ML20009D447
| ML20009D447 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 07/22/1981 |
| From: | Mccaffrey B LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, TASK-2.B.2, TASK-2.B.4, TASK-2.F.1, TASK-3.D.3.3, TASK-TM SNRC-602, NUDOCS 8107240066 | |
| Download: ML20009D447 (38) | |
Text
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LONG ISLAND LIGHTING COMPANY g
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SHOREHAM NUCLEAR POWER STATION P.O. BOX 604, NORTH COUNTRY ROAD e WADING RIVER, N.Y.11792 July 22, 1981 SNRC-602 Mr. Harold R. Denton Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 SHOREHAM NUCLEAR POWER STATION - Unit 1 Docket No. 50-322
Dear Mr. Denton:
Enclosed herewith are sixty (60) copies of LILC0 responses to specific NRC concerns which were previously identified as requiring additional information to complete NRC review. Attachment A provides a list of the specific responses included.
If you require additional information or clarification, please do not hesitate to contact this office.
I Very truly yours, kWWh#
B. R. McCaffrey Manager, Project Engi ering Shoreham Nuclear Power Station Enclosures g
cc:
J. Higgins 8107240066 810722
^
PDR ADOCK 05000322 E
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SNRC-602, July 22, 1981 Attachment A Additional information is provided for the following items:
1
- 1) NUREG-0737 Item II.B.2 - Plant Shielding
- 2) NUREG-0737 Item II.B.4 - Training for Mitigating Core Damage
- 3) NUREG-0737 Item II.F.1 - Accident Monitoring Instrumentation (Attachments 1, 2, 3, and 5)
- 4) NUREG-0737. Item III.D.3.3 - Radiation Monitoring
- 5) SER Open Item No. 37 - Secondary Containment Bypass Leakage
- 6) SER Section 6.3.1 - HPCI/RCIC 50% Plugged Strainer Test 1
I
SNPS-1 FSAR II.B.2 Plant Shieldina. to Provide Access to vital Arean and Protect Safety Equipment for Post-Accident Oneration NRC Position With the assumption of postaccident release of radioactivity equivalent to that described in Regulatory Guides 1.3 and 1.4 (i.e.,
the equivalent of 50 percent of the core radioiodine, 100 percent of the core noble gas inventory, and 1 percent of tne core solids are contcined in the primary coolant), each licensee shall perform a radiation and shielding-design review of the spaces around syste:.3 that
- may, as a result of an accident, contain highly radioactive materials.
The design review should identify the location of vital areas and equipment, such as the control
- room, radwaste control
- stations, emergency power
- supplies, motor control centers, and instrument areas, in which parsonnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during postaccident operations of these systems.
Each ' licensee shall provide for adequate access to vital areas and protection of safety equipment by design
- changes, increased permanent or temporary shielding, or postaccident procedural controls.
The design review shall determine which types ot corrective actions are needed for vital areas throughout the facility.
The purpose of this item is to ensure that licensees examine their plants to determine what actions can be taken over the short-term to reduce radiation levels and increase the capability of operators to control and mitigate the consequences of an accident.
These actions should be taken pending conclusicns resulting in the long term degraded core rulomating, which may result in a need to consider additional sources.
Any area which will or may require occupancy to permit an operator to aid in the mitigation of or recovery Irom an accacent is designated as a
vital area.
For the purposes of this evaluation, vital areas and equipment are not necessarily the same vital areas or equipment defined in 10 CIR 73.2 for security purposes.
The security center is listed as an area to be considered as potentially vital, since access to this area may be necessary to take action to give access to other areas in the plant.
The control
- room, technical support center (TSC),
sampling station and sample analysis area must be included among those areas where access is considered vital after an accident.
(See Item III.A.1.2 for discuscion of the TSC and emergency opo-tions facility).
The evaluation to determine the necessary vital areas should also include, but not be limited to, censideration or the post-LOCA hydrogen control system, containnent isolation rese-control area, manual ECCS alignment area (if any), motor control
- centers, instrument
- panels, emergency power supplies, security II.B.2-1
-n
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SNPS-1 FCAR
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center, and radwaste control panels.
Dose rate determinations need not be for these areas if they are determined not to be vital.
As a
- minimum, necessary modifications must be sufficient to provide for vital system operation and for occupancy of the control room, TSC, sampling station, and samplo analysis area.
' In order to assure than personnel can perform necessary postaccident operations in the vital
- areas, the following guidance is to be used by licensees to evaluate the adequacy of radiation protection to the operators:
f,1)' Source Term
'2hc minimum radicactive source term should be equivalent to the source terms recommended in Regulatory Guides 1.3, 1.4, and 1.7 and Standard Review Plan 15.6.5 with appropriate decay trmes based on plant design (i.e., you may assume the radioactive decay that occurs before tission. products can De transported to various systems).
(a)
Liquid-Containing Systems:
100 percent of the core equilibrium noble gas inven' ory, 50 percent of the core equilibirum halogen invc cory, and 1 percent or all others are assumed to be mAxed in the reactor coolant and liquids recirculated by residual heat removal (IULR),
high-pressure coolant injection (LPCI), and low pressure coolant injection (LPCI),
or the equivalent of these systems.
In determining the source term for recirculated, depressurized cooling
- water, you may assume that the water contains no noble gases.
(b)
Gas-Containing Syt.
100 percent of the core equilibrium noble gas 11.s-.ttory and 25 percent of the core equilibirum halogen activity are assumed to be mixed in the containment atmosphere.
For vapor-containing lines connected to the primary system (e.g.,
BNR steam lines),
the concentration of radioactivity shall be determined assuming the activity is contained in the vapor space in the primary coolant system.
(2) Systems Containing the Source Systems assumed in your analysis to contain high levels of radioactivity in a poataccident situation should include, but not be limited to, containment, residual heat rencval system, safety j
injection systems, chemical and volume control system (CVCS),
containment spray recirculation
- system, sample lines, gaseous radwaste
- systems, and standby gas treatment systems (or equivalent of these systems).
If any of these systems or others that could contain high levels of radioactivity,were
- excluded, i
you should explain why such systems were excluded.
Radiation from leakage of systems located outside of containment need not t
II.B.2-2
SNPS-1 FSRR be considered for this analysis.
Leakage measurement and reduction is treated under Item III.D.1.1 " Integrity of Systems Outside Containment Likely To Contain Radioactive Material for PWRs and BHRs".
Liquid waste system need not be included in this analysis.
Modifications to liquid waste systems will be considered after completion of Item III.D.1.4, "Radwaste System
. Design Features to Aid in Accident Recovery and Decontamination".
~ (3) Dose Rate Criteria The design doce rate for personnel in a vital area should be such that the guidelines of GDC 19 will not be exceeded during the course et the accident.
GDC 19 requires that adequate radiation protection be provided such that the dose to personnel should not be in excess of 5 rem whole body, or its equivalent to any part of the body for the duration of the accident.
When determining the dose to an
- operator, care must be taken to determine the necessary occupancy times in a specific area.
