Text

Rev 2 to 10855-D7.5, Environ Design Criteria for Hope Creek Generating Station
	ML20125B183
Person / Time
Site:	Hope Creek
Issue date:	10/29/1984
From:	; BECHTEL GROUP, INC.
To:	;
Shared Package
ML20125B156	List: ML20125B154; ML20125B183;
References
	10855-D7.5, NUDOCS 8506110340
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10855 REV.

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DESIGN CRITERIA DOCUMENTS DISCIPLINE D7.5 REVISION STATUS SHEET 2

PAGE 2

of 96 ED-46 (8-74) srP assea

o 10855-D7.5, Rev. 2 TABLE OF CONTENTS Page I.

SCOPE 6

II.

DESCRIPTION 7

III.

REQUIREMENTS 7

IV.

USE OF TABLES 1 through 7 10 d.

ADDITIONAL ENVIRONMENTAL CONDITIONS 11 V

VI.

REACTOR BUILDING ENVIRONMENT 15 VII.

AUXILIARY BUILDING ENVIRONMENT 22 VIII.

TURBINE BUILDING ENVIRONMENT 25 IX.

INTAKE STRUCTURE ENVIRONMENT 26 X.

GENERAL BUILDING ENVIRONMENT 26 XI.

GENERAL CRITERIA, STANDARDS AND GUIDES 28 XII.

REFERENCES 29 l

4 OO86H 3

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1085'5-D7.5, Rev. 2 LIST OF TABLES i

Pages 1.

ANTICIPATED ABNORMAL EVENTS 31 1

2.

POSTULATED DESIGN BASIS EVENTS 33

,3.

DESIGN BASIS EVENT ENVIRONMENTAL CONDITIONS 34 4.

RADIATION CONDITIONS INSIDE PRIMARY 39 CONTAINMENT 5.

ENVIRONMENTAL CONDITIONS INSIDE PRIMARY 40 CONTAINMENT - DBE 6.

NORMAL AND MAXIMUM PLANT ENVIRONMENTAL 41 CONDITIONS 7.

REACTOR BUILDING INTERNAL FLOOD DEPTHS 82 LIST OF FIGURES 1.

BWR DRYWELL PRESSURE ENVELOPE 86 2.

8WR DRYWELL TEMPERATURE ENVELOPE 87 3.

SAMPLE ENVIRONMENTAL DATA SHEET - GENERAL 88 4.

SAMPLE ENVIRONMENTAL DATA SHEET - DUCTS 89 AND VALVES 5.

PRIMARY CONTAINMENT ZONES 92 6.

PRIMARY CONTAINMENT TEMPERATURE RESPONSE - LOP 93 OO86H 4

. ~... _. _ _ _

_.. _, _ - - _ _ _ _ _ _ _ _ -. ~. _ _.. _ -. _ _ _ _. _ _ _. _.. _. _ _ _ _ _. - _, _ _.

10855-D7.5, Rev. 2 7.

PRIMARY CONTAINMENT PRESSURE RESPONSE - LOP 94 8.

WETWELL PRESSURE ENVELOPE - DBE 95 9.

WETWELL TEMPERATURE ENVELOPE - 08E 96 0086H

10855-D7.5, Rev. 2 I.

SCOPE A.

This document specifies the bounding plant indoor environmental data to be used for the design and procurement of Bechtel supplied equipment. This document provides the basis for the overall E.Q.

program. During the procurement cycle, the equipment location is determined so that the environmental condition may be extracted from the appropriate tables. Certain equipment however may not be qualified to the stated environmental conditions. See paragraphs IIC, IIIC and IVA for further details.

8.

Seismic and missile design requirements are not included in the scope of this document.

C.

The plant environmental conditions contained in this document have been taken from the following sources.

1.

Normal Conditions - Bechtel design criteria or calculations for the heating and ventilating and radiation shielding design for Hope Creek Unit 1.

2.

Abnormal Conditions - Bechtel design criteria or calculations performed to bound the environment caused by such an event. Abnormal condition calculations were performed for areas containing or affecting safety-related equipment.

3.

Design basis event (DBE) conditions - Bechtel design criteria or calculations of bounding environment for the DBEs are identified in Table

2. The bounding environment for other than radiation conditions is always due to a pipe

break, e.g., PBIC, PROC, instrument line break.

The LOCA is always the bounding condition for post-accident radiation conditions, with the following exceptions:

a.

HPCI and RCIC system doses are determined by the anticipated transient without a scram (ATWS) accident.

b.

Spent fuel pool cleaning and cooling system doses are determined by the fuel handling 4

accident.

4 i

f 0086H 6

j

10855-D7.5, Rev. 2 4.

General Electric document 22A2928, Revision 2, SWR Equipment Environmental Interface Document 5.

General Electric document NEDO-10698, Environ mental Qualification of Class I control and Instrumentation Equipment 6.

Reactor building flood levels - Bechtel calculations defining flood depths D.

This document does not describe the requirements for methods of qualifying equipment to environmental conditions. These requirements are contained in other design criteria and specifications.

i II.

DESCRIPTION A.

Incorporation of appropriate environmental design data is necessary to ensure proper functional performance of the system or equipment during all design modes of operation.

B.

This document indicates certain conditions to which the equipment may be exposed and may be required to i

operate, but does not provide loading combinations for the design cf safety related structures.

C.

Special cases not covered by this document may be necessary for some equipment or special locations.

See Paragraph IA.

III.

REQUIREMENTS i

Equipment shall be designed and tested to meet the requirements of 10 CFR 50, Appendix A, General Design Criteria 4 titled " Environmental and Missile Design Basis."

In order to meet the above requirements, the environmental I

conditions have been specified for three event categories:

normal, abnormal and design basis event. The three categories are defined and discussed in the following i

sections.

i 1

r 0086H 7

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1 I

10855-D7.5, Rev. 2

~

A.

