ML20058N543

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Proposed Tech Specs Re Defueled SAR
ML20058N543
Person / Time
Site: Shoreham File:Long Island Lighting Company icon.png
Issue date: 08/10/1990
From:
LONG ISLAND LIGHTING CO.
To:
Shared Package
ML20058N542 List:
References
NUDOCS 9008140189
Download: ML20058N543 (30)
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s ATTACHMENT TO SNRC-1740' Insertion Instructions .

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i Remove Old- Insert New  ;

Page 7-2 thru 7 Page 7-2 thru 7-4 Pages 9-3'thru 9-18 Pages 9-3.thru 9-17 ,

. Table 9.'2.1-1 (pages 1~& 2) Table 9.2.1-1 (pages'1.& 2)

Page 10-2 Page 10-2 ,

'6 Page 11-1 Pape 11-1 .

'I Pages 11-12'thru'11-14 Pages 11-12 thru 11 Pages.11-20, 11-21 Pages 11-20,.11-21 Page.11-23 Page 11-23 Page 11-25 Page 11-25 3

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.'- SHOREHAM DSAR 4-

[' -7.1.1.1.6 Reactor Manual' Control System This system is~not needed to. support the storage of the fuel.in the fuel pool, therefore it is not included in the DSAR.

I 7.1.1.1.7 Reactor Vessel Instrumentation Reactor vessel instrumentation monitors and transmits-information concerning:-key. reactor vessel operating variables. Portions of this system will-only be used if a wet-layup~of the reactor vessel is utilized.-

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.7.1.1.1.8 Reactor Recirculation System This system is not needed to support the storage of-the fuel in

'the. fuel pool, therefore it11s not included in the DSAR.

7.1.1.1'.9- Feedwater Control System

This system is not needed to support the storage 1of the-fuel in

'the fuel pool, therefore it is not' included in the DSAR.. 1 I'

7.1.1.1.10 Pressure' Regulator and Turbine-Generator Controls This system is not needed to support the storage of the fuel in j the' fuel pool, therefore it is not included in the DSAR. I l

7.1.1.1.11 Remote Shutdown System i i

.This system is not needed to support'the storage of the fuel in-the fuel = pool, therefore'it-is not included in the DSAR. l 7.1.1.1.12 Screenwell Pumphouse Ventilation System ]1 The screenwell pumphouse ventilation system instrumentation and i controls remain functional and'are-designed'to-ventilate each of '!

the.two rooms of the1 building using separate, 100 percent outside ,

air-ventilation systems.- _

7.1.1.1.13-- Process Computer System I i l

This system is not needed to support the storage of-the fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.14 Reactor Core Isolation Cooling System This system is not needed to support the storage of the fuel in

.the' fuel pool, therefore it is not included in the DSAR.

8 7-2 Rev. 1 Aug. 1990 n

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- ~SHOREHAM DSAR'

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7.1.1.1.15 Standby Liquid Control' System

, This system is not'needed to support the storage of the fue1~in

'the fuel pool, therefore it is not included in the DSAR.

4 7.1.1.1.16- Reactor Water Cleanup System

'The reactor. water cleanup (RWCU) system instrumentation and controls provide manual initiation of-system' equipment to maintain 1high water purity in the reactor water. This system. i will be used only if the reactor syst'em is placed in a wet layup l condition.;  !

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'7.1.1.1.17 Leakage Detection System ,

This system is not needed to support the storage of the fuel in.

the fuel pool, therefore it is not included in the DSAR.-

7.1.1.1.18 Reactor Shutdown Cooling Mode-RHR System This system is'not needed to support the storage of the' fuel in the fuel pool, therefore it is not included in the DSAR.

7.1.1.1.19 Radwaste System Radwaste system instrumentation and controls support manual .

processing and~ disposing of the radioactive process wastes.

7.1.1.1.20 Emergency Diesel Generators This system is not needed to support the storage of the fuel in the. fuel' pool. >

7.1.1.1.21 Turbine Building Closed Loop Cooling Water System The turbine building closed loop cooling water (TBCLCW) system instrumentation and controls remain functional to maintain the turbine ~ building cooling water system at design temperature and monitor. system performance. The.TBCLCW system also cools the equipment in the radwaste building and supports the station pressurized air compressors.

7.1.1.1.22 Service Water System Under' normal conditions the service water system provides cooling ,

for the plant components.

7.1.1.1.23 Recirculation Pump Trip System This system is not needed to support the storage of the fuel in l the fuel pool. 3 7-3

I. g SHOREHAM DSAR

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7.1.1.1.24 Reactor Building Standby Ventilation System. .[

.The filtration' portion.of the system is not needed to support-the E

storage of the fuel in the fuel pool. Certain fans and air operated. valves.will remain functional to support RBNVS T operation. See DSAR section 9.4 for additional'information.

7.1.1~.1.25 Reactor Building Closed Loop Cooling Water-System This system is not needed to support the storage of the fuel in the fuel _ pool.

'7.1.1.1.26 Primary Containment Atmospheric Control System +

This system is not needed to support the storage of the fuel in the fuel pool, t

7.1.1.1.27 Fuel Pool Cooling and Cleanup Systems 1 Fuel pool cooling and cleanup systems-instrumentation and controls remain unchanged except that the coo' ling portion is protected because' evaporative' cooling is sufficient to remove the -

small amount of decay heat.:

Control Room Air Conditioning System

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7.1.~1.1.28 The' control room air conditioning (CRAC). system instrumentation and controls for one of the two redundantzsubsystems'are l functional to maintain-the' main control room at design temperature during normal and emergency; conditions, monitor system performance,~and' permit manual as well'as automatic

~ initiation of an air supply fan. l l 7.1.1.1.29 Chiller. Equipment Room Ventilation System

.This system is not needed to support storage of the fuel in the fuel pool.

7.1.1.1.30 Diesel Generator. Room Emergency Ventilation Systems i This system is not needed to support the storage of the fuel in l

the fuel pool and is protected.

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7.1.1.1.31 Relay Room, Emergency Switchgear Rooms, And Computer

' Room Air Conditioning System q S

. The relay room,. emergency switchgear rooms, and computer room air conditioning system instrumentation and controls for one of the two redundant subsystems are maintained functional to automatically control the ventilation system to maintain these rooms at their design temperature and system performance.

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iu SHOREHAM DSAR 19.1.4.3 Description of Fuel Transfer The fuelthandling system provides a safe and' effective means for transporting and handling fueltfrom the time it reaches'the plant until<it. leaves the plant after post-irradiation cooling. The preceding: subsection describes the equipment and methods used in fuel handling. The following paragraphs describe the integrated fuel transfer system, which ensures that the design bases of the

-fuel-handling system and the requirements of Regulatory Guide l'13 are satisfied._

9.1.4.3.1 Arrival of Fuel On Site 1k) new fuel is expected to arrive on site. Therefore this section of the USAR is not required. -l 9.1.4.3.2 ~ Refueling Procedure No refueling is planned. Therefore this section of the USAR is s not required.