For example, areas requiring continuous occupancy will require much lower dose rate than areas where minimal occupancy is required.
Therefore, allowable dose rates will be based upon expected occupancy, as well as the radioactive source terms and shielding.
- However, in order to provide a general design objective, we are providing the following dose rate criteria with alternatives to be documented on a case-by-case bases.
The recommended dose rates are average rates in the area.. Local hot spots may exceed the dose rate guidelines.
These doses are design objectives and are not to be used to limit access in the event of an accident.
(a)
Areas Requiring Continuous Occupancy:
<15 mrem /hr (averaged over 30 days).
These areas will require full-time occupancy during the course of the accident.
The control room and onsite technical support center are areas where continuous occupancy will be required.
The dose rate for these areas is based on the control room occupancy factors contained in SRP 6.4.
(b)
Areas Requiring Infrequent Access:
GDC 19. These areas may require access en an irregular basis, not continuous occupancy.
Shielding should be provided to allow access dt a frequency and duration estimated by the licensee.
.The lant radiochemical / chemical analysis laboratory, radwaste panel, motor control
- center, instrumentation locations, and reactor coolant and containment gas sample stations are examples of sites where occupancy may be needed often, but not continuously.
(4) Radiation Qualification of Safety-Related Equipment i
The review of safety-related equipment which may be unduly degraded by radiation during postaccident operation of this equipment relates to equipment inside and outside of the primary containment.
Radiation source terms calculated to determine II.B.2-3
SUPS-1 FSAR i
environmental qualification of safety-related equipment consider the following:
(a)
LOCA events which completely depressurize the primary system should consider releases of the source term (100 percent noble
- gases, 50 percent
- lodines, and 1 percent particulates) to the containment atmosphere.
(b)
LOCA events in which the primary systen may not depressurize should consider the source term (100 percent noble
- gases, 50 percent
- iodines, and 1 percent particulate) to remain in the primary coolant.
This method is used to determine the qualification doses for equipment in close proximity to recirculating fluid systems inside and outside of contaiment.
Non-LOCA events both inside and outside of containment should use 10 percent noble
- gases, 10 percent
- iodines, and 0 percent particulate as a source term.
The following table summarizes these considerations:
Contailuaent LOCA Source Tern Non-LOCA (Noble Gas / Iodine /
High-Energy Line Break Source Term Particulate)
(Noble Gas / Iodine / Particulate)
Outside (100/50/1)
(10/10/0) in RCS in RCS Inside Larger of (10/10/0)
(100/50/1) in RCS in containment y
(100/50/1) in RCS LILCO Position Areas where access is vital after an accident have been analyzed for post accident personnel access.
The Shoreham position is that access is only needed to the control room, the Technical Support Center (TSC), and the Post Accident Sampling and Analysis Faci.'ity (PASF).
The other areas suggested as vital post accident in NUREG-0737 either do not apply for Shoreham or are not needed.
The hydrogen recombiner system is controlled remotely from the control roam.
The containment isolation reset control area is also located in the main control room.
Shoreham l
has no manual ECCS alignment area; all vital ECCS valves are automatic and operable remotely from the control room.
Motor control centers do not need to be accessed, nor do the instrument panels located outside of the control room.
No operability access is needed for emergency power supplies since they are II.B.2-4
S14PS-1 FSAR remotely operable from the control room.
The radwaste control panels are not needed for accident mitigation since they control no safety related functions.
Access to the Security Center is not needed to gain access to the control room, TSC or PASF.
Since the control room is always manned, entrance could be gained manually at any time, should automatic security systems fail.
Access to the PASF can be gained from any of the Emergency Operations facilities without passage through a building containing high radiation areas.
Keys are available to authorized personnel to manually open the doors to the TSC or PASF should the automatic security systems fail.
The control room and the permanent TSC are both habitable (less than 15 mrem /hr) (30 day avg) in a post-accident environment.
The PASF is designed for limited habitability to the extent necessary to obtain and analyze samples per the specific requirements of Item II.B.3.
The PASF habitability is prirarily affected by the LOCA cloud around the facility which is drawn in through the fil-tered intake.
The worst-case gamma dose rate for this case will occur at about t=8 hours and will be less than 100 mrem /hr within the manned area of the facility.
The worst case accident which impacts the normal radiochemistry laboratory (in the turbine building) habitability is the control Rod Drop or the failure of the air ejector lines.
Both of these increase the airborne hazard in the laboratory to higher levels than a LOCA.
They will result in gamma dose rates of approximately 200 mrem /hr at the worst part of the accident.
Detailed radiation calculations were performed to ensure adequate environmental qualification of safety related equipment within the harsh, post accident environment of the reactor building.
SOURCE TERM l
Radioactive sourcc release and distribution assumptions for Shoreham are as follows:
Radioactive Source Release 1.
The percentages of core inventory radioactive fission products assumed to be released from the fuel rods are:
Noble gases (Kr, Xe) 100%
Iodine 50%
Others 1%
2, This entire release is assumed to occur instantaneously at the start of the accident.
hcdioactive Source Distribution In order to envelope the full spectrum of break sizes and depres-surization rates, two bounding events and source distributions were considered.
II.B.2-5
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SNPS-1 FSAR 4
l.
LOCA -(both pressurized and depressurized events)
'The following fission products are considered to be uniformly mixed in the following volumes:
a.
Suppression Pool Noble gases 0%
Iodine 50%
Others 1%
b.
Combined Drywell/Wetwell Air Space
~
-Noble gases 100%
Iodine 25%.
Others 1%
c.
. Reactor Coolant System Steam Space The following distribution is used for determining reactor building pipe shine doses due to HPCI, RCIC and MSIV-LCS operation.
Noble gases 100%
Iodine 25%
Others 0%
Using the above distribution, time history radiation zones were established within the primary containment
-and within the-secondary containment as follows:
1.
Primary Containment - Previously described total integrated dose accident radiation levels (Table 3.11.2-1) adequately bound all LOCA events for equipment within the primary containment and no new analyses were required.
ii.
Secondary Containment - Time history radiation zones were established for the secondary con-tainment'using'the above sources distributed in the steam and liquid piping in the following fluid systems which were conservatively assumed to' operate concurrently:
4 1.
HPCI 2.
RCIC 3.
RHR (all essential modes) 4.
Core Spray 5.
RBSVS 6.
MSIV-LCS 7.
PCAC in addition to radiation shine from the above
=
system piping and components, the primary con-tainment was assumed to leak at technical speci-fication limits 1
II.B.2-6
i SNPS-1 FSAR resulting in an airborne source term which was included in the radiation zoning.
As provided by NUREG-0737, no additional leakage was assumed.
iii. Excluded Systems a.
All-piping whic'. could potentially carry undiluted reactor coolant into the secondary containment is isolated and is nonessential (e.g., RWCU, shutdown cooling mode of.RHR).
Accordingly, the undiluted reactor coolant liquid source
~
discussed above was excluded.
Adequate means are provided using the post accident sampling system to ensure safe coolant activity levels exist prior to use of any of these nonessential systems.