Normal Conditions Normal environmental conditions are defined as those conditions existing during routine plant operations including startup, shutdown, refueling and maintenance operations. Under these conditions l

systems and components required to shutdown the reactor and maintain it in a safe shutdown condition and which also have to mitigate the consequences of a design basis event (D8E) shall be designed to remain I

functional after exposure to the most extreme environmental normal conditions that can occur and i

under which the component must function. These environmental conditions are as follows:

l 1.

Temperature, pressure, and humidity must be at least as high as the maximum design conditions s

maintained at the equipment location by the l

cooling or ventilating systems during normal operation, i

2.

Maximum expected integrated radiation doses must i

be as high as the predicted 40 years dose at the j

equipment location during normal operation (i.e.

100% load factor and 100% of rated power). The same criterion also applies to the expected i

maximum dose rates.

B.

Abnormal Conditions Abnormal conditions are those that may be experienced during the operating lifetime of the plant. These i

conditions are anticipated operational occurrences, j

such as loss-of-offsite power, see Table 1, and i

should not be interpreted to be accident conditions.

Equipment required for a safe shutdown of the plant shall be designed to remain functional during the abnormal environmental conditions to which it may be exposed as shown in Table 6.

1.

Temperature, pressure, and humidity must be at

~

least as high as the conditions maintained at the equipment location by the cooling or ventilating system under the abnormal condition.

l 2.

Abnormal event radiation doses and dose rates, i

where identified, are included in the normal i

doses and dose rates reported in Table 6.

0086H 8

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10855-D7.5, Rev. 2, The total duration of an abnormal condition will be the algebraic sum of all dependent, and succeeding events initiated by any postulated abnormal event.

For a discussion of some of these occurrences and their duration, see Section V.

C.

Design Casts Event Conditions The design basis event (D8E) conditions are those that would exist during an accident event, see Table

2. The D8E conditions shown in Table 6 are those that would exist for the most severe pipe break with the exceptions as delineated in Section I.C.3. For the duration of DBE environmental conditions two different periods are used. For temperatures, pressures and humidities a conservative duration of 100 days is applied. Ambience is reached at the end of this period. For the post-accident integrated radiation exposures a period of 180 days is used. The systems and components required to mitigate the consequences of the DBE shall be designed to remain functional after exposure to the following environmental conditions:

1.

Temperature, pressure, and humidity must be as high as expected after a D8E. The expected temperature, pressure and humidity should be applied for a period of 100 days following a D8E. For some specific components however, the temperature, pressure, or humidity may be reduced to cover just their required functional period after the accident.

2.

Maximum expected total integrated radiation j

doses (TID), in general, should be as high as predicted for a 180 day period following a LOCA.

However, for some specific components, the TID can be reduced to cover just their functional period after the accident which may be much less. The radiation source terms and calculation methodology are based on the requirements of l

NUREG 0588.

Post LOCA doses are calculated to 180 days by which time the total integrated doses have essentially i

i saturated. Not all safety related equipment is required to function for this entire period but none should fail during the 180 day period in a way that would jeopardize a safety function.

I 0086H 9

j i

lt 10855-07.5, Rev. 2 I

IV.

USE OF TABLES 1 THROUGH 7 A.

General i

Care should be exercised when extracting information from the tables. The conditions are those that would exist in the specific area during normal, abnormal, and the most severe D8E conditions. The data provided in this DITS was obtained from an area or room by room analysis of the most severe heating / cooling loads. This conservatively bounds the environment caused by postulated events. As a result, certain equipment may actually experience a less severe environment than that specified in this DITS. When detailed calculations or documentation exists to justify a less stringent environment, then these lower values may be used when necessary to qualify 2

specific equipment. Examples of these situations 4

follow:

1.

Certain equipment will not be required to operate for the entire 100 or 180 days following the D8E. The conditions specified will reflect only that time the equipment is required to operate.

2.

Some equipment which is part of the primary containment boundary and which is located outside primary containment but inside the reactor building takes suction from the drywell (e.g., hydrogen recombiner) or suppression pool (e.g., RHR pump). This equipment will have the surrounding environmental conditions specified in addition to the internal conditions (radiation, temperature, and pressure) due to its communication with primary containment.

Depending on the equipment's function, this could apply to normal, abnormal, and DBE or only D8E conditions.

3.

Materials that are purchased for several locations, such as valves, could have conditions specified for the most extreme environment any single item will experience. For example, nuclear class valves that will be located in both the drywell and reactor building could have the drywell conditions specified in the i

i OO66H 10

,,,m.,_,,,,.____ _ _ _.. _

10855-D7.5, Rev. 2 procurement documents for qualification purposes. A general envelope to be used for equipment located in more than one building is presented in Sections X.A-C.

l Some of the conditions described in 1 through 3 above are j-outside the scope of this document and will have to be developed on a case by case basis and will have to be documented in design criteria and specifications, as l

required.

B.

Margin Margin is defined as the difference between the most i

severe specified service conditions of the environment and the conditions used in type testing to account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance. The following applies to margin as addressed by this document:

l 1.

Any conservatism included in the development of 1

the conditions specified may not be considered i

margin. These tables, therefore, do not include margin. Margin must be addressed ^by the procurement documents or the vendors qualification plan.

2.

A method of including margin acceptable to the NRC is presented in NUREG-0588.

V.

ADDITIONAL ENVIRONMENTAL CONDITIONS 4

There are several additional conditions that could exist during the operating life of the plant. These conditions 4

l have the potential to affect equipment qualification.

These conditions are identified and defined along with any

limitation on their use in the following sections:

A.

Design Basis Tornado (DST)

The design basis tornado is considered to be one of i

the design basis events. However, the depressurization resulting from the tornado is not included in Table 6. Depending on the translational velocity of the tornado, the entire plant or only portions will experience the maximum depressurization j

of -3 psig (11.7 psia). The duration of the DBT is 4

fror 3 seconds up to one (1) minute.