9.1.4.3.3 Departure of Fuel from Site t

This section applies as written in the USAR.

.In addition:

1. The spent fuel will be removed from the site in certified fuel shipping casks.
2. The casks will be leak tested prior to shipment.

The remainder of USAR Secticn 9.1.4 is applicable.

9.2 WATER SYSTEMS 9.2.1 ' Service Water System The Service. Water (SW)-System is as. described in USAR Ttctions

.9.2.1.1 thru.9.2.1.5 with the following changes because of the- l reduced heat removal requirements with the plant in the de-fueled state.  !

'a)- The RBSW system is considered non-safety related because it does'not provide-cooling water to any plant equipment required to-perform a safety function.

b)= One RBSW pump will supply cooling water to one RBSVS/CRAC chiller condenser and to all Turbine Building Service Water (TBSW)' cooling loads. (See item e below.) No service water

.is required for RHR, diesel generator cooling, RBCLCW, drywell cooling, and makeup water to the reactor vessel 9-3 Rev. 1 Aug. 1990

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w V SHOREHAM DSAR ultimateLeooling; connection (UCC). The testable: check valve t

-in the~UCC will_not require testing'to. verify forward flow.-

-Emergency service; water to the spent fuel pool is not required (per DSAR Chapter 15) because of the very low heat generation by the fuel, c): Automatic' start / initiation due to accident signals are_not required.

9 d) The double isolation valves which split the RBSW from-the TBSW subsystems will-belopened to intertie the subsystems. l

, , e)' Normal operation will now consist of only'one RBSW pump in l

-use because of the minimal heat load imposed by the TBCLCW-system to. support the station air compressors. It will

- supply cooling water to one TBCLCW heat exchanger, the circulating water pump bearing and the-fish retention pool.

Cooling water for the-vacuum priming pump seal. cooler is not E; required. The second RBSW pump will remain in standby. _l l

f) The_only tests and inspections to be performed on'the TBSW system in the defueled condition are those that are deemed-to; q

-be' required.for proper operation and maintenance. 1

9) Table 9.'2.1-1 has been revised. q

.9.2'.1.5 ~ Instrumentation Application >

This section remains unchanged except that only the .

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'instrumentat' ion needed for the Service Water System'as described -  ;

in 9. 2.1 ' a) through g) is required.  ;

9.2.2 Reactor Building Closed Loop = Cooling Water (RBCLCW)-System This system is not needed to support the storage of fuel in the  ;

spent fuel pool. ]

9 ; 2. 3- Makeup Water Demineralizer System The description contained under this h'eading'in the latest revision of the Shoreham USAR remains unchanged except as ,

follows: ,

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1. SBLC, RBCLCW, seal water injection,'and vacuum priming are no.

longer users:of. demineralized water in the defueled' D condition.  ;

2. The HPCI suction line=from the condensate storage tank is not required to be maintained as safety related in the defueled t condition.

9-4 Rev. 1 Aug. 1990

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For further information-relative-to'this system refsr.to the

'USAR.

9. 2. 4 - ' Potable'and Sanitary Water Systems ,

The description contained under thic heading in the latest  ;

revision of the Shoreham USAR' remains unchanged in the defueled condition. For further information on this subject refer to the

.USAR.

9.2.5 Ultimate Heat Sink >

With the fuel in-the Spent Fuel Pool, the ultimate heat sink

-(Long. Island Sound)-no longer has any safety significance, since the decay-heat of the. fuel is insignificant. However, the ultimate heat sink will' continue to be used as a source of cooling _ water for normal plant needs (refer'to DSAR Section

  • 9.2.1). ,

9.2.6 Condensate Storage Facilities While'in the defueled condition:the condensate storage facilities provide makeup water for the fuel storage pool. The description

.of this system in the USAR remains unchanged.except as follows:

1. Condensate, feedwateri reactor systems, HPCI and RCIC will no longer be users.  ;

'2. HPCI test discharge and CRD pump return lines to the CST are '

not required to be active.

'3 . The:first'three. paragraphs of USAR 9.2.6.3 are no longer applicable.

4. The last paragraph of USAR 9.2.6.4 and 9.2.6.5 are no longer. 3 5

. applicable.

'9.'2.7 sTurbine Building Closed Loop Cooling Water System-The-description contained under this heading in the latest .

revision of the Shoreham'USAR remains unchanged in the defueled condition. The only exception is that many of the coolers listed ,

in DSAR Table 9.2.7-1.will normally be valved out of service  !

while'the plant remains in the defueled state.

For further information on this subject refer to the USAR.

9.2.8 Main Chilled Water System This system will not be maintained as an operable system since it i is not needed with the plant in the defueled condition.

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.. SM0REHAM DSAR-9.2.9- Reactor Building: Standby' Ventilation System And Control Room Air Conditioning Chilled Water System Redundancy in this system _is no't needed'since the RBSVS; system is not required to operateJin the defueled condition. The heat loads generated by the electrical equipment in.the control room, i relay room and the emergency switchgear room are greatly reduced, such that only one. chiller is required to_ maintain the' control room, relay room and_switchgear room at design conditions. The 3 operating chiller and associated pumps will be manually-

. controlled from the' control room. Aside from the above, the

. system design remains unchanged and further information can be found under the above heading in the Shoreham USAR.

9.3- PROCESS AUXILIARIES 9.3.1 Compressed Air Systems The description contained under this heading in the latest '

revision of the USAR remains unchanged in the defueled condition-except.for the following:

1. Piping that has'been installed as ASME III code class 2 is no longer considered safety related and is reclassified QA

, Category' IIA.

2. Nitrogen will no longer be used for inerting the primary _

containment or for equipment within the' primary containment. ,

3.- Safety 1related functions of_the compressed air system no ,

longer exist. -No pneumatically operated valves al i required i

for safe shutdown.

-For further information on_the' compressed air-system, refer to the USAR.

L' 9.3.2 Process Sampling System The Process Sampling: System provides monitoring of.certain -

process operations while fuel is in the spent fuel pool for -i either-short or long term' storage. The process monitoring is

, accomplished as necessary by means of' measuring, analyzing-and/or recording for conductivity, pH, and silica ccncentration, as shown on DSAR Table 9.3.2-1. l 9.3.3 Equipment and Floor Drainage System With the Reactor defueled and the fuel assemblies stored in the Fuel Pool, large portions of the Equipment and Floor Drainage System are not required.

9-6 Rev. 1 Aug. 1990

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System Description

This' system is described in the USAR. Changes in status are

addressed below.