~
b.
The post accident sampling lines were excluded from specific evaluation due to their size (typically 3/8 in tubing) and because they are flushed after each use.
Enough conservatism in integrated dose calculations exists to bound any effects of these small sources.
c.
The gaseous radwaste system lines were also excluded because they were not sources for any safety related equipment.
They are also located outside the reactor building and are isolated.
Pipe break inside containment and nonpipe break events are bounded by the LOCA event above.
2.
Pipe _ Break in Secondary Containment As specified in NUREG-0737, secondary containment airborne time history dose and dose rates were established for this event using the following fission products uniforr.ly mixed in the primary coolant system:
Noble gases 10%
Iodine 10%,
Others 0%
4 I
II.B.2-7
SNPS-1 FSAR Reactor. building radiation zones and time history dcsc and dose rate data from the above analyses are shown on Figs. II.B.2-1 through 17.
The above radiation conditions are being used in conjunction with other environmental conditions
-(pressure, temperature, and humidity) for the equipment qualification. program.
Safety related equipment is being qualified in accordance 'iith NUREG-0588.
II.B.2-8
D
.9 9
4 W
FIGURES FOR ITEM II.B.2 REMAIN UNCHANGED.
REFER TO SNRC-563 DATED 5/15/81 FOR FIGURES.
--e,,-
-II.B.4 - Training For Mitigating Core Damage - Additional Information NRC Position:
Licensees are required to develop a training program to teach the use of installed equipment and systems to control or mitigate accidents in which the ccre is severely damaged. They must then implement the training program.
LILCO Response: The program for Training For Mitigating Core Damage, as outlined in our previous response to this issue (See SNRC-579 dated May 29,1981), will be in compliance with the GUIDELINES FOR TRAINING TO RECOGNIZE AND MITIGATE THE CONSEQUENCES OF CORE DAMAGE from The Institute of Nuclear Power Operations, Document Number STG-01, Rev. 1, dated January 15, 1981. The detailed course description is currently under preparation and will be submitted for NRC Staff review by October 1, 1981.
Ihe Shoreham Shift Technical Advisors and operating personnel, from the Plant Manager through the operations chain down to the licensed operator level will receive all of the training indicated in En' closure 3 to Mr. H. R. Canton's March 28, 1980 clarification letter.
Supervisory personnel and technicians in the Instru-mentation and Control (I&C), health physics, and chemistry sections will recieve the necessary training commensurate with their responsibilities.
II.B.4-3 7/23/81
SNPS-1 FSAR j
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II.F.1 Additional Accident Monitoring Instrumentation Introduction Item II.F.1 contains the following subparts:
1.
Noble gas effluent radiological monitor; 2.
Provisions for con tinuous sampling of plant effluents for postaccident releases of radioactive iodines and
~~
particulates and onsite laboratory capabilities (this requirement was inadvertently omitted from NUREG-0660; see Attachment 2 that follows, for position);
3.
Containment high-range radiation monitor;
- 4..
Containment pressure monitor; 5.
Containment water level monitor; and 6.
Containment hydrogen, concentration monitor.
Attachments 1 through 6 present the NRC position on these matters.
It is important.that the displays and controls added to the control room as n' result of this requirement not increase the potential for operator error. A human-factor analysis should be performed taking into consideration:
1.
the use of this information by an operator during bc ah normal and abnormal plant conditions, 2.
integration into emergency procedures, 3.
integration into operator training, and 4.
other alarms during emergency and need for prioritization of alarms.
II.P.1-1
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SNPS-1 FSAR NOBLE GAS EFFLUEST MONITOR ATTACm1EST 1 NRC Position effluent monitors shall be installed with an extended Noble gas range designed to function during accident conditions as well as during normal operating conditions.
Multiple monitors are considered necessary to cover the ranges of interest.
~~
1.
Noble gas effluent monitors with an upper range capacity of 10 Ci/cc (Xc-133) are considered to be practical 2,.'
5 and should be installed in all operating plants.
.2.
hJ'le gas ef fluent monitoring shall be provided for the r
total range of concertration extending from normal condition (as low as reasonably achievable (ALARA))
5 concentrations to a maximum of 10 Ci/cc (Xe-133).
Multiple monitors are considered to be necessary to cover the ranges of interest.
The range capacity of individual monitors should overlap by a factor oi ten.
Licensees shall provide continuous monitoring of high-leve'1, postaccident releases of radioactive noble gases from the plant.
Gaseous effluent monitors shall meet the requirements specified in the attach Table II.F.1-1.
Typical plant effluent pathways to be conitored are also given in the tabic.
The monitors shall be capable of functioning both during and following an accident.
System designs shall accommodate a design-basis release and then be capable of following decreasing concentrations of noble gases.
9 Offline moaitors are not required for the PWR secondary side main steam safety valve and dump valve discharge lines.
For this application, externally mout.ted monitors viewing the main steam line upstream of the valves are acceptable with procedures to correct for the low energy gammas the external monitors would not detect.
Isotopic identification is not required.
Instrumentation ranges shall overlap to cover the entire range of effluents from normal (ALARA) through accident conditions.
The design description shall include the following information.
1.
System description, including:
a.
Instrumentation to be
- used, including range or sensitivity, energy dependence or response, calibration frequency and technique, and vendor's model number, if applicable; b.
monitoring locations (or points of sampling), including deceription of methods used to assure representative measurements and bcckground co.rection; II.F.1-2 r,
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SNPS-1 FSAR r.
- c.
location of instrument readout (s) and method of recording, including description of the method or
{C procedure for transmitting or disseminating the information or data; 4
d.
assurance of the capability to obtain readings at least every.15 minut.:s during and following an accident; and 4
c.
the source of power to be used.
2.
Description of procedures or calculational methods to be used 2,.'
-for. converting instrument readings to release rates per unit time, _ based on exhaust air flow and considering radionuclide spectrum distribution as a function of time after shutdown.
LILCO Position The effluent monitor categories in Table II.F.1-1, which apply to the Shoreham Nuclear Power Station are:
(a)
"BWR reactor building exhaust air",
(b)
"other release points",
and (c) gases".
" buildings with systems containing primary coolant or See Fig. 'II.F.1-1 for a simplified diagram of Shoreham's gasechs effluent layout.
The maximum anticipated primary containment leakage rate is 0.005 5
volumes per day (volume of primary containment is 1.93x10 cu ft) into the secondary containment which has a volume of 2x10' cu ft.
(
The primary containment Icakage is highly diluted in the L'
secondary containment atmosphere.
This mixture will be discharged after passing through high efficiency particulate absolute filters and charcoal absorber banks via the reactor building standby ventilation system'(RBSVS) disc.harge pipe, at the top of 'the station vent exbaust.
Two Class IE radiation monitors (RE-021 and 022) serve this system dcunstream of the filters and adsorbers along with a post accidert Class 1E monitor (RE-134), which is added to the system for higher ranges.
The RBSVS monitors are supplied with power from vital instrument buses. These monitors read out in the control room and are located in the control building to permit access during an accident for collection of their radiciodine and particulate sample media for laboratory analysis.