1 OO86H 11 4

.c

10855-D7.5, Rev. 2 Although discussed above to some extent, it is not necessary to specify the DST as one of the environmental conditions in procurement documents for equipment as the depressurization is not expected to be so severe as to have an effect on the safety-related systems. A plant specific tornado depressurization study has shown that damage 4

resulting from the DBT will be limited to non pressure tight doors springing open and some HUAC ducting experiencing deformation.

8.

Equipment Submergence 1.

Drywell Submergence is generally not conside.*ed to be applicable to the drywell. Any water that would be introduced into the drywell during normal operation would be removed by the sump pumps. In the event of sump pump failure, the water level would reach a depth of approximately one (1) foot when it would begin to flow through the drywell vents to the suppression pool.

I 1

2.

Reactor Building The flood levels for the reactor building and main steam tunnel are presented in Table 7.

These levels are based on the largest moderate energy line in a compartment breaking and blowing down fo.r 30 minutes before isolation occurs. The levels shown in Table 7 include a 4

50% safety factor over the calculated levels.

The equipment function should be very carefully considered before specifying that qualification for submergence is required. In general, any equipment that has the potential of being submerged will not be required to operate after submergence.

3.

Control, Auxiliary and Turbine Buildings Internal flooding for the control and auxiliary buildings need not be considered for equipment qualification. All flooding potentials were examined during the separation review program, and it was determined that there is no accident I

l 0086H 12 i

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j 10855-D7.5, Rev. 2 scenario severe enough to flood safety related equipment and components located in these buildings. There are no essentia1' safety related components located in the turbine building.

4.

External Flooding Flooding of the reactor, control and auxiliary buildings from external sources need not be considered as all entrances, openings etc. into these buildings have been designed watertight up to the highest flood level of elevation 126.2 feet. There are no essential safety related equipment or components in the turbine building.

l C.

Dust j

l In a nuclear power plant there are three major sources of dust: transport from outside the plant through the ventilation system; deterioration of l

uncoated concrete floors and; residue from maintenance activities. However, for the reasons discussed below, dust is not expected to be a factor in equipment qualification.

1.

Plant Ventilation All major ventilation units supplying the plant with outside air have two filters in series to insure the cleanliness of the air. The first, or low efficiency, filter has a minimum rating of 55% removal and the second, high efficiency, filter has a rating of 80-85% efficiency.

Certain HVAC units such as the reactor building ventilation system (ROVS) have a higher rated first filter for additional air cleanliness. The ducting for these systems is arranged in such a manner that the recirculated air is mixed with the~outside air before the HVAC filters. This will guard against any contamination from i

compartments where dust may exist. The pressure drop across the filters is monitored by pressure differential indicators so that the filters may be replaced to ensure that the air quality will not be degraded.

i l

OO86H 13 4

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10855-D7.5, Row. 2 Concrete Coatings 2.

In general, areas containing equipment that may be sensitive to dust, such as electrical switchgear, will have the floors coated with a sealing medium. Surfaces that are not coated in the above manner but are subject to radiological contamination are coated with a material intended to ease the decontamination process which will also minimize dust generation.

3.

Housekeeping Housekeeping, cleanliness, and fire protection administrative procedures will prevent the accumulation of debris during normal operation and subsequent maintenance.

4.

Equipment Maintenance The equipment suppliers of equipment known to be sensitive to dust shall be requested to provide cleaning and maintenance procedures to prevent dust accumulation.

D.

Non-Seismic Vibration i

Individual equipment vibration is not anticipated to i

be a condition affecting other equipment qualification. Piping or equipment in the plant is supported so that any vibration originating at rotating equipment is restricted by the supports. Any vibration that is identified during startup testing or operation will be considered at that time and the necessary corrective measures will be taken. A separate program will be developed for addressing and analyzing the non-seismic vibrations on a case by case basis for those cases where there are no corrective measures available.

i J

14 0086H

10855"D7.5, Rev. 2 E.

Loss of HUAC All areas containing safety-related equipment with i

the exception of the diesel generator electrical switchgear, battery and remote shutdown panel rooms, l

are supplied by redundant HVAC systems. (Redundancy

~

is achieved through availability of extra equipment train.) Loss of a HVAC system is assumed to occur once per year for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per occurrence.

F.

Loss-of-Offsite-Power (LOP)

Loss-of-offsite power will occur when all power on the offsite power distribution system is lost. This occurrence will cause the diesel generators to start and supply power to those systems that are safety-related and a small number of non-safety related systems. The conditions described in the abnormal conditions section of Table 6 will occur during this time. A LOP is conservatively assumed to i

occur four (4) times during the 40 year plant life,

for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per occurrence.

VI.

REACTOR BUILDING ENVIRONMENT l

A.

Outside Primary Containment /Inside Reactor Building 1.

Normal Conditions a.

Normal conditions are listed below:

Temperature Pressure Humidity (deerees F) finch WG)

(1)

Maximum 115 1.0 90 Average varies

-0.25 40 Minimum 40

-0.25 20 l

The above conditions are for the general reactor building environment. The minimum temperature of j

40'F is both for shutdown and operating l

conditions. The 115'r temperature also envelopes the conditions existing during equipment testing. The area specific conditions, including maximum radiation dose rates and total integrated doses are shown in Table 6.

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10855-D7.5, Rev. 2 b.

Rooms which have the potential of spreading radioactive material (air, vapor, etc.) to adjacent rooms with lower concentrations of radioactivity will be kept at slightly negative pressures to prevent contamination of adjacent areas. The pressure range for these rooms is -0.1 to -0.25 inch WG.

2.

Abnormal conditions The anticipated events leading to abnormal conditions are listed in Table 1.

The abnormal conditions existing at the various locations in the reactor building are shown in Table 6.