' Reactor Building ,,

The only source of radioactive waste to the Equipment and Floor Drainage System in the Reactor Building.is the Fuel Pool and-associated service equipment leakage. Sources.in the USAR that are no longer applicable are the Drywell Equipment Drain System and the' Reactor Recirculation Pumps Drainage System. The Drywell Equipment Drain Tank is no longer required _One or more floor drain sumps are no longerTrequired,.as applicable.

Turbine Building, 3 The Turbine Building Floor Drain and Equipment Drain Systems are-no longer required, as applicable, except for the Decontamination Sump drains and associated equipment. There is no1 steam and the turbine is no longer required','southat'the only source of radioactive waste is the Chemical" Laboratory.

Radwaste Building .

The=Radwaste' Building Equipment and Floor Drainage System is:

maintained operational. The Dirty Waste Sump and Pumps (IN52-TK ,

114 and'1N52-P-187A/B) and Regenerant Recovery Sump and Pumps 1 (1N52-TK-115 and 1N52-P-181A/B) are no longer required.

-t 9.3.4 Chemical, Volume Control, and Liquid Poison Systems The Standby Liquid Control System is no. longer required in the i defueled condition. The RWCU System is also no longer required.

unless the-Reactor is' layed up wet.

9.3.5 Failed Fuel Detection System With the fuel.in the pool, the description in the USAR Section-is

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no longer applicable.

In the event of gross fuel rod failure in the fuel pool (see

" Worst Case Fuel Damage Accident" in DSAR Chapter 15), the refueling floor process radiation monitors will detect this radioactivity if it.becomes airborne.

- .9.3.6 Suppression Pool Pumpback System This system not required to support storage of fuel in the fuel pool.

9.4 AIR CONDITIONING, HEATING, COOLING, AND VENTILATION SYSTEMS 9-7 Rev. 1 Aug. 1990

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SHOREHAM-DSAR a:

1 9. 4 .1 : Control-Room Air ~ Conditioning System C '

.The Control Room AC system remains unchanged in design and Loperating functions. However, the system is reclassified.to QA Category IIA, the filter portion of the system will no longer be L required and one of each of the redundant fans and ACUS will no  ;

< longer'be required. The AC system will only function to provide-an OSHA' environment for the operators.during the fuel storage period. This requires the operation of only one RBSVS/CRAC chiller. Automatic initiation systems and interlocks for the habitability-portion of the system will be protected and the AC i system will be manually controlled from the control room. For further discussion on this system refer to the Shoreham USAR.

9.4.2 Reactor Building Normal Ventilation System 9.4.2.1 Design Basis The RBNVS remains unchanged in design and operating function except that the system will only: [

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1. Provide ventilation by introducing. filtered outside air in the reactor building at a rate of approximately 2.7 air changes per hour
2. Remove heat generated by solar and external heat '

transmission, lighting and the fuel pool.

3. Support monitoring system for radioactive release'through the- l )

exhaust air system.

4. Induces negative pressure in reactor building for secondary
  • containment integrity'.

For further discussion on this system refer to the'USAR.

- 9.4.3 Radwaste Building Ventilation l

-The description contained under this heading in the latest Shoreham USAR. remains unchanged, except that:the charcoal exhaust  ;

filtration system is no longer required and one of the two

. redundant supply.and exhaust fans, mechanical refrigeration units and circulating pumpsiare also no: longer. required. . Refer to the USAR for information on this subject.

9.4.4 Turbine Building Ventilation System And Station Exhaust System A) -Turbine Building Ventilation System This system is not required to support the storage of fuel in the spent fuel pool.

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B) ._ Station Exhaust System This: system will expel. the exhaust air from the radwaste building l ,

and the reactor building. IJowever, only : one f an will be needed-s Lfor this purpose, allowing one fan to be-protected and still- l

. maintain a fan on standby. This will ensure that the Isokinetic-nozzles located'in the upper level of the exhaust duct will see a-sufficiently high velocity to be operational. For further

~ discussion regarding this system refer to the Shoreham USAR.

9.4.5.

Battery Room Heating And Ventilation Tho' description contained under this heading in the latest- -

-revision of Shoreham USAR remains unchanged. Refer to USAR for 1 information on this subject. This system is reclassified to 0.A.

Category IIA. f 1

9.4.6, Drywell Air Cooling: System ini's--system is not needed while the fuel is stored.inLthe spent fuel pool. ,

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9.4.7 Screenwell Pump House Heating And Ventilation The description contained under this. heading in the latest- {

revision of Shoreham USAR remains; unchanged. . Refer to USAR for '

This sytem is reclassified to 0.A. l informa' tion on this subject.

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19.4.8 Plant Heating {

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.The' description 1 contained under-this heading in the latest  !

revision of Shoreham USAR remains unchanged.- Refer'to USAR for information on this-subject. ]

9.4.9 Primary Containment Purge System x

'This system-is not needed while the fuel is stored in the spent j fuel pool.

9.4.10 Diesel ~ Generator Room Ventilation f

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1This system is net needed while the fuel is stored in the spent i i

fue11 pool. This system is reclassified to 0.A. Category IIA. l l

9.4.11 Relay Room, Emergency Switchgear Room And I y Computer Room Air Conditioning System ,

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The description contained under this heading in the latest revision of-Shoreham USAR remains unchanged with the exception i that only one train of equipment will remain functional. Refer  !

to USAR for information on this subject. This system is reclassified to QA Category IIA.

9-9 Rev. 1 Aug. 1990

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9.5 OTHER AUXILIARY SYSTEMS 9.5.1 Fire Protection System-9.5.1.1L Design Basis

.The. design basis section applies with the following addition j

The' basic premise of the fire protection discussions in the USAR and FHAR is protection from fire for. safety related areas

including areas containing equipment or circuits that are (1) required for safe shutdown, or (2) required.to prevent or  !

mitigate radiological releases comparable to 10CFR 100 limits.

'Since safe shutdown is assured by non-operation of-the plant, and all of the nuclear fuel is in the fuel storage pool, the only remaining safety related' area is the Reactor Building.

Structures, systems components and administrative controls in ,

place to. protect. areas, equipment or circuits previously' identified as safety related will be maintained as required for property loss prevention purposes and should be considered the

.same as those fire protection features described in the USAR'for  ;

protection of non-safety related areas.

Three documents which were used in the design of the plant's fire

< protection' features and continue to be part of the fire '

protection program are:

l. ' Evaluation of the SNPS Fire Protection-Program as compared to '

10CFR50, Appendix R criteria submitted via SNRC 572 dated May-21, 1981. ,.

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2. Fire Hazards Anal'ysis Report.
3. Cable Separation Analysis Report:  !

SNRC 532 dated February"10, 1981 SNRC 811 dated April 13,;1983 However,.the term " safety related", as used in those documents l

and in USAR section 9.5.1," applies'only to the' Reactor ~ Building. -l t

Section 6 of the Fire Hazards Analysis Report (FHAR) contains technical requirements that formerly were fire protection technical specifications.