The ' criteria in Table II.F.1-1 for other release points and buildings with systems containing primary coolant or gases are applicabic to the station vent exhaust monitor (RE-042) and the station vent post accident high range monitor (RE-126).
Normal ventilation discharges from the reactor building, che turbine 4
building =, and the radwaste building are mixed, thereby providing dilution prior to being exhausted through the station vent exhaust. When RBSYS is operating, and the reactor building normal ventilation system (RBSYS) is isolated, the loss of normal reactor building ventilation flow is compensated by opening
()
louvers at the, station-vent exhaust to permit 90,000 cu ft/ min of II.F.1-3 k
SNPS-1 FSAR outside air for dilution and to maintain a constant air velocity through the station vent.
This single discharge point for the combined vent ilat ion flow from all potentially contaminated buildings is monitored by a noble gas radiation monitor (RE-042) and post accident high range ef fluent rudiat ion monitor (RE-126).
The monitor (RE-042) is supplemented by in-line RE-069 with a high upper range.
In addition, the individual building ventilation flows to the station vent exhaust are each analyzed by a high range in-line radiation monitor (RE-066, 067, and 063).
All these monitors, except RE-042 and 126, are powered from a
~~-
vital instrument bus.
Where practical, initial calibration includes detector response for a minimum of three decades using standard sources of three dif ferent energies and intensities.
These calibration curves are initially generated using both gaseous. ' solid sources, where practical.
Routine calibration of these monitors is in accordance with technical specifications provisions using solid sources related to the initial calibration.
Calibration sources used are Sr-90, Cs-137, and Co-60 f or low ra- ;c monitors, and CS-137 for high range monitors.
The conversion of the instrument readings to release rates are determined using the energy response of the detectors obtained during calibration. Accident release rates are then calculated based on anticipated radionuclide inventories following a design basis loss of coolant ac c id ent.
Actual releases nay be determined by analyzing a grab cample and correcting the release rate calculated.
Continuous strip chart recording and CRT dis-play are provided in the control room.
Digital readout for the high range effluent monitors RE-126 and 134 will assure the
' availability of continuous reading in the control room during or after an accident.
The effect of background radiation on readings of the RBSVS noble gas monitors (RE-021, 022, and 134) will be minimized during an accident, due to their location in the control building and the detector's location in a 4n lead shield.
For the station vent exhaust monitor (RE-042 and 126),
background radia t ion in the vicinity of the monitor within the secondary containment will have minimal effect on the noble gas detector, due to its location in a 4n lead shield and the fact that the detector is a thin beta scintillator.
This type of detector is very inefficient for detecting ganma radiation which might penetrate the lead shield, while it is efficient for detecting the beta radiation associated with the sample stream's noble gases brought in close contact with the detector.
For a listing of the radiation monitors with the rangen provided, refer to Table II.F.1-4.
II.F.1-4
SNPS-1 FSAR KITACIIt1ENT 2 SA!!PLING AND ANALYSIS OF PLANT EFFLUENTS NRC Position
.C.
Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for effluent monitoring of radiciodines for the accident condition shall lua provided with sampling conducted by adsorption on charcoal or other media, folicwed by onsite laboratory ar.alysis.
Licensees shall' provide continuous sampling of plant gaseous effluent for postaccident. releases of radioactive iodines and particulates to meet the requirements of the enclosed Table II.F.1-2.
Licensees shall also provide onsite laboratory capabilitics to analyze or measure these samples.
This requirement should not be construed to prohibit design and development of radiciodine and particulata monitors to provide online sampling and analysis for the accident' condition.
If gross gamma radiation measurement techniques are used, then provisions shall be made to mi.nimize noble gas interference.
The shiciding design basis is given in Table II.F.1-2.
The sampling s; stem design shall be such that plant personnel could remove samples, replace sampling media and transport the sampics to the onsite analysis facility with radiation exposures that are not in excess of the criteria of GD0 19 of 5-ren whole-body exposure and 75 rem to the extremitics during the duration of the
- ~
accident.
The design of-the systems for the sampling of particulates and lodines should provide for sample nonale entry vclocities which are approximately isokinetic (same velocity) with expected induct or instack air velocitics.
For accident conditions, sampling may be complicated by a reduction in stack or vent effluent velocitics to below design IcVels, making it necessary to substantially reduce sampler intake flow rates to achieve the isokinetic condition.
Reductions in air flow may well be beyond the capability of availabic sampler flow controllers to maintain isokinetic conditions; therefore, the staff will accept flow control devices which have the capability of maintaining isokinetic conditions with variations in stack or duct design flow velocity of i 20 percent.
Further departure from the isokinetic condition need not be considered in design.
Corrections for non-isokinetic sampling conditions, as provided in Appendix C of ANSI 13.1-1969 may be considered on an ad hoc basis.
Effluent streams which may contain air with entrained water, le.g., air ejector discharge, shall have provisions to ensure that the adsorber is not degraded while providing a representative
- sampic, e.g., heaters.
O II.F.1-5
a l
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SNPS-1 FSAR 7
LILCO Position
' The normal station vent exhaust monitor (RE-042) is not pouered from'a vital instrument bus,- however, it is powered from a
- dependable backup power supply to normal ac.
Due to its location in the secondary containment, it may be inaccessible daring an
~
accident.
This would preclude obtaining the radioiodine and particulate-sample media f rom the monitor for analysis. However, "J.
inability to obtain these samples is compensated for by the fact that -the turbine building and radwaste building ventilation flows are each sampled for. radioiodine and particulates by the equipment _ associated with the normal range noble gas monitors for these flows (RE-057 and 055). These monitors are both located in the turbine building, peruitting access for collection of the sample media during an accident in order that laboratory analysis may be performed. Adding the results obtained for radioiodine or particulates from the turbine building and radwaste building ventilation flows will give the radiciodine or particulate release at the station vent exhaust should the secondary containment be inaccessible. Under these circumstances, RBSVS is operating and there is no reactor building ventilation contribution to the station vent exhaust. As discussed
- above, the RBSVS release is monitored separately for noble gases and continuous collection of camples for particulates and radioiodine releases _(RE-021, 022, and 134).
These monitors are capable of representative monitoring and sampling for all accident cond*
ons except for pipe break outside containment (refer to Appe* dix 3C). The monitors associated with the reactor, radwaste ano LuJoine buildings ventilation systems are not powered from a
' ital bus. -This is consistent with the design of the monitored
- v systems. The station vent exhaust monitors (RE-042 and 126) radioiodine and particulate sample media can be obtained for analysis if the secondary containment is accessible.
The addition of the high range station ventilation exhaust monitor (RE-126) assures continuous sampling of radioiodine and particulates during accident conditions. Continuous sampling is achieved with isokinetic sampling during normal operation and sampling probes during accident conditions. Provisions have been made to comply with ANSI N13.1-1969 to the maximum extent Epractical to assure representative sampling.