Temperature Pressure Humidity Varies atmospheric 100% maximum 20% minimum The specific temperatures are listed in Table 6.

3.

DBE Conditions Safety related systems that are required to operate following a design basis event (DBE) shall be capable of operating in the conditions associated with that event and after perforning its function, shall not fail in a way that would jeopardize a safety function. DBE conditions; can be grouped into two categories, bulk environmental conditions and local steam conditions which are described below.

a.

Bulk Atmosphere Condition This DBE atmospheric condition is experienced uniformly throughout the entire building excluding those rooms containing high energy line breaks.

Equipment needed for safety related system operation, such as pumps, valves, motors, wiring, controls and instrumentation, as a minimum requirement shall be capable of operatfor, for the required duration and subsequently 0086H 16

10855-D7.5, Rov. 2

~

shall remain in a safe condition, when subjected to an accident environment of 148'r at 100% relative humidity for 30 minutes and 148'F and 95% relative humidity for the remainder of the 100 days. Operator action accounts for the decrease in the relative humidity after 30 minutes, b.

Local Steam conditions i

This atmospheric condition is experienced locally for a pipe break inside a room or in the path that steam takes in escaping from the room. Temperatures in these areas will be much higher than the bulk atmospheric conditions experienced i

uniformly throughout the plant.

i Under this condition, safety related system equipment shall be capable of functioning for the required duration and shall subsequently remain in a safe condition when subjected to the local (proximity to break) steam environmental conditions if the equipment is:

(1)

Required to detect a failure; (2)

Required to perform and/or maintain isolation function; (3)

Required to perform a water line isolation function and could be subjected to the steam environment, such as electrical cable or valve operator; (4)

Required for safety related system operation, and is located so a failure in some other system expbses the safety related system equipment to the local accident environment.

(5)

Required to track the post accident environment conditions, such as pressure, temperature, and radiation monitors.

0006H 17

10855-D7.5, Row. 2

~

(6)

Required to control the environmental consequences of the failure.

The local steam environment from a steam line break can have higher localized room temperatures and pressures as shown in Table 6. The maximum temperatures and pressures shown in Table 6 as X-30 minutes represent a temperature or pressure of X for 30 minutes with a temperature of 148'r and pressure of 0 psig for the remainder of the 100 day DOE, one exception is in torus compartment area. Temperatures may be either 302*F for 30 minutes, or 175'F for 9 days. 148'F would always follow for the remainder of the 100 days.

Max relative

)

humidity of 100% has a duration of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to account for the time to complete vessel depressurization.

The worst D8E conditions must be used to qualify walls, isolation dampers, and 1 solation valves which are located in these areas. Nonsafety-related equipment is not expected to function.

1 c.

Radiation Conditions 4

The radiation conditions existing inside the reactor building for a DOE are presented in Table 6.

S.

Inside Drywell 1.

Normal Conditions a.

Temperature, Pressure and Humidity Conditions Temperature Pressure Humidity l

(deerees F) feste)

(1)

Maximum 150 2.0 90 Average 135 0

50 Minimum 40 (not

-0.5 20 operating) 60 (operating) l 0086H le l

10855-D7.5, rov. 2 b.

Radiation Conditions (outside 81oshield)

Total integrated gamma plus neutron dose:

2.5 x 107 rads /40 yr. (4.2 x 106 rads g{g for neutrons).

Total integrated neutron fluence: 4.2 x 10 14 neutrons /cm2 for 40 years (11.0 MeV) 2.

Abnormal Conditions a.

Temperature, Pressure and Humidity Conditions The temperature and pressures in the drywell resulting from a loss of offsite power (LOP) event are shown in Figures 6 and 7. The humidity will be in the range of 20 to 90%.

b.

Radiation Conditions The radiation exposure does not increase above the normal operating levels during LOP event.

3.

DOE Conditions a.

Temperature, Press'are, and Humidity conditions are provided in table 5.

These environmental conditions are based on a combination of DOE's to ensure conservative b'ounding values. The 100%

Relative Humidity for 100 days is also conservative, and encompasses that environment produced by containment spray actuation.

b.

The primary containment post-LOCA total integrated doses (TID) and dose rates are provided in Table 4.

Both gamma and beta doses were calculated for the primary containment. The doses were calculated by assuming that 100% of the 0086H 19

10855-D7.5, Rev. 2 core noble gas inventory, and 50% of the core halogen inventory are initially l

released to the drywell atmosphere in accordance with NUREG-0508. The time-dependent and location-dependent distritution of the radioactivity is then calculated mechanistically using the i

i methodology described in NUREG-0508, Rev.

1, Section 1.4 and Appendix D.

The beta doses and dose rates were calculated assuming a semi-infinite cloud l

i geometry. For most components, the beta i

I doses in Table 6 can be reduced by performing a detailed study which accounts l

l for the actual geometry and thickness of the component. In addition (See Reference S), the beta contribution may be reduced or j

neglected provided:

i i

l 1.

The material of interest is enclosed' in a hermetically sealed enclosure with a thickness equivalent to at j

laast 70 mils of material with a o 7sity of one, or

[

11.

The coa.ponent will not be affected by i

changes within a distance equivalent to 70 mils of its surface with unit density material.

111.Formaterialsenclosedinahousin!y I

the beta plateout dose can general 4

be neglected.

In all cases the supplier shall justify any beta reduction factor used in equipment qualification.

4.

Section VI.8 and Tables 3, 4 and 5 are to be used for specifying the conditions to be used in equipment qualification for the primary I

containment, rieures 1 and 2 are a generic set of conditions which envelope the environmental conditions in a SWR drywell. The design conditions, other than the radiation environment, for coatings are based on a small f

4:

i 0086H 20 l

10855-D7.5, Rov. 2 reactor coolant pressure boundary (RCPB) line break in the drywell which does not immediately depressurize the reactor pressure vessel (RPV).