/HAR Chapter 6 reflects reductions in the technical requirements that--are consistent with the text of this DSAR Section 9.5.1.

Types of Fires The " types of fires" section upplies with no changes. i Design Criteria 9-10 Rev. 1 Aug. 1990 1

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LThe." design" criteria"Lsectionvapplies with the following addition:

'As. discussed above,-this design will be maintained for property' loss prevention purpones. HowcVer, the " safety related" application ofxthe listed' documents, particularly.NRC's Branch LTechnical; Position APCSB 9.5-1 and Appendix A thereto, is limited j to the Reactor Building.

Locations-of Fires' 4  !

The " locations of fires":section applies with the following changes: .

The rooms listed. parenthetically as examples of safety.related-areas having a concentration-of-cables are' reclassified to Q.A.

-CategoryLIIA._ The rooms listed as examples of where oil fires '

could~ occur nearl safety rel'ated' equipment ~no: longer fit that description becauseithese areas, arc' reclassified to QA Category

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l IIA' .Furthermore, the' fire ~ hazard' associated with this equipment is significantly reduced while the equipment is not being used

because1the ignition sourcescassociated.withethe operating  ;

equipment ~have been. eliminated. j intensity of fires  ;

This.section applies without change.

0 Fire Characteristics ~

I This.section applies without change.

" Building Arrangement and Structural Features  ;

The " building arrangement-and structural features" section applies ~with the following. changes:

n ~ Inithe; response to NRC question'3, as shown in FHAR revision 3, "

LSNPS has stated our intention to replace existing motorized. fire

' dampers 1with newly designed fire dampers. All of-the areas where Lthese'new dampers were-to be installed are in the control  !

Building and are reclassified to Q.A. Category IIA. Therefore, this' proposed modification will not be implemented. The CO systems for those rooms are in electric lockout. When a fi$e is detected, the CO 2 system controls would cause the dampers to close on an electrical signal- . As a backup, the fusible link of o

=each of the existing fire-dampers is sufficient to cause closure

of a damper in-the' event of a fire, thus assuring integrity of theLfire barriers.

9-11 Rev. 1 Aug. 1990 l.

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In, contrast with this'USAR sect i on, an unprotected HVAC. opening.

- exists in the east wall of.each of-the-three diesel generator-rooms within 50 feet of an_ oil-filled 1(Reserve Station Service) s i transformer. This deviation was reported to the NRC on Licensee

' Event Report 87-021. The-proposed corrective action'was.to- g install.a deluge 1 water curtain system below the existing missile

. shield ~ wall betweenithe transformer and the wall openings. Since the diesel generator rooms are reclassified to QL Category IIA, this modification will~not be implemented. The partial 1 protection provided:by the missile barrier-is considered- l sufficient for non-safety related areas. {

i Seismic Design This section applies _without change.

-I Water Requirements

's The " water requirements" section applies with the following additional statement:

Although some areas previously identified as safety related are reclassified to QA Category IIA, the water supply is not being reduced.

Codes and Standards q This section applies without changes. SNPS will continue to meet i the requirements of the applicable NFPA codes for fire protection systems that-remain functional. .,

9.5.1.2 System Description i

'The " System Description" section applies with the following

' changes: ,

'As' discussed earlier, all fire protection features remain in place.: Several rooms / areas listed in this section as. safety related~are reclassified'to Q.A. Category IIA. Essential- '

circuitry installed for' safe shutdown of the plant is no longer needed for that purpose.'.'Nofremoval of such cable or change in i its physical separation-is contemplated. Similarly, the service water.line inside the Reactor Building, where a spare connection 5 exists for manual hookup to the-fire protection water system, is reclassified to Q A. Category IIA. Modifications that would degrade its seismic design are not contemplated at this time.

9.5.1.3 Safety Evaluation Electrical Insulation Fires This section applies without change.

9-12 Rev. 1 Aug. 1990 l

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[t- Charcoal Fires n 'This section upplies without change.

P" 011 Fires h The " oil fires' section of the safety evaluation applies with the

!: folicwing: change:

b As discussed earlier, the fire hazards associated with non-operating-equipment are significantly reduced because the primary ignition sources - electrical energy and hot surfaces'-

are eliminated.

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! Severity, Inten. 'ty and Duration of Fires This section arm es without changes, Time Estimate 9 This section applies without changes.

Failure Mode and Effects Analysis This section applies without changes.

Accidental Initiation of Fire Protection System The " accidental initiation of fire protection system" section applies with the following change:

LAreas protected by CO systems are among those that are no longer L consideredsafetyrelkted.

Single Failure in Fix. < cote,ction Systems This'section applies without change.

Pipe Breaks in Fire Protection Systems This section applios without changes.

Failure of Fire-Protection System Affecting Safety Related-Equipment This section applies with the following change:

Of the areas listed, only the Reactor Building is still  ;

considered safety related.

l, 9-13 Rev. 1 Aug. 1990 ..

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.- SHOREHAM DSAR 4,

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'9.5.1.4- Tests and Inspectionst b,

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9.S.1.5- Personnel Qualification and Training This section applies.without changes.

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'9.5.2 Communications System t

9.5.2.1- Desion' Bases This!section of the USAR remains unchanged.

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.9.5.2.2' System Description n

i This-section of the USAR remains unchanged except for the-following:

1. For the very low. frequency (VLF) portable-radio systems, one low-powered VLF radio base station will be used in conjunction with two mobile car units to provide offsite

-radio communications (instead of two VLF base stations and four' mobile car units) .

'2. :The Emergency Operations Facility (EOF)'.is not required,

, since no emergency requiring EOF activation can occur with the' fuel;in the Spent Fuel Pool.

9.5.2.3 ' Tests and Inspections This:section"of the USAR remain unchanged.

t: '9.5.3 Lighting Systems-LWhilefin the defueledJcondition this system will provide.all the

.necessary~ required lighting *to the plak.t and the< site.- The description of-this system in the USAR remains unchanged except for'the following:

l'. ,Section 9.5.3.2, item #2 - the standby AC-lighting system will receive power.from plant service bumes which are powered.

42 from offsite.

2. -Same'section,' item f5 - the fifth lighting subsystem will' receive power from DC battery sources while the plant remains .j

.in the defueled condition.

3. LThe last paragraph of the same section, the independent power sources for lighting, remains unchanged but the source of l' power will be from plant service buses and DC battery sources uit needed.

9-14 Rev. 1 Aug. 1990 g

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- SHOREHAM DEAR ,

b -!

'9.5.4 Diesel Generator Fuel Oil Storage and Transfer System Since emr,rgency power is no longer required with the fuel in the Spent Psel- Pool, the Emergency Diesel Generators are not ,I required, and sections 9.5.4 - 9.5.7 of the USAR no longer apply.