The sampling collector will initiate an alarm in the control room when it 2
a concentration of 10 ACi/cc and 30 min collection time.
reaches At this time the microcomputer associated with RE-126 transfers the flow to the next particulate and iodine assembly, isolates the alarmed assembly, and indicates to the operator the need to replace the collector assembly and transfer it to the laboratory for analysis.
The sampling media is paper with more than 90 percent collection ef ficiency for 0.3 micron particles and a charcoal cartridge with more than 90 percent collection efficiency for methyl iodide.
f II.F.1-6 I
--a
~
SNPS-1 FSAR 1
The radioio/.ine and particulate sampling nedia is analyzed in the counting room at Shoreham.
Charcoal cartridges are purged with nitrogen or air to remove entrapped noble gases. A separate counting station is provided which serves as a backup for the counting facility in the radiochenistry laboratory. At least one of these locations will remain a low-contamination, low-background area for all postulated accident conditions.
Tha above meets the requirements of Table II.F.1-2.
Further, procedures will be prepared fo.? conducting all aspects of the measurement and analyses correctly and in a manner to minimize personnel exposure.
e II.F.1-6a i
SNPS-1 FSAR ATTACllMENT 3 CONTAINMENT llIGIl-RANGE RADIATION MONITOR
-NRC Position In containment radiation-level monitors with a maximum range of 8
10 Lrad/hr shall be installed. A minimum of two such monitors 7.ha t are physically separated shall be provided. Monitors shall be developed and qualified to; func t ion in an accident environment.
Provide two radiation monitor systems in containment which are documented to meet the requirements of Table II.F.1-3.
The - specification of 108 rad /hr in the above position was based onla circulation of postaccident containment radiation levels that included both particulate (beta) and photon (gamma) radiation. A radiation detector that responds to both beta and gamma radiation cannot bc qualified to post-LOCA (loss-of-coolant accident) containment environments but gamma - sensitive instruments can be so qualified.
In order to follow the course
.of an accident, a containment monitor that measures only gamma radiation is adequate.
The requirement. was revised in the October.30, 1979 letter to provide for a photon-only measurement 7 R/hr.
with an upper range of 10 Tbc monitors shall be ' located in containment (s) in a manner as to provide a reasonabic assessment of area radiation conditions inside containment. The monitors shall be widely separated so as to provide independent measurements and shall "vicu" a large fraction of the containment volume.
Monitors should not be placed in areas which 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 dif ficulties.
For BWR Mark III containments, two such monitoring systems should be inside both the primary containment (drywell) and the secondary containment.
. Th'e monitors are required to respond to gamma photons with energies as low as 60 kev and to provide an essentially flat response for gamma energies between-100 kev and 3 McV, as specified in Table II.F.1-3.
Monitors that use thick shielding to increase the upper range will under-estimate postaccident radiation levels in containment by several orders of magnitude because of their insensitivity to low energy gammas
,d are not acceptable.
LILCO Position Two physically separate monitors are located inside the drywall,
- El 78'-7",
fee photon radiation.
One is located adjacent to the equipment hatch and the ot)er being adjacent to the personnel hatch, (1800 separation). See Fig. II.F.1-2.
These locations II.F.1-7
~
SNPS-1 FSAR have been selected to provide an unobstructed, large view of the containment volume, and to ensure case of access for replacement, maintenance, and calibration.
Calibration will be performed during routine refueling outages.
These monitors are each powered by a vital instrument hus, are seismic qualified, and are designed to withstand the temperatures, pressures, humidity, and total radiation in the dryw
.1 contain-ment through an a cc id ent.
Monitor readouts are displayed
~'-
Ca tegory I panel in the main continuously and recorded on n control room.
These monitors provide unshielded, unattenuated containment. rad ia t ion readings during an accident and neet the requirements of Table II.F.1-3.
For a listing of the radiation monitors with the ranges provided, refer to Table II.F.1-4.
a II.F.1-8
~
SNPS-1 FSAR
-A'ITACIDIENT 4 CONTAINMENT PRESSURE MONITOR NRC Position A continuous indication of containment pressure shall be provided in the control room cf each operating reactor.-
Measurement and indication capabili.y shall include three times the design pressure of the containment for concrete, four times the design pressure for steel, and -5 psig for all containments.
Design and qualification criteria are outlined in Appendix B to NUREG-0737.
Heasurement and indication capability shall extend to 5 psia for subatmospheric containments.
TWo or t ;*e instruments may be used to cect requirements.
However, instruments that need to be switched from one scale to another scale to meet the range requirements are not acceptable.
Continuous display and recording of the containment pressure over the specified range in the control room is required.
The accuracy and response time specifications of the pressure monitor shall be provided and justified to be adequate for their intended function.
LILCO Position
' Currently installed instrumentation provides continuous display
' and recording of containment pressure in the control recm.
- Pressure transmitters and associated instrumentation have been replaced in order to provide the capability to measure three times the design pressure of the primary containment.
The range of pressure instrumentation is from
-5 to +150 psig.
The pressure transmitters have an accuracy of 0.25 percent of span and at 100 F the response times are 0.2 sec (63 percent of the The components provided meet the design criteria outlined time).
in Appendix B to NUREG-0737 to the maximum extent practical.
e 9
II.F.1-9
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KITAClit!ENT 5 CONTAIN.'!EST VATER LEVEL !!OSITOR
~
NRC Position s
A continuous indication o f. containment water level shall be provided in the control-room for all plants.
A narrow rangn instrument shall be provided for PWRs and cover the range from the bottom to the top of the containment sump.
A vide range instrument shall also be provided for PWRs and shall cover the range from the bottom of the. containment to the elevation equivalcat.to a 600,000 gallon capacity.
For BURS, a wide rang _
instrument shall be provided and cover the range from the bottom
-to 5 feet above the normal water Icyc1 of the suppression pool.
ig The containment wide-range water IcVel indication channels shall meet the. design and qualification criteria as outlined in Appendix. B to NUREG-0737. The narrow-range channel shall meet the requirements of Regulatory Guide 1.89.
The measurement capability of 600,000 gallons is based on recent plant. designs.
For older plants with smaller water capacitics, licensees may propose deviations from this requirement based on the availabic water supply capability at their plant.
. Narrow-range ' water Icyc1 monitors are requirad for all sizes of sumps but are not required in those plants that do not contain sumps inside the containment.
.(.
.For BWR pressure-surpression containments, the emergency core cooling system (ECCS) suction line inlets may be used as a
starting reference point for the-narrow-range and wide-range water level monitors, instead of the bottom of the suppression pool.
The accuracy requirements of the water 1cvel monitors shall be
.provided and justified to be adequate for their intended function.
LILCO Position Containment wide-range water level indication channels meet the design and qualification criteria as outlined in Appendix B to NUREG-0737 to the maximum extent practical.
For Shoreham the lower limit is at the elevation of the center line of the ECCS suction lines containment penetrations.
In order to provide
" suppression pool water level ecasurement with an upper limit of 5 ft above the normal water 1cvel (26'-6")
the currently installed instrument taps have been relocated to 31'-6".