1 Both drywell spray trains spraying 75'F water with 100% efficiency into a drywell purged of air is conservatively assumed. These conditions are more severe than a DBE because the coating i

is subjected to a longer time at temperature and pressure followed by a high depressurization rate. The radiation environment to which coatings are to be qualified are those in Tables 3

4.

j C.

Inside Torus (Wetwell) i 1.

Normal Conditions l

i a.

Temperature, Pressure and Humidity.

T L*fj, P ( p s i n )-

H (%)

Maximum 150

+2.0 90 i

Average 100 varies varies Minimum 40

-0.5 20 t

b.

Radiation Conditions l

[

Totalingegratedgammadose:

~

3.5 x 10 rads (40y)

_gg 2.

Abnormal Conditions '

a.

The abnormal conditions for the torus areas will be enveloped by the temperature and pressure functions shown in Figures 6 and

7. Humidity range will be same as under normal conditions.

b.

Radiation Conditions The radiation exposures will not increase i

I above the normal operating levtis.

i l

1 0086H 21 1

i 10855-D7.5, Rev. 2 3.

DOE Conditions a.

Figures e and 9 show the temperature and pressures inside the torus (Wetwell) following a LOCA. Humidity will be 100% for 100 days following the DSE.

b.

Radiation Conditions For the 180 days following the DSE the total integrated doses are:

6.3 x 106 rads (Gamma)

'i 3.5 x 108 rads (Beta) jg VII.

AUXILIARY BUILDING ENVIRONMENT 4.

Radweste Areas 1.

Normal Conditions During normal conditions, the temperature in most parts of the radweste area will be between 40 and 115'F. The pressure in these areas will be atmospheric and the humidity will be between 20 and 90%. These conditions are intended to envelope the most severe normal operation conditions in these areas. Specific conditions may be less severe and are shown in Table 6.

2.

Abnormal Conditions The only abnormal condition directly and noticeably affecting the radumste areas of the auxiliary building would be a loss of HUAC.

Temperatures could rise to a maximum of 120*f.

3.

DOE Condition A number of safety related components are located at the lower elevations of the radweste areas. Also, the post-accident sampling station is located in the radwaste areas.

Temperature, pressure and humidity conditions 0086H 22

k 10855-D7.5, R09 2 l

will not be significantly more severe than during normal operation or abnormal events.

Radiation levels, however, will increase due to direct radiation shine from the reactor building and from airborne cloud activities.

Integrated 08E radiation doses for the general radwaste areas are: Gamma 1.0 Rads /180d and Beta 20 Rads /180d. (due to environmental airborne activities only) Radiation levels at the post accident sampling station area will be higher.

B.

Diesel Generator Areas 1.

Normal Conditions Tex.peratures will be maintained between 40 and 104*F, except the diesel generator and H&V equipment rooms which will reach a maximum of 120*F.

Certain other rooms such as battery rooms will have more stringent temperature requirements. Specific conditions may be found in Table 6. The total integrated dose for 40 years is 180 rads.

2.

Abnormal Conditions l

l The temperatures in the diesel generator area I

will reach 120*F during diesel generator operation. The pressure will be atmospheric but may vary between 0.25 and -0.25 inch water and the relative humidity will be between 20 and 90%. A minimum of 40*F may be reached if a loss of heating occurs. CO2 injection will result in a lower local temperature, however, no specific temperature need be specified. Fire protection equipment will be specified for this service.

3.

DBE Conditions The conditions in the diesel generator area will not be significantly more severe during a 08E than during normal operation.

The D8C radiation level in this area is due to the environmental airborne doses, which will be 0.6 rad /180d for gammas and 20 rads /180d for betas.

0086H 23

l 10855-D7.5, Rev. 2 l

C.

Service Areas 1.

Normal Conditions The temperatures will be maintained between 68 and 80*F for personnel areas and 60 and 104*F for equipment areas. The pressure in these areas will be atmospheric and the' relative humidity will be between 20 and 90%.

The total integrated, 40 year dose for these areas will be 180 rads and 880 rads for uncontrolled and controlled areas respectively. Specific conditions for these areas can be found in Table 6.

2.

Abnormal Conditions A minimum of 40'F may be reached if a loss of heating occurs.

3.

D8E Conditions The D8E radiation level in this area is due to the environmental airborne doses. They are 0.6 rad /180 days for gammas and 20 rads /180 days for betas.

D.

Control Area and Technical Support Center 5

1.

Normal Conditions Temperature in the control room will be maintained at 76*F i2*F, with the relative humidity between 40 and 50%.

The pressure will always be higher than the surrounding areas. The total integrated dose will be 180 rads for 40 years.

2.

Abnormal Conditions.

The control room will maintain a temperature of f

76*F 12*F during abnormal conditions. The pressure and humidity ranges also will not differ from those during normal operation. Other locations in the control area will be maintained at different temperatures depending on specific services. These conditions are presented in Table 6.

0086H 24

I 10855-D7.5, Rev. 2 3.

DBE Conditions The temperatures in the control room will be maintained the same as those for normal l

operation. The pressure will be positive to guard against the possibility of inleakage of airborne contamination. The radiation dose resulting from the DBE will be less than 20.0 reds. Humidity ranges will not differ from those during normal operation.

VIII.

TURBINE BUILDING ENVIRONMENT A.

Normal Conditions During normal operation, temperatures will be maintained between 40 and 104*F, except in some of l

the high radiation areas where a maximum of 120*F will be maintained. Pressure will be 1" water gauge maximum -0.1" water gauge minimum.

Relative humidity will be between 20 and 90%. These general conditions l

are specified to envelope the most severe normal operation conditions of any in the area. Specific areas may have less severe conditions as identified in Table 6. The total integrated doses for the various areas in the turbine building are given in Table 6.

8.