9.5.5 Diesel Generator cooling Water System 9.5.6 Diesel' Gen.erator Starting System i 9'.5.7 Diesel Generator Lubrication System f 9.5.8- Primary containment Leakage Monitoring System With the fuel'in the Spent Fuel Pool, the Primary Containment Leakage Monitoring Systep is not required. l b 9.5.9 Storage of Gases Under Pressure \

The quantities and; type'of ga'ses stored'in pressurized containers "in the defueled condition is reduced from that previously on hand._ The design bases remain unchanged. Storage facilities are l provided for the following gases as shown in. Table 9.5.9-1:

1. Carbon Dioxide for fire protection. 1 '
2. Halon 1301 for fire protection.
3. Air for instrument, control, breathing and service.
4. Nltrogen for glycol and HW heating..
5. Propane for auxiliary boiler ignition'.

The following gases are no longer used or required to be stored ,

in the defueled condi* ion l'. Hydrogen for mac generator.

2. Hydrogen and oxyven for gas analyzers.
3. Nitrogen for containment inerting.  !
4. . Nitrogen for drywe.'.1 floor seals.
5. Nitrogen =for electrohydraulic control. .
6. -Air for MSIV accumulators (inboard and outboard). .

7.. Air:for long term accumulators. ,

8. Air for standby diesel generators.  ;

The-statement in the USAR relative to maintenance and laboratory

ases remain unchanged. The safety evaluation discussed in section 9.5.9.3 of the USAR is only applicable for air for  !

instruments, service breathing, and control and for carbon dioxide and halon. Statements relative to the pressure relief '

valves and gas release hazards remain as discussed in the USAR.

Gas use for safe shutdown is no longer necessary in the defueled -

condition.

P 9-15 Rev. 1 Aug. 1990

p q i .- .

I . SHOREHAM DSAR

{ ' Appendix 9A FUEL CRITICALITY ANALYSIS i

The Shoreham Spent Fuel Rack (SFR) is of a stainless _ steel and water neutron flux trap design which uses no additional poison.

j A description of the storage racks is provided in 9.1.2. The criticality analysis of this rack dtsign is described in detail in Appendix 9A of the Shoreham USAR. The reactivity results which are summarized lln USAR Table 9A-4 remain valid for the conditions existing at Shoreham after defueling. Furthermore, due to the differences in U-235 enrichment-between the SFR designed and the current Shoreham fuel, a large negative reactivity credit should be taken into account. Xi;is is explained as follows:

The Shoreham SFR design is based on a maximum U-235 enrichment of 3.1 wt. %. - The resulting ,asic cell k is calculated to be 0.9129 without uncertaAnty and model adjustments (Table 9A-4, Appendix 9A, Shoreham USAR). The Shoreham Cycle 1 fuel loading uses three (3) enrichments. Of the 560 fuel assemblies in the core, 340 bundles.have the highest bundle average U-235 enrichment of 2.19 wt.  %, 144 bundles of 1.76 wt. % and 76 remaining bundles uses natural uranium.

If the six inch natural uranium segments at the top and-bottom of the fuel are excluded, the average enrichment of a 2.19 wt. % bundle becomes 2.33 wt. %. Using this enrichment and linearly. extrapolating the reactivity vs. U-235 6nrichment results given in Figure 9A-5 of Appendix 9A, Shoreham USAR, the reactivity difference between the SFR design enrichment of 3.1 wt. % and the current maxiumum loading enrichment of 2.33 wt. % is found to be about -6.0%

in k ( k -0.060). This brings the basic cell k under nominal storage conditions for the. current fuel down to 0.85, which is well below the regulatory acceptance criterion o# _k 0.95. All the corrective and uncertainty adjustments listed in Table 9A-4 of the Shoreham USAR remain applicable.

During the period from July, 1985 to June, 1987, Shoreham went  !

-through three' separate-stages of low power testing (less than 5%

of rated power), which resulted in a total core exposure of approximately 48 mwd /MT as determined by a series of core-follow analyses. . The net effect of=the core exposure is a slight decrease -in reactivity- ( -0.002 in - k) mainly due to the offsetting contributions from the formation of Sm-149 and the slight depletion of the burnable cd poison in the fuel bundlec. '

In light of the large reactivity margin described previously (k O.85), no additional credit will be claimed here. i

}

9-16 Rev. 1 Aug. 1990 G:  !

l [,,

,, SHOREllAM DSAR i 9B EVALUATION OF SPENT FUEL POOL MAKEUP REQUIREMENTS An analysis was performed to determine the rate of water loss ,

from the spent-fuel pool through_ evaporation under the worst case ,

scenario described below. The following conservative. assumptions d are used in the analysis to maximize the calculated pool  ;

evaporation rates. ,

1) The spent fuel pool temperature is 110'F. l
2) The ambient temperature above the spent fuel' pool is. l conservatively assumed to have zero relative: humidity.  !

-t 3): The reactor building air flow exists due to normal l ventilation system operation to maximize evaporation.  ;

The result of the calculation shows that the maximum evaporation.

rate from the pool'is approximately 0.6 gpm which translates to a )

. pool level depletion rate of one foot per eleven days. Based on this very low maximum depletion rate, external cooling of the.

-spent fuel pool is not required. Technicalsspecifications e require that the water level above the' spent fuel be a least  ;

twenty-one feet. In addition, it should be noted that_ pool water i level is: alarmed-in the1 control room andfalarm-response i procedures exist to' provide' appropriate operator action. l t

i t

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e k

i-i e-1, sev. 1 Aue. 1,,e  ;

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g SHOREHAM DSAR: - _

W mg . - . Q $-ii -

s, _ _- , .1

. TABLE 9.2.1-1; -

SERVICE WATER SYSTEM COMPONENT DESIGN DATA Nominal Capacity Number of Components Each. Utilized in. -

- f-Componeg Quantity (qpm) Defueled Condition R accor Building Service Water Pumps' 4 8,600 1

. Turbi.ne Building Service Water Pumps 3 '8,000 - - -

React >r Building Subsystem Components:-

Reactor Building Service Water Strainers 4 250 1 Diesel Generator Jacket Coolers 3 700 -

Residual Heat Removal Heat Exchangers, 2. 8,000 ---

Reactor Building; Closed Loop Cooling--- '2 6,370 ---

Water Heat Exchangers Reactor Building Standby Ventilation 4 525 1 System Chiller Condensers Main Chilled Water System Chiller. 3 l',500 ---

l 400 i_ Condensers ,1 Drywell Cooling Bocrter Heat Exchangers 2 1,460 --

d l

L

~

1 of 2 - Rev. 1 Aug. 1990-

c. .,

p .,, ,__,m,,, ,__ _ , , _, _ _

-"7'" aw++r-v 9--,w_y ~yf9 ,,- yy. , ., y,, .

n -- .. n ,, . - ..c--- ,- - _; .

