NRC CRITERIA (11)
Servicing, testing, and calibrating programe should be
~
~.
epecified to maintain the capability of the monitoring instrumentation. - Por thoseIinstrumente where the; required: -
interval between testing uitt be tese than the normat time interval between generating station shutdowne, a capability for testing during pouer operation should be provided.
~
DPC RESPONSE
~
The containment. hydrogen concentration indicators will be calibrated at least each refueling shutdown.
...O
/
.r-E l '
WP1 4-8 j
d
/
5 "C
~NRC CRITERIA (12)
Whenever meane-for removing channele from service are included
.in the design, the design should facilitate administrative-control of-the access to euch removat meane.
./
e.
mDP[ RESPONSE-
^
.c 3
Th'c contai6menp ' hydrogen concentration indicators can be removed f rom service under administrative control, if required.
As they do not activate 'anp' auxiliaries or equipment, no bypass features are require'd '..
1 0E fj Ij f RJ A (13)
The design should facilitate administrative control.of'the access to all setpoint adjustmente, module calibration adjust-
~~
' f mente, anbi test pointe.
r s
i
/
DOCo ESPONSE-The design, includes the analyzer in the turbine hall and the indicators in the control L room.
All equipment is accessible. The
~
Dut'y Shif t Supervisor exercises administrative control by.
Maintenance-nequest Forms over this equipment.
NRC CRITERLA (14)
Thejnonitoring instrumentation design should minimize the devstopetent of conditione that would cause meters,
,[
,annunciatore, recorders, alarme, etc., to give anomalous
_indicatione potentially confusing to the operator.
DPU RESPONSE m
~
The instrumentation is designed to minimize ' anomalous indication.
The indicator is simple and requires no conversion to interrupt.
NRC CRITERIA (15)
The instrumentation should be designed to facilitate the recognition, tocation, replacement, repair, or adjustment of malfunctioning componente or modules.
DPC RESPONSE The instrumentation is well designed for' repair and adjustment with a minimum number of components.
WP1 4-9 F
NRC CRTTERIA (16 ).
To the entent practical, monitoring instrumentation inpute should be from sensore that directly measure the desired variables.
DPC RESPONSE The sensors directly measure containment hydrogen' concentration.
NRC CRITERZk (17)
To the entent practical, the same instrumente should be used for accident monitoring as are used for the normai operations of the plant to enable the operator to use, during accident situatione, instrumente vich which the operator is most familiar.
However, where the required range of monitoring instrumentation results in a toes of instrumentation sensitivity in the normat operating range, separate instrumente should be used.
DPC RESPONSE Not applicable.
No measurable. hydrogen concentration is present'in the reactor containment during normal operation.
NRC CRITERIA (18)
Periodic testing should be in accordance with'the applicable portions of Regulatory Guide 1.118: pertaining to. testing of instrumente channete.
I DPC RESPONSE The calibration is performed at least each refueling outage.
1 i
WP1 4-10
ATTACHMENT 5 VERIFICATION OF OUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION SYSTEMS NUREG-0737, ITEM II.K.3.28
e-VERIFY OUALIFICATION OF ACCUMULATORS ON AUTOMATIC DEPRESSURIZATION SYSTEM VALVES NUREG 0737 ITEM II.K.3.28 NRC POSITION & CLARTFICATION Position Safety analysis reporte claim that air or nitrogen accumuttaore for the automatic deprnacurination system (ADS) vatoce are provided with sufficient capacity to cycle the valves open five times at design pressures.
GE has also stated that the emergency core cooling (ECC) systems are designed to uithstand a hostile environment and still perform their function for 100 days follouing an accident.
Licensee should verify that the accumulatore on the ADS valves meet these requiremente, even considering normat teakage.
If this cannot be demonstrated, the licensee must show that the accumulator design is still acceptable.
Clarification The ADS valves, accumulatore, and associated equipment and instrumentation must be capable of performing their functione during and following emposure to hostile environmente and taking no credit for nonsafety-related equipment or instrumentation.
A dditionally, air (or nitrogen) teakage through valves must be accounted for in order to assure that enough inventory of compressed air is available to cycle the ADS valves.
DPC RESPONSE LACBWR ha a Manual Depressurization System (MDS) rather than an Automatic Depressurization System.
The MDS System consists of two 4" parallel control valves in the condensate return line from the shutdown condenser.
The two MDS valves are located on the shutdown condenser platform at an elevation of approximately 712' in the north end of the Reactor Containment Building.
The two valves vent directly to the containment upon actuation.
The MDS valves are diaphragm-operated globe-type control valves which require 30 psi of nitrogen to close and are spring loaded to open on a loss of nitrogen.
Nitrogen is admitted to each valve operator through individual solenoid valves which must be e ne rg ized to release the closing pressure thus opening the associated vent valve.
Operation of the individual solenoid valves is from the control room by key locked switches.
LOCA environment solenoids and control c able are presently scheduled for upgrade completion in accordance with LACBWR's ef forts to obtain electrical environmental qualification of safety related electrical equipment.
WP1 5-1
It can be considered that the nitrogen supply system to the MDS valve diaphragm accuators constitutes an accumulator system.
This nitrogen supply system consists of two standard DOT 3AA2400 Nitrogen cylinders, a Bastian-Blessing Type 18230580-2-E Duplex Manifold, and copper tubing all of which. are located in the Reactor Containment Building between the 701' and 716' elevations
.on the north side of the containment.
These nitrogen supply components cannot be adequately demonstrated to sustain a LOCA environment above a temperature of 165'F for a duration of up to 100 days.
However, a failure of. the. system would result in a loss of nitrogen gas pressure which would position.the vent valves in the open position, thus providing the venting capability, if so needed.
Venting would then be controlled through the inlet valves of the shutdown condenser.
In addition, an event that would result in a temperature -rise situation above 165'F would indicate a failure approximating the conditions subsequent to a MDS actuation.
In view of the above considerations, a loss of nitrogen control to the MDS valve operators due to supply system failure will not result in an unsafe condition during and subsequent to an accident.
LOCA qualification of the nitrogen system is, therefore not judged necessary.
WP1 5-2
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