The accuracy of the water level monitors are 0.2 percent of span.
II.F.1-10
-II.F.3, Attachment 5 - (Cont'd) - Additional Information The ECC3 suction lines (including RCIC) currently penetrate the containment at El 24-0.
The lines then turn downward via an elbow connection and are attached to a suction strainer. The elevation of the strainer connections are approximately as follows:
~~-
HPCI-:EL120-10 RCIC-~EL 22-3.5
- Separate level monitoring instrument penetrations are provided in the suppres-sion pool at El 24-0, and represent the approximate nominal elevation for adequate system. performance. Operation below the suction line penetration would be operating under suction lift conditions; le.,'if pump operation were
- stopped, a subsequent. restart could not be guaranteed since a vacuum condition could exist with possible air in leakage. Also, operation near or at the minimum' pipe suctiori elevation would be subject to vortex concer,ns and pump performance could again not be guaranteed.
l As discussed above, the level monitoring instrument penetrations are approp-riately located at El 24-0.
In addition, it should be noted that there are no remaining ' spare penetrations below that elevation.
It is not feasible to provide additional penetrations at this time. The Shoreham containment construction is complete and drilling through the liner and concrete would have a significant schedule impact. Also, it is not desirable to provide connections below the minimum water level from the standpoint of containment integrity.
Shoreham has'also considered alternative instrumentation fcr level measure-ment of the' suppression pool. We have in the past evaluated instruments
.which could be mounted inside the primary containment such as those presently utilized in' Pressurized Water Reactors. However, due to the structural loads impressed by Mark II Containment suppression pool swell, it was judged impractical since the level instrumentation was of the float type and the
~ manufacturers had no test data to evaluate the effects of such an event.
Non-mechanical ultrasonic type level instrumentation was evaluated and
+
found unsuitable for use inside the primary containment due to a history L
of poor reliability and a lack of any manufacturer who could provide such l
anLinstrument for Category I service. We believe that the use of proven, reliable instrumentation mounted outside the primary containment seismically t
and environmentally qualified, and accessible for periodic calibration and i
test Lis the most acceptable arrangement.
i-In summary, continuous, repetitive and reliable system operation would not be expected below the existing penetration elevation and should not be a system design requirement.
l II.F.1-10a 7/22/81
-b ATTACllMENT 6 CONTAINMENT IIYDR0 GEN MONITOR NRC Position 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.
Design and qualification criteria are outlined in Appendix B of NUREG-0737.
a..
The continuous indication of hydrogen concentration is not required during normal operation.
If an indication is not available at all times, continuous indication and recording shall be functioning within 30 minutes of the initiation of safety injection.
The accuracy and placement of the hydrogen monitors shall be provided and justified to be adequate for their intended function.
LILCO Position The hydrogen concentration in the primary containment atmosphere 1is continuously monitored by the hydrogen analysis system.
This system consists of two redundant subsyste1s, each including two hydrogen analysers to sample the drywell and the suppression chamber atmosphere (see Figure 6.2.5-1).
Each analyzer is provided with dedicated instrument penetrations to ensure continuous monitoring.
The range of the analyzer is from 0 to 10_ percent hydrogen concentration by volume over a pressure range of
-2 to' +60 psig.
The accuracy of the hydrogen conitors are 2 percent of -full scale.
Containment hydrogen concentration measurement channels meet the design and qualification criteria as outlined in Appendix B to NUREG-0737 to the maximum extent practical.
II.F.1-11 N
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.. i SNPS-1 FSAR TABLE II.F.1-1 0
- HIGH-RANGE NOP>LE GAS EFFLUENT MONITORS
-REQUIREMENT Capability to detect
.and measure concentrations of noble gas fission products in' plant.
gaseous effluents during and following an accident.
All potential accident release paths shall be monitored.
To provide the plant operator and emergency PURPOSE planning agencies with information on plant releases of noble gases during and following an accident.
DESIGN BASIS MAXIMUM RANGE Design range values may-be expressed in Xe-133 equivalent values for monitors employing. gamma radiation detectors or in microcuries per cubic centimeter of air at standard temocrature and pressure (STP) for monitors employing beta radiation detector (Note :
1R/hr el ft-= 6.7 Ci Xe-133 equivalent f or point source).
Calibrations with a higher energy source are acceptable.
The decay of radionuclide noble gases after an accident (i.e.,
the distribution-of noble. gases changes) should be taken into account..
O Undiluted containment exhaust gases (e.g., PWR 105.pci/cc reactor building
- purge, PWR drywell purge through the s tandby gas treatment system).
Undiluted PWR condenser. air removal system exhaust.
~ 10 4 pCi/cc Diluted containment exhaust gases (e.g.,
10:1
- dilution, as with auxiliary building exhaust air).
BWR r e a c t o'r-building (secondary containment) exhaust air.
PWR secondary containment exhaust air.
-103'pci/cc Buildings with systems containing primary coolant or primary coolant offgases (e.g., PWR auxiliary buildings, EWR turbine buildings).
PWR steam safety valve discharge, atmospheric steam dump valve discharge.
102 pCi/cc Other release points (e.g.,
radwaste
.()
, buildings, fuel handling / storage buildings).
1 of 2
~
SNPS-1 FSAR TABLE II.F.1-1 (CONT *D)
Not required; monitoring the final release REDUNDANCY point of several discharge inputs is acceptable.
(None) Sampling design criteria per ANSI N13.1.
SPECIFI-CATIONS Vital instrument bus or dependable-backup
- POWER SUPPLY power supply to normal ac.
Calibrate monitors using gamma detectors to CALIBRATION Xe-133 equivalent (1 R/hr D 1 f t = 6.7 Ci Xe-133 equivalent for point source).
Calibrate
-monitors using beta detectors to Sr-90 or similar long-lived beta isotope of at least.
0.2 MeV.
Continuous and recording as equivalent Xe-133 DISPLAY concentrations or gCi/cc of actual noble gases.
QUALIFICATION -
The instruments shall provide sufficiently accurate responses to perform the intended function in the environment to which they will be crposed during accidents.
Offline monitoring is acceptable for all ranges
- DESIGN CONSIDERATIONS of noble gas concentrations.
Inline (induct) sensors are acceptable for 102 pCi/cc to 105 pCi/cc noble gases.
For less than 102 pCi/cc, offline monitoring is recommended.
Upstream filtration (prefiltering to remove radioactive iodines and particulates) is not required; howe.ver, design should consider all alternatives wi"Ja respect to capability to monitor effluents following an accident.
For external mounted monitors (e.g., PWR main stemn line), the thickness of the pipe should be taken into account in accounting for low-energy gamma radiation.
O e
e 2 of 2
,1 SNPS-1 FSAR TABLE II.F.1-2 SAMPLING AND ANALYSIS OR MEASUREMENT OF HIGH-RANGE RADIOIODINE AND
~~
-PARTICULATE EFFLUENTS IN G ASEOUS EFFLUENT STIdIJ4S Capability to collect and analyze or-measure EQUIPMENT representative samples of radioactive iodines and particulates in plant gaseous effluents during and following an accidenr.