^

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+ .;j SHOREHAM DSAR 1 - y -:

~

TABLE 9.2.1-1 j - -

~

SERVICE WATER' SYSTEM COMPONENT-DESIGN DATA (Cont'd.) _

4 Capacity Number of' Components

.Each Utilizediin.

Component ' Quantity (qpm) ~ Defueled Condition- -

_: S Turbine Building Subsystem Components:

Turbine Building: Service Water: Strainers '2 420' -- _

l

~

^

Circ Water Pump Bearing Cooling .4- -6, Note 1 l~

Fish Retention Pool l' .185 I s

Turbine Building Closed Loop Cooling 2 '14,200 1 Water Heat Exchangers .

Vacuum Priming Pumps Seal Water. Coolers 3 100 ---

'e Note 1: One circ water pump bearing cooler will be needed if a; circ wateripump is used to provide dilution of chlorine.

-T

.c 1

i 4

2 of 2 rec 1; Aug. 1990"

___..-A--

o O 4 j; .

, SHORERAM OSAR l10.4.4 Turbine Bypass System The purpose for which.the turbine bypass system was built no longer exists. The components of this system.as_ described in the USAR are not required in the defueled condition.

The-portion of the bypass system upstream of the bypass valves was' built to'ASME III cc2 criteria. Asithe function of the bypass syr, tem no longer exists ~in the defueled condition, the l bypass system is reclassified 0.A. Category IIA.

10.'4.5 Circulating Water' System The purpose for which the circulating water system was built no >

longer exists.. The components of this system as described in the USAR will.not be required in the defueled condition. The only j exception is that a circulating water pump and the circulating water discharge system may be used to provide dilution capacity for elimination of liquid radwaste and SPDES limits on chlorine and, suspended solids to the Long Island Sound.  ;

10.4.6 Condensate Demineralizer System

.Since there is no fuel in the Reactor and no Reactor steam produced,-there is no need for the Condensate Demineralizer

, System.. This system will be protected. However, the Acid and

- Caustic Storage Tanks :(lN52-TK-035 and -TK-036) will remain operable to provide regeneration chemicals for the continued operation of the Domineralizer and Makeup Water System (P21).

The Chemical Waste Sump (lN52-TK-ll3) will remein operable as a pathway for further treatment of non-radioactive regenerant. waste from the Demineralizer and Makeup Water System.

-10.4.7 Condensate and Feedwater System  ;

The purpose for which the condensate and feedwater. system was built no longer exists. The components of this system as described in the USAR will not be required in the defueled ,

condition.

Piping built and des.igned to ASME III cci is considered Q.A. 1 Category IIA while in the defueled condition. Inservice '

inspection according to ASME XI need not be performed while in the defueled condition.

F 10-2 Rev. 1 Aug. 1990

C >

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'SHOREHAM DSAR l 3  !

i CHAPTER 11 ]

RADIOACTIVE WASTE MANAGEMENT 11.1 RADIATION SOURCE TERMS The description contained under this heading in the latest revision of Shoreham USAR remains unchanged as it'is used to 1 detalop the basic design criteria of the plant. Ilowever, the actual source terms in the plant's present defueled condition are as follows:  ;

a..

~

Liquid Radioactivity Sources As of August 1989, since all SNPS' fuel had been placed in the. ,

spent fuel pool, there were no liquid sources with nuclide ,

concentrations greater than the Lower Limit of Detection (LLD),

outside of radwaste streams. It must be recognized that in the q future some concentrations greater than LLD will be seen (e.g.,

as sludge at the bottom of sumps is processed to Radwaste). l _;

!!owever, these should be mir.or and temporary occurrences.

Sources related to the decontamination and decommissioning should also be minor, as the degree of overall plant contamination is low.- These liquid sources would be dealt with in accordance with the Liquid Radwaste, ALARA, and Health Physics programs as I discussed in DSAR Sections 11.2, 12.1, and 12.5, respectively.

Isotopic concentrations above the LLD levels in the Radwaste System as of 6/30/89 are indicated in Table 11.1-1, from

References:

2, 3 and 4.

c

b. Gaseous Sources There are no detectable' gaseous sources at SNPS,.either present  ;

or' anticipated. 'This statement is supported by the fact that the Semi-Annual Radiological Effluent Release Report for the first and second quarter 1989,(Reference 11) indicates there;were no detectable' releases during the six-month period,'either from the ,"

offgas system or_the various building exhaust systems.

c. Activated Materials Sources It is expected'that materials which were located in the reactor vessel during low power testing (eg, control rods, TIPS, IRMs,.

and the like) have been activated to some extent. With the exception of some portions of the liquid radwaste system (10 mrem /hr maximum), dose rates outside of plant. systems are very These low dose rates are indicative low,.less than 0.5 mrem /hr.  !

of a low deposition of sources within plant systems.

i 11-1 Rev. 1 Aug. 1990 1

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. .SHOREHAM DSAR

meteorological, lar.d, and water use data. The range of analyses l performed on a sample depend on the type of sample taken.

+

Sampling' locations are designated as either indicator or control.

Indicator locations provide representative measurements of radiation and radioactive materials for those exposure pathways and radionuclides'(from SNPS) that lead to the highest potential radiation exposures. Control locations are placed sufficiently far from SNPS so that they will be beyond the measurable influence of SNPS or any other nuclear facility. This monitoring program implementsSection IV.B.2 of Appendix I to 10CFR Part 50, by verifying that measured concentrations of radioactive materials and direct radiation are representative of.the actual contamination levels and doses to the public.

SNPS' REMP has been subdivided over three distinct time intervals: Preoperational REMP (prior to SNPS' initially achieving criticality), Operational REMP (from initial criticality until removal of the fuel from the core), and Post-Defuel REMP (after the core was transferred to the spent fuel pool).

Preoperational REMP was performed to identify and determine background levels of-environmental activity around SNPS, Preoperational REMP also served to verify that indeed.the media

.being sampled and analyzed is sensitive to radiological fluctuations in SNPS' environs (indicator locations) and to provide effective monitoring of potential critical pathways.

Preoperational and Operational REMP samples within the aquatic environment included surface water, algae, fish, invertebrates (clams, lobsters, etc.) and sediment. The atmospheric environment war sampled for airborne particulates, iodine, and noble' gases. Milk, potable water, precipitation, game and food products were obtained from thu terrestrial environment. -Direct radiation was measured using thermoluminescent dosimeters (TLDs).

The range of analyses for each sample were gamma spectrometry,  !

Sr-89 and Sr-90 I-131; H-3, gross beta, direct radiation and .

noble gases. Under Post-Defuel REMP, several of the above l sampling-locations and/or range of analyses are discontinued. i The current Post-Defuel REMP program is outlined in Tables.