Abnormal Conditions The only safety-related components located inside the turbine building are reactor protection system f

switches on the main turbine and condenser. None of' L

the anticipated abnormal events listed in Table I will create environmental conditions inside the turbine building which could prevent the reactor l

protection system from performing its safety i

function.

C.

DBE Conditions The only safety-related components located inside the turbine building are reactor protection system switches on the main turbine and condenser. None of the postulated design basis events listed in Table 2, will create environmental conditions inside the turbine building which could prevent the reactor protection system from performing its safety function.

0086H 25 i

l

10855-D7.5, Rev. 2 IX.

INTAKE STRUCTURE A.

Normal Conditions J

During normal conditions, the temperature in the l

intake structure (except fan rooms) will be between 40 and 104*F.

During shutdown conditions, the

}

temperature will be maintained at 60*F or higher. The l

pressure in this structure will be atmospheric and the humidity is not controlled and will be the same as ambient. The total radiation exposure for this area will be 180 rads over the 40 year design life.

The room specific temperatures can be found in Table 6.

B.

Abnormal Conditions The abnormal conditions are shown in Table 6. The pressure will be atmospheric and humidity will be between 20 and 100%.

C.

DBE Condit1uns The temperatures, pressures and humidity will be the same as for normal operating conditions. The radiation doses in this area are negligible.

X.

GENERAL BUILDING ENVIRONMENT There are specifications which cover equipment located in several areas and buildings throughout the plant. To avoid assigning different environmental conditions for each piece of equipment, a single set of conditions may be assigned to cover all areas except the intake structure.

It should be noted that these conditions are specified to envelope the most severe conditions and individual areas l

may have less severe conditions. If necessary, a note will '

be added in procurement documents stating that these conditions will be used in qualification unless area specific conditions are used.

A.

Normal Conditions Temperature Pressure Humidity l

l (dearees F)

(inch WG)

(%)

Maximum 120 1.0 90 Minimum 40

-0.25 5

(5* room 5423)

I 0086H 26 1

10855-D7.5, Rev. 2 Total integrated dose: 2.0 x 107 rads /40y B.

Abnormal Conditions 4

Temperature: 148'F (195'F room 4416)

Pressure: +1.0 inch WG to -0.25 inch WG Humidity: 100%

C.

D8E Conditions Temperature:

340*F for 30 minutes 175'F from 30 minutes to 9 days 148'F from 9 days to 100 days Pressure:

16.3 psig for 30 minutes O psig from 30 minutes to 100 days Humidity:

100% for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 95% 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 100 days t

DBE integrated dose:

I Shine:

3 x 107 rads (180 days)

Airborne:

2.5 x 104 rads (180 days)

Gamma 1.1 x 106 rads (180 days) 8 eta Examples of areas outside the drywell that exceed 50 rads /hr or 2.0 x 10+7 rads /40 yrs are:

RWCU backwash receiver TK room, RWCU phase separator room, Spent fuel storage pool, RWCU filter /demin cubicle, i

Air ejector room Waste sludge separator room Primary recombiner cells Offgas holdup delay pipe chase Drum storage hayloft 0086H 27

)

l

1 e

10855-D7.5, Rev. 2 Table 6 shall be used to determine the exceptions to the above conditions. The above conditions shall not be used for equipment which will only be placed in a single location. A sample environmental data sheet to be used with each specification is shown in Figures 3 and 4.

XI.

GENERAL CRITERIA, STANDARDS AND GUIDES A.

The following IEEE Standards, NRC documents, and Hope Creek project specifications are to be referred to as applicable for environmental qualification of equipment. However, it is not to be considered all inclusive.

1.

IEEE Std 278, dated 1967 Guide to Classifying Electrical Materials Exposed to Neutron and Gamma Radiation 2.

IEEE Std 317, dated 1976 Electrical Penetration Assemblies in Contain ment Structures 3.

IEEE Std 323, dated 1971 Standard for Qualifying class IE Equipment for Nuclear Power Stations 4.

IEEE Std 334, dated 1974 Type Test of Continuous Duty Class I Motors 5.

IEEE Std 382, dated 1972 Type Test of Class I Electric Valve Operators 6.

IEEE Std 383, dated 1974 Class IE Electric Cables, Field Splices, and Connections 7.

Hope Creek Specifi-Environmental Qualifi cation 10855-G-013(Q) cation of Safety Related Equipment 8.

Regulatory Guide 1.89, Qualification of Class November 1974, Rev. O IE Equipment for Nuclear Power Plants d

OO86H 28

10855-D7.5, Row. 1 9.

NUREG 0588 Interim Staff Position on Environmental Qualification of Safety Related Electrical Equipment 8.

Ducts and dampers used for ventilation are subjected to different environmental conditions:

1.

Normal conditions inside and outside the ducts and 2.

Conditions inside and outside the ducts for a DBE.

In order for the manufacturer to properly qualify the 4

equipment to the most stringent conditions all of the above information must be given on a data sheet shown in Figure 4.

C.

All reactor building isolation valves must be bubble tight for a negative pressure of 0.25 inch WG.

XII.

REFERENCES A.

GE Documents 22A2928 Revision 2 "6WR Equipment Environmental Interface Data".

8.

GE Document NEDO-10698 " Environmental Qualification of Class I Control and Instrumentation Equipment".

C.

8echtel Design Guide N2.2.2 " Pressure and Temperature conditions for Environmental Qualifications of Coatings Inside 8WR Containments with a Generic Set of Conditions for both BWRs and PWRs".

E.

Hope Creek Design Criteria 10855 D2.2, " Design Criteria for Reactor Building".

F.

Hope Creek Design Criteria 10855-03.47 " Turbine Building Heating, Ventilating, and cooling Systems (HAUC)".

G.

Hope Creek Design Criteria 10855-D3.48, " Reactor Building HUAC".