The capability to sample and analyze for radioiodine and particulate effluents is not required for PWR secondary main steam safety valve and dump valve discharge lines.
To determine quantitative release of PURPOSE radioiodines and particulates for dose calculation and assessment.
102 pCi/cc of gaseous radiciodine and particu-DESIGN BASIS SHIELDING lates deposited on sampling media; 30 minutes ENVELOPE sampling time, average gamma energy (E) of 0.5 MeV.
SAMPLING MEDIA Iodine 90%
effective adsorption for all forms of gaseous iodine.
Particulates
>' 90%
effective retention for 0.3 micron (v) diameter particles.
SAMPLING CONSIDERATIONS Representative sampling per ANSI N13.1-1969.
Entrained moisture in effluent stream should not degrade adsorber.
. Continuous collection required whenever exhaust flow occurs.
Provisions for limiting occupational dose to personnel incorporated in sampling
- systems, in sample handling and transport, and in analysis of samples.
ANALYSIS Design of analytical facilities and preparation of analytical procedures shall consider the design basis sample.
Highly radioactive samples may not be compatible with generally accepted analytical procedures ;
in such
- cases, measurement or emissive gamma radiations and the use of shielding and distance factors should be considered in design.
1 of 1
,a
.r SNPS-1 FSAR d
TABLE II F.1-3 CONTAINMENT HIGH-RANGE RADIATION MONITOR The capability to detect and measure the REQUIREMENT radiation level within the reactor containment during and follouing an accident 1
rad /hr to 108 rads /hr (beta and gamma) or RANGE.
alternatively 1 R/hr to 107 R/hr (gamma only) 60 kev to 3 MeV photons, with linear energy RESPONSE' response i 207.) for photons of 0.1 MeV to 3
MeV.
Instruments must be accurate enough to provide usable information A minimum of two physically separated monitors REDUNDANT (i.e.,
. monitoring widely separated spaces within containnent).
Category I instruments as described in Appendix
- DESIGN AND
-QUALIFICATION B to NUREG-0737, except as listed below In situ calibration by electronic signal SPECIAL CALIBRATION substitution is acceptable ior all range decades above 10 R/hr. In situ calibration for at least one decade below 10 R/hr shall be by means of calibrated radiation source.
The original laboratory calibration is not an acceptable position due to the possible I
differences after in situ installation.
For high-range calibration, no adequate sources exist, so an alternate was provided.
Calibrate and type-test representative speci-SPECIAL ENVIRONMENTAL mens of detectors at sufficient points to dem-QUALIFICATIONS onstrate linearity through all scales up to 106 R/br.
Prior 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.
0 e.
e 4
e 1 of 1
S11PS-1 FSAR TABLE II F 1-4 RADIOACTIVITY CONCE!?PRATION RA"GES FOR SilOREll AM GASEOUS EFFLUENT RADIATION MONITORS-RNNGE
( oCi/cc) '
_r
____. GASEOUS EFFIUENT MONITOR M.
Reactor Building Standby
~
~ Ventilation RE-021, RE-022*
1x10-6._to 1x10+2-Post Accident Reactor Building 3x10-3 to.2x10+5
~ Standby Ventilation RE-134*
Reactor Building Normal
.. to 1x10-1 1xTO-6 Ventilation RE-029*
Turbine Building Ventilation-1x10-6.to 1x10-1 RE-0574 Radwaste Building 'Jentilation 1x10-6 to 1x10-2 RE-055*
1x10-6 to 1x10 -l' i
Station Vent Exhaust RE-042*
Post Accident Statir. Vent 3x10-3 to 2x10+r-Exhaust RE-126*
Reactor Building ITormal*
1x10-2 to 1x10+3 Ventilation PE-068 1x10-2 to 1x10+3 Turbine Building Ventilation RE-067
- Radwaste Building Ventilation 1x10-2 to 1x10+3 RE-0 6 6* ~
1x10-2 to 1x10+3 Station Vent Exhaust RE-069
- 1x100 to 1x107 R/hr Drywell Monitors RE-085A, B i
- Ranges shown for thene radiation monitors aro for the noble gas portion of the monitor.
9 J,
ATMOSPHERE RESVS RELEASE POINT I
RE-063 RE-022 l RE-021
' FILTER S 7
ll Y__
CHARCOA!.
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PAM RE-042 1
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0-EXHAUST FILTER L RSSVS/B
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- STATION VENT EXH AUST FANS '
RE-r oT
[EiAl RE-067
!l l PAM l OUTSIDE DILUTION AIR TURB1NE
! (NORM ALLY CLOSED)
BUILCING l
RE-029 R E-055 RE-OG8 RM Prt RE-066 lPAM l l PAM l ti REACTOR RADWASTE g
U BUILDING BUILDING (RBNVS)
RM: R ADI ATION MONITOR. THESE MONITORS DETECT NOBLE G ASES, AND CONTINUOUSLY COLLECT SAMPLES FOR RADIOlODINE AND PARTICUL ATE RELEASE AN ALYSIS.
PAM: POST ACCIDENT MONITOR. THESE ARE HIGH RANGE IN-LINE R ADIATION MONITORS.
PAM-RMs POST ACCIDENT HIGH RANGE EFFLUENT MONITOR.
plg,7[,p,l_g THESE MONITORS DETECT HIGH RANGE NOSLE GASES AND GASEOUS EFFLUENT RADI ATION MONITORS CONTINUOUSLY COLLECT SAMPLES FOR RAD!OiODINE AND PARTICULATE RELEASE ANALYSIS.
SHOREHAM NUCLEAR POWER STATION-UNIT 1 THERE ARE ADDITIONAL GASEOUS STREAM RADI ATION MONITORS FIN AL SAFETY ANALYSIS REPORT IN THE SHOREH AM PL ANT. THIS StMPLIFIED DIAGRAM SHOWS ONLY THOSE DISCUSSED IN TEXT.
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l DIVISION M FIGURE II.F.1-2 GEfiERAL ARRANGEMENT NOTE:
FOR THE REACTOR BUILDING
.SEE FIGURE 7A FOR CRD PL AN EL. 7 8'- 7" HYDRAULIC UNIT DETAILS SHOREHAM NUCL E AR POWER STATION - UNIT I FIRE HAZARD ANALYSIS REPORT I
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III.D.3.3 Innlant Radiation Monitoring
}gC Position Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine s
concentration in areas within the f acility where plant personnel
,may be present during an accident.
Ef fective monitoring of increasing iodine levels in the buildings. -
under accident conditions must include the use of portable.
. instruments using sample media that.will collect iodine
- selectively over xenon (e.g., silver zeolite)..for the following reasons:
a._
The physical size of the auxiliary and/or fuel handling :_:
building precludes locating stationary monitoring-
. instrumentation at all areas where airborne iodine concentration data might be required.
=.
b.
Unanticipated isolated
" hot spots" may occur in locations where no stationary monitoring instrumentation is located.
c.