11.6.3-1 & 11.6.3-2.

Preoperational REMP began in February 1977 and continued through 1 l:

1984, although the official Preoperational REMP period; i.e. the time frame against which the data base from Operational REMP was l compared f occurred during 1983 and 1984. The Operational REMP  ;

L began on February 15, 1985 when initial criticality was achieved. I L Except for reactor operator training programs which required the I I

reactor to operate at '0.0% power' (during January 1989), SNPS has not generated radioisotopes since the last 5.0% power test, completed on June 6, 1987. Comparisons between the 11-12 Rev. 1 Aug. 1990 _1 l l

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1 SHOREHAM DSAR-p ..

above-two phases of REMP were documented in each Semiannual '

Radiological Effluent Relcase Report.

As of August 9, 1989, SNPS' core was transferred to the spent fuel pool -- as part of the agreement between LILCO, state and local governments not to operate Shoreham. This transfer '

prevents criticality from being reestablished. In addition, since SNPS' last 5.0% power test was completed'during June 1987, per Ref. 9, virtually all iodines and gaseous effluents have decayed away. Consequently, the surveillance requirements for  !

,f SNPS' Post-Defuel REMP were reduced to below the-operational  !

level.

Justification for Reducing REMP to Post-Defuel Surveillance Levels-t Pursuant to Reg Guide 4.1, once the initial core of the licensca has reached the point (in time) of maximum burnup, and the licensee has demonstrated (using results from environmental media or calculations) that the doses and concentrations associated witt. a particular pathway are sufficiently small (comparable to preoperational levels), then the number of media sampled in the .

pathway and the frequency of sampling may be reduced to normal Tech Spec requirements. Since (as of August 9, 1989) the core has been in the spent fuel pool, the initial core has " exceeded" the point of maximum burnup. ]l

~

It should be noted that the concept of " normal" Tech Spec r requirements as referred to in Reg.. Guide 4.1, refers to a fully operational station with normal surveillance requirements. Reg.

Guide 4.1 does not account for the unique condition at SNPS.

Consequently, the justification for the reduced surveillance program will be performed in two steps. Step one reduces Operational REMP to the level mandated when SNPS was to become operational. Step two reduces the sur .;11ance program further, .

to the= revised requirements corresponding to the defueled condition. ,

. Dose calculations to SNPS' environs (1983 - 1988) were performed by analyzing positive concentrations of radioactivity detected in collected samples. Table 11.6.1-4 compares the radiological

-impact from-each major pathway to the public during SNPS' preoperational and operational REMPs. Specifically, the radiological impact during SNPS' 5.0% power testing program (1985

- 1987) was compared to preoperational REMP.

In all cases, the calculated doses during both the operational and preoperational phases of REMP were comparable. Therefore, no .

environmental radioactivity was identified (during any of the ]

5.0% power tests) as having originated at SNPS. These results satisfy the criteria established in Reg. Guide 4.1 for reducing L post-defuel REMP to the level originally mandated by SNPS' license. The sampling points not required by the license are:

11-13 Rev. 1 Aug. 1990 l

l

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. SHOREllAM DSAR

1) Game; 4) Rain Water; and
2) Aquatic Plants 5) Noble Gases.
3) Aquatic Sedimentt Justification for reducing REMP to the revised requirements (after the core was defueled) is given based on the above informat. in; i.e., the measured environmental impact due to 5.0%

power testing was comparable to that of preoperational REMP, and as of August 9, 1989, the core was removed from the reactor pressure vessel. SNPS' last 5.0% power test was completed on June 6, 1987, and per Ref. 9, with the exception of I-1.'.9 and Kr-85 (4.0 mC1 and 1560.0 Ci, respectively), all iodines and gaseous effluents have since decayed away. In addition, c=dwaste systern activities are quite low (listed in DSAR Sections 11.1 &

12.2). As a result, the only remaining radioisotopes (and their release pathwaye) are:

I sotope ( s) Source Effluent Pathway

1) Kr-85 Spent Fuel Gaseous
2) Solubles and Radwaste Gaseous and Liquid Particulates SNPS' Post-Defuel REMP Surveillance Program Outline
1) DIRECT RADIATION: Reduce from 36 to 18 locations Quarterly Surveillance Frequency
2) AQUATIC-
a. Aquatic Plants and - Delete, not required Beach Sediments
b. Fish, Surface Water - Retain, may be impacted and Invertebrates from liquid release path to L.I. Sound Perform Semiannual surveillances as available
3) AIRBORNE
a. Iodine - Delete, insignificant quantity
b. Particulates and. - Retain, particulates and Gross Beta and solubles still exist.
c. Noble Gas - Delete Noble Gas, not required.

Quarterly Surveillance Frequency 11-14 Rev. 1 Aug. 1990

i; o J7 o

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o-' SHOREHAM DSAR ,

, [l c 111.6.3.1.2 A,tmospheric Environment j l The atmospheric environment is examined by analyzing airborne particulates collected on Gelman Type A/E filters using low volume air samplers (approximately 1 cfm). The samplers used l  !

are equipped with vacuum recorders for sample volume correction i I

and to indicate sample validity and maintenance problems when they occur. Should the sampler lose vacuum due to a leak the i v60uum level reading will drop to zero. Since this may occur i

without a corresponding loss of electric supply the exact time of t the maintenance problem will be evident en the recorder chart. l Sample volumes are measured using dry gas meters and corrected for differences between the actual pressure that the volume meter i sees and the average atmospheric pressure. Sample volumes _are corrected to_ standard pressure using average weekly barometric  !

pressure (measured at. Environmental Engineering Department.  !

Melville) and air sampler vacuum readings. Time totalizers indicate the duration of time the sample is taken.

t 11.6.3.1.S Terrestrial Environment I

j The terrestrial environment is examined by analyzing samples of milk and fo xi products. When available, milk samples are F

-collected quarterly, except during the pasture season (May ,

through October) when the sampling is increased to monthly. Milk

. samples are prepared for shipment in accordance_with the

instruction of the laboratory performing the analysis. Food products consisting of vegetables and fruit are collected from j area farm ~ stands and shipped fresh to the laboratory.  ;

11.6.3.1.4 Direct Radiation j Direct radiation levels in the environs are measured with energy compensatedLealcium sulfate (CaSO4:Dy) TLDs, each containing four separate readout areas. The TLDs are annealed by LILCO prior to placement in the field._ One TLD is placed at each of the 18 locations,_and exchanged on a quarterly bases; these locations correspond to the 16 meteorological sectors in the general areas ,

of the site boundary, plus.two control locations (actual locations are listed in Table 11.6.3-1). The units are then '

packaged and shipped to the laboratory for analysis.