1 OO86H 29

l 10855-D7.5, Rev. 2 H.

Hope Creek Design Criteria 10855-03.48A, " Technical Support Center HVAC".

I.

Hope Creek Design Criteria 10855-D3.49, "Drywell Cooler System".

j 3.

Hope Creek Design Criteria 10855-D3.50, " Auxiliary f

Building Control Area HVAC".

)

l K.

Hope Creek Design Criteria 10855-D3.51, " Auxiliary Building Diesel Generator HUAC".

L.

Hope Creek Design Criteria 10855-D3.52, " Auxiliary Building Radwaste Areas HVAC".

M.

Hope Creek Design Criteria 10855-D3.53, " Auxiliary Building Service Area HUAC".

N.

Hope Creek Design Criteria 10855-03.54,

" Administrative and Miscellaneous Areas HUAC".

O.

Hope Creek Design Criteria 10855-D3.57A, " Service Water Intake Structure and Miscellaneous Yard Buildings HUAC".

f P.

Nuclear Regulatory Commission Regulatory Guide 1.3

" Assumptions Used for Evaluating the Potential Radiological Consequences of a Less-of-Coolant Accident in a Boiling Water Reactor".

Q.

Nuclear Regulatory Commission Regulatory Guide 1.76

" Design Basis Tornado for Nuclear Power Plants".

R.

Nuclear Regulatory Commission Regulatory Guide 1.117

" Tornado Design Classification".

S.

NRC Bulletin No.79-018, " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors," January 14, 1980.

1 0086H 30

--,-,.,-y-n,..e-.

10855-D7.5, Rev. 2 Table 1 Anticipated Abnormal Events

+

Inadvertent Closure of MSIU Inadvertent Closure of All MSIUs Pressure Regulator Fails Open Loss of Aux. Transf. (Loss of Power)

Loss of Grid Conn. (Loss of Power)

Grid Tie-Line Disturb. & Recou.

Loss of Condenser Vacuum Turbine Trip without Bypass Turbine Trip with 8ypass Ger;erator Load Reject w/o Bypass Generator Load Reject w/8ypass l

Single SRU Opens-Depressurize (SORU)

Trip of both Recirc. Pumps Recirculation Failure Decrease Flow Recirculation Failure Increase Flow Feedwater (FW) Controller Fails-Max. Demand Loss of All FW Flow Loss of FW Heaters-Auto Loss of FW Heaters-Manual Trip One FW Pump & Recovery Worst in Sequence Rod Error OO86H 31

=

10855-D7.5, Rev. 2

~

Table 1 cont.

Instrument Ranging Error Rod Withdrawn Erro.- at Power Inadvertent (or manual) Scram Inadvertent HPCI Injection Table 1 (Cont'd)

Inadvertent RCIC Injection Loss of HVAC in Steam Tunnel Loss of HVAC in Drywell Loss of HVAC in Aux. Building Loss of HVAC in Reactor Building Loss of HVAC in Turbine Building O

e e

OO86H 32

1 10855-D7.5', Rev. 2 Table 2 Fostulated Design Basis Events Small High Energy (H.E.) Pipe Break in Drywell Small H.E.

Line Pipe Break in Reactor Building (R.B.)

Outside Drywell Large H.E.

Pipe Break in Drywell Large H.E. Pipe Break in R.B. Outside Drywell i

Large H.E.

Pipe Break in Outside R.B.

Open Recirc Valve in Cold Loop (Reverse Flow)

Start Recirc Pump in Cold Loop (Forward Flow)

Seizure of Recirc Pump Inadvertent LPCS Injection Inadvertent ADS-Depressurize i

Worst ATWS-MSIV Closure No Scram, Two Pump Trip Control Rod Accident (Drop)

Reactor Overpressure Backup Scram Reactor Drain Shutoff Improper Shutdown of Plant (RHR Suct thru S/RU and Supp Pool)

Improper Startup of Plant Hot RWCU Fuel Handling Accident Tornado 0086H 33

.~.

10855-D7.5, R,v.2 nesign poclo coext r.xvironnextc1 Conditions f

a InoIde Prleary Centainment t

Condition Canponent I

1 Core spray injection Temperature 349'r 349'r 320*r 259*r 299'r

!'l' eheck valve Freneure

-2 to 62 pelg

-2 to 35 palg

-2 to 35 pelg 9 to 25 psig 9 to 20 pelg

(

'l t.PCI-pua Injection met. humidity ISS percent ISS percent ISS percent 199 percent ISS percent j i check valve, puration (1) 45 seconds 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 6 hosers I day 199 days meactor sheetdown i

cooline suctina j

valwe includtng operator and cable, i

Relief valve Includirup operator and cable, j l Vesset level indicator l

Structural components (e.g., loop restraints, s

vessel skirt, etc.)

1 4

2 Feedwater Check Temperatiste 349'r 349'r 329'r

valves, Pressure

-2 to 62 pelg

-2 to 35 pelg

-2 to 35 pelg HvCI stese line met. humidity ISS perces.t ISS percent ISS percent loolation valve pescation 45 seconds 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 6 hours includisw) operator and cable,

}

RCIC steam line 1

Spolatlan valve Includlews operator j

and cable, Reactor Water Clean-up nuction valve I

including operator I

and cable, I

meactor water i

nample line valve i

including operator i

and cable, r,Ines 2 inches and i

l maaller (Isolation Valves, Operatorn, 3

Cabling),

Cdiles to intermediate range penitorti and power range menitot i

meneter vessel head spray isolation valve includiews operator and cable i

I I

R33/9

3y 4

-- ~.