Unexpectedly high background radiation levels near
... stationary monitoring instrumentation after an.accidcnt may interfere with filter radiation readings.
d.-
.The time required to retrieve samples af ter an accident.
- . may result in high personnel exposures if these filters are located in high-dose-rate areas.
After January 1, 1981, each applicant and licensee shall have the -
capability to remove the sampling cartridge to a low-background, low-contamination area for further analysis. -Normally, coun tng rooms in auxiliary buildings. will not have sufficiently low.
backgrounds for such analyses following an accident.
It the low background. area, the sample should first to purge 6 of any entrapped noble gases using nitrogen gas or cican air. free of noble gases.
The licensee shall have the capability to. measure.
accurately the iodine concentrations present on these samples under accident conditi'ons.
There should be sufficient samplers to sample all vital areas.
LILCO Position The inplant iodine concentration will be determined by using either portable, semi-portable, or fixed air samplers to draw a
known quantity of air through either a charcoal filter or silver zeolite cartridge.
Fixed samplers are located in ventilation streams and have charcoal filter collection capability for radiciodine.
III.D.3.3-1
9-SNPS-1 FSAR There shall be at least three (3) semi-portabic continuous air monitors capable of de'ecting particulate, iodine, and noble gas concentrations, (Eberline Instrument Corporation PING-3 0:-
equivalent).
Also included shall be at least four (4) portable air samplers with the capability of using either a charcoal filter or silver zeolite cartridge for collection of radiciodine (Eberline Instrument Corporation RAS-1/ICll-1 or equivalent).
If the presence of interfering noble gas activity is confirmed, silver zeolite shall be used where feasible for iodine sampling.
Prior to
- analysis, the charcoal filters will be purged with bot' 1.ed nitrogen or clean air to remove entrapped noble gases; this is not necessary for the silver zeolite sample.
The sample will be counted according to normal operating health physics procedures using instrumentation capabic of accurately measuring iodine concentrations.
Instrumentations used for this analysis will be located in both the radiochemistry counting room and the alternate on-site counting room.
At least one of these locations will remain a
low-contamination, low-background area for all postulated accident conditions.
III.D.3.3-2
1 Item # 37 - Secondary Containment Bypass Leakage I.
An evaluation was conducted of all fluid systems to examine the lines or penetrations that pass through the primary containment and extend, without being vented to the secondary containment, outside
'he secondary containment boundary. A list of those lines and the results of our review are as follows.
A.
'Feedsater System Primary containment isolation is provided by a check valve inside and a testable check valve outside of primary containment.
In addition to the containment isolation valves, there is a motor operated stop check valve located outside the primary containment.
The two check valves located outside of primary containment are equipped with positive closure measures which ensure valve closure and seating following an accident. After isolation of the HPCI and RCIC systems, there are three valves in series between the inside of primary containment and the environment.
B.
HPCI and RCIC Suctions The HPCI and RCIC systems draw suction initially from the condensate storage tank and then from the suppression pool. The condensate storage tank, during HPCI and RCIC operation, is isolated from suppression pool water by means of a check valve and motor-operated
. gate valve in each system's suction line. After operation of the HPCI and RCIC systems is terminated, a second, closed, motor-operated gate valve provides additional isolation.
C.
-Core Spray Suctions A locked-closed globe valve in each suction line of the core spray system isolates suppression pool water from the condensate storage tank.
D.
HPCI and RCIC Test Return Lines Two motor operated, normally closed gate valves on the test return lines isolate the condensate storage tank from either the HPCI or RCIC systems. After HPCI or RCIC operation is terminated, further isolation is provided by closure of another motor operated gate ' val ve.
4 01 37-1 i
i
-. -. -. ~ _ -
3 E.
Condensate Fill Connections to the HPCI, RCIC, Core Spray and RHR Systems The condensate transfer system is used as the alternate fill source to the RHR, HPCI, Core Spray and RCIC systems. The condensate connections are isolated by means of normally closed globe valves
'as well as check valves in the lines.
. F.
RWCU Connection to the Condenser During an accident the RWCU System is isolated from the RPV and y
RWCU is not expected to.contain highly radioactive water.
The i
condenser is further isolated from the RPV by two closed gate valves in the. blowdown line.
G.
Suction and Reci'rculation Lines for the RCIC Loop Level Pump from the Condensate Storage Tank Three (3) 1" spring-loaded check valves are used to isolate the i -
RCIC Loop level fill system from the RCIC system.
Further isolation is provided when operation of RCIC is terminated.
0 H.
Service Water System & Ultimate Cooling Connection to the RHR System The ultimate cooling water connection to. the RHR system is protected i
. against leakage in either direction by dual isolation valves with a drain-off connection between the two valves.
I.
CRD System i
The design configuration of this system is discussed in FSAR Section r
6.2.4.3.2 (Containment Isolation System)
J.
RBCLCW System
'This system including the provisions for leakage detection is discussed in FSAR Section 9.2.2.
1 K.
Instrument & Service Air System This system is discussed in FSAR Section 9.3.1 and TMI Item Response II.K.3.28.
The service air portion of the system is not hard piped to the containment penetration. As shown on Figure 9.3.1-1D, a base
-connection with a quick disconnect type fitting is removed during plant operation. This arrangement effectively vents any leakage past the isolation valves to the secondary containment.
01 37-2
(
II. None of the lines or penetrations identified in Section I need be considered as potential bypass leak paths of containment atmosphere.
Lines A through H, by system design, functions, and location,will not.._
expose the internals of the primary containment isolation valves to containment atmosphere. Water exists on both sides of the containment.,
isolation valves thereby preventing contact with containment atmospherF-and acting to seal the valves against atmosphere leakage. The isolation valves will.not be exposed to containment atmosphere for at least 30 days.
Within the primary containment Systems I, J and K are closed and they are not exposed to containment atmosphere. Systems I and J are used ~
d9"i.ng nermal operation, and potential system degradation would.be.
,11dontified during normal maintenance. Systems I and J are wa;er filled which further act as a seal against containment atmospheric leakage.
In addition, System u is vented to the secondary containment via the RBCLCW head tank.
The Instrument Air Subsystem of System K is at all times at a system pressure greater than the containment peak pressure ensuring that any air leakage would be into containment.
III. N/A IV. N/A V. Closed Systems I & K meet all of the requirements of Position 9, BTP CSB 6-3.
01 37-3
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J SER Section 6.3.1: - Additional Information i
HPCI/RCIC - 50% PLUGGED STRAINER TEST N
LILCO Response LILC0 will perform tests to verify the flow capabil.ity of the HPCI pump and the RCIC pump, with their respective strainers 50% plugged, during
- the Preoperation Startup Test Program.
These tests will be performed using auxiliary boiler steam. The RCIC test will demonstrate full flow capability whilE the HPCI test will verify a flow less than full flow due to auxiliary boiler steam cap.acity limitations. Based on this partial flow data, calculations and extrapolations using the HPCI pump curvc viill be performed to demonstrate full HPCI pump capability..The test procedures, data, analysis, and l
results will be on file as part of the Pre-op Program.
a k.
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