11.6.3.2 Sampling Locations and Freguency ,

Typical REMP sampling locations and frequency are given in Table 11.6.3-1. These locations are described in Table 11.6.5-2 and <

shown in Figures 11.6.3-1 and 11.6.3-2, 11.6.4 NOT USED IN THE DSAR (Data Incorporated Into Section 11.6.1)

L 11-20 Rev. 1 Aug. 1990 0 _ __ - _ _ _ _

+. .o. ..

[\

e

]

T NL .I SHOREHAM DSAR

~

l

["' 11.6.5 Data Analysis, Presentation and Interpretation

."he discussion contained in the latest version of the Shoreham UCAR (Section 11.6.5, 11.6.5.1, and 11.6.5.2) continues to apply.  ;

11.6.6 Program Statistical Sensitivity  ;

F The discussion contained in the latest version of the Shoreham t 1USAR (Section 11.6.6) continues to apply.

REFERENCES In Section 11.6

1) Regulatory Guide 4.1 " Programs for Monitoring Radioactivity i p in the Environs of Nuclear Power Plants"  !

~

t

2) Not Used j

(-

I

3) Not Used

, 4) Radiological Branch Technical Position, Rev. 1, Nov. 1979 j l

5) Reg. Guide 4.15, Rev. 1, February 1979, " Quality Assurance j For Radiological Monitoring Program (Normal Operation)  :

Effluent Streams and the Environment" '

5

6) .SNPS Technical Specifications ,

3/4.12 Radiological Environmental Monitoring 3/4.12.1 Monitoring Program Table 3.12.1-1 " REMP"

7) Not Used  ;
8) SNPS' Operational REMP Annual Reports: January 1, to December  ;

31, 1983, 1984, 1985, 1986, 1987, & 1988 issued ~by Nuclear Engineering and Environmental Engineering Departments of LILCO. j i .9) :C-RPD-476, Rev. O, 10/21/88, "SNPS Core Thermal Power After (

Shutdown" -

t i

Y i

11-21 Rev. 1 Aug. 1990

- :g

' 5:

SEOREHAM DSAR ,

B BLE 11.6.1.-4' -

Cmmrison Of Operational - Frevperational REMP Data .

i Operational REMP ) (- Prevperat.ional REN -)

(

1986 1985 1984 1983 SAMPIE TYPE Unit / Isotope 1988 1987 240 - 410 140 - 450 130 - 420 150 - 290 120 - 540 70 - 220 Potable Water pCi/l (H-3) 35.1 - 6490 54 - 3230 992 - 4330 641 - 5340 34.0 - 6310 came pCi/Kg(Cs-137) 76.7 - 9270 1.9 - 5.7 3.0 - 6.2 2.7 - 6.9 2.3 - 5.7 Direct (gamn) mrem Monthly 2.3 - 5.2 2.8 - 6.9 2.9 - 4.9 2.8 - 5.5 3.1 - 6.2 2.8 - 5.4 Quarterly 2.7 - 4.8 2.9 - 5.0 Radiation 5.0 - 44.0 4.0 - 32.0 5.0 - 360 6 - 47 4.2 - 61. 5 - 54 Air: Gross Beta [x1.0E-3]

LT 0.8 LT 0.8 0.11 - 0.27 LT 0.8 LT 0.07 1.3 - 1.4 Particulate Sr-90 pCi/m' x 1.E-3 LT 10.0 35 - 1230** LT 10.0 LT 10.0 LT 30.0 Iodine-131 pCi/m 2 x 1.E-3 LT 10.0 LT 1.0 LT 1.0 6.8 - 27.

  • 33. LT 20.0 Aquatic pCi/Kg (Sr-90) LT 1.0
  • 85.5
  • 47.9
  • 45. 69.7 - 140. 36 - 55 Plants pCi/Kg (Cs-137) LT 6.0 pCi/1 (Sr-90) 0.76 - 6.00 0.61 - 5.70 0.98 - 13.0 0.86 - 4.60 0.69 - 5.3 0.9 - 7.7 5.90 - 11.5
  • 4.4 9.6 12.9 - 14.1 Milk pCi/l (Cr-137) 6.00 - 14.8 7.0 - 8.9 2.1 - 4.8 LT 0.20 LT 0.20 ra pCi/1 (I-131) LT 0.20 LT 0.20 LT 4.0 .LT 4.0 LT 4.0 LT 3.0 1%

Food pCi/Kg (I-131) LT 4.0 LT 5.0 LT 5.0

  • 12.2 LT 5.0 LT 5.0

I l

  • Ranges are not given since only one data point contaired an identified isotope.

1 l

Bev. 1 Aug. 1990 11-23 L __ _

a .mo -

1 O

s

SIDPD M DSAR ' TABLE 11.6.3-1 i Post-operational Radiological Enviromental Mmitoring Program (RINP) ,

Media Sanpling Locations Sanpling Prequency Analysis Direct (1) 131,2A2,3S1,4S1,5S2, Quarterly ' Gama Exposure  ;

Iwiiation 6S2,7A2,8A3,9S1,10A1, ,

'11A1,12A1,13S3,14S2, ,

t 15S1,16S2,*5E2,*6El Fish and (2) 3C1, 14C1, *13G2 Smi-annually Gama-isotopic Invertebrates or when in scuon l Fruits, (3) 8B1, 6B21, *12H1 At tine of Annual Gama-isotopic and Vegetables  !!arvest Airborne (4) 6S2,2A2,3S1,7B1,*11G1 Quarterly Gross-Deta Particulates Ganma-isotopic I

Milk (5) 13B1,*10F1, or *8G2 Quarterly. During Gama-isotopic Pasture Season, l l Monthly Surface Water 3C1 or 14C1, and *13G2 Smiannual Gama-isotopic Grab Sanple 11 - 3 l _ (*) Designates Control locations (1) Eighteen nonitoring stations, DR1 through DR18, (16 indicatcr and 2 control). l are used. One indicator location is positioned in each noteorological sector near the site boundary. One dosimeter or continuously measuring dose rate instrunent is placed at each location.

(2) At each Indicator location, one sample of each comercially and recreationally inportant species. One sanple of same species in control location.

(3) Sanple three different kinds of broad leafy vegetables grown nearest to two indicator locations.- having highest predicted average ground level D/O (when milk sanples not available) . Also take one sample of same leafy vegetation grown nearest to Control location.

(4) 'Ihree sanples (near SNPS), one frm each of the three Meteorological *ectors

-having the largest annually averaged ground-level D/0, are taken. One sanple (near a ommlnity) also having the highest calculated temually averaged ground-level D/O is taken. Establish one Control location.

(5) Indicator sanples frm milking animals having highest potential dose.

i Sanple within 5 km distance (preferably), within 5 to 8 km where doses are l

calculated to exceed 1 mrm/yr (second choice) or frm 8 to 17 km. Control location is 15 to 30 km frm SNPS and in the least prevalent wind direction.

11-25 Rev. 1 Aug. 1990

.

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