_ ~-

..~

evues-ve.4, eiew. 4 Tablo 3 (Cont'G)

Design Casic Event EIvironmentc1 Conditionn i

inside Primary Containment i

condition Component f

3 Main steam teclation Temperature 340*F 340*F 1

valves includines Pressure

-2 to 62 psig

-2 to 35 pelg operator and cable, met. humidity 100 percent 100 percent Main steam drain Duration 45 seconde 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months i-toolation valves

?

including operator and cable, i

Standby liquid contron injeetlos check valve 1

i l

4(2)

Recirculation valves Temperature 340*F 340*F 320*F (main valves, hypass Pressure

-2 to 62 pelg

-2 to 35 peig

-2 to 35 poig I

valves, equaliser Ret. heetetty 100 percent 100 percent 100 percent valve) including Duration 45 seconde 3 houre 4.5 houre I

operators and cables, neactor Protect ton i

i System t

Neutron Monitoring System i

)

l

=

l 5

i 3[

{

10855-D7.5, Rev. 2 Desiga saata r.vext hvisonmental caxditions Inside Primary Cont.alsument Conditton Crepanent valven not registred to he operahle but meest not fall to open uneler the fofleiring conditinna 5

Feedwater e; heck Temperature 250*r 200*r valve. IIPCI and Prensiere

-2 to 25 palg

-2 to 20 psig steam Ilne istoation Ret. humidity 100 percent 100 percent valven including Duration 1 day 100 days operators and cables, pectreolatlan valves (main valves hypass valves, equalizer valves), includi nel operatore and cables, peactor vennel head spray isolation valve including operator and cable, Reactor water clean-up suction valves includinel operator and cable, peactor water nample line valven including operator and cable, f.ines 2 inches and smaller (inolation valves, operators, cabling) 6 Main steam isolation Neaperature 340*r 320*r 250*r 200 *r valves including Pressure

-2 to 35 pelg

-2 to 25 psig

-2 to 20 pasig operator and cable, Rel. humidity 100 percent 100 percent les percent 100 percent Main steam drain Disration 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months 6 hours-I day 100 days Isolatlon watwas including operator and cable, Stendhy ligisto control injection check valve (1) Duratione showet are termination times measured from the Initiatlon of the postulated accident, i.e. Condition 1, the 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months duration in the period from 45 seconds throtegh 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , the one day duration is the period free 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months through I day (24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

(2) Por the rectreislation valves to perform their matety tienetton they munt clone followinre a recirculation line break, no that the core flooding can he carried out in the required time.

Por this safety requirement the environmental conditions ullt esot exceed 3fn*r at 62 pe lg for 1/2 hour. The specified conditions in (4) above are to enable a normal vessel shutdown cooling procedure durine a steam leak.

34

10855-D7.5, Rev. 2 Table 3 (Cont'd) t The following is a compilation of basic DBE environmental pressures and temperatures together with the time durations expected. The full spectrum of simultaneous environment possibilities is not presented in a series of curves, but rather as an exposition of the boundaries within which designated equipment must operate at discrete times during the cycles / modes of the reactor's operation.

Temperatures 340*F Upper boundary on maximum superheat temperature for a steam leak with the reactor vessel at 400 - 500 psi, containment at 50 psia 320*F Maximum superheat temperature during shutdown cooling line flush after reactor has been depressurized to 150 psia 250*F Maximum long term temperature in the containment during the first day following a postulated Design Basis Event j

200*F Extended long term temperature in the containment following a postulated Design Casis Event Pressures

-2 psig Assumed minimum pressure of the primary containment

+62 psig Positive design pressure of the primary containment 35 psig The containment pressure corresponding to all the non-condensibles initially in the drywell being transferred to wetwell 37 0087H

10855-D7.5, Rov. 2 Table 3 (Cont'd) 25 psig Upper boundary on long term pressure response up to one day following a postulated Design Basis Event 20 psig Upper boundary on extended long term pressure at one day and longer following a postulated Design Basis Event Durations 45 seconds Conservative time duration to cover peak containment pressure 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months conservative time duration during which valves that must isolate automatically on low Reactor Pressure Vessel pressure or high drywell pressure, must be operable 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Conservative duration of time to depressurize the Reactor Pressure Vessel at a rate not exceeding 100 degrees / hour, down to 150 psia 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Conservative time at which shutdown cooling system flush is complete. Normal shutdown cooling necessitates closure of recirculation line valves.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Conservative duration of time to complete vessel depressurization to approximate containment pressure. This time includes Reactor Pressure Vessel depressurization to 150 psia not exceeding a rate of 100*F per hour, flushing of system and depressurization to approximate containment pressure.

38 OO87H

20055-D.75 Rev.2.

Tablo 4 1

Radiation Conditions Inside Primary Containment kadiation Operating Dose Design Basis Event Integrated Dose (1,2,5)

Area (fig.5)

TYPE mate (1)

Type Dose mate (1,3)

Normal Det Drywell, Inside Gamma 2.3 x 104 LOCA S.1 a 109 Vessel Shield Neutron 0.0 a 103 SEE NOTE 6 3.1 x 109 SEE 900TE 7 Beta (4)

Outside vessel Shield 1

zone 1 Gamma 2.1 m 10 LOCA 7.4 a 106 Above Core Neutron 3.0 SEE NOTE 6 1.1 a 100 SEE NOTE 7 Beta (4) j sone 2 Gamma 59 LOCA 1.1 a 107 Core Region Neutron 12 SEE 180FE 6 4.2 a 10' SEE NOTE 7 Seta (4)

W I

to sone 3 Gamma 54 LOCA 1.9 a 100 under vessel taeutron 12 SEE NOTE 6 4.2 s 10 SEE NOTE 7 Beta (4)

Sone 4 Gamma 44 LOCA 1.5 a 107 Near Recirc.

Neutron Q.6 SEE NOTE 6 2.1 a 105 SEE soort 7 Beta (4) sone 5 Gamma 15 LOCA 5.3 a 106

> 15 ft. Recirc.

Neutron 1.1 SEE NOFE 6 3.9 a 105 SEEIsoft 7 seta (4) at 1.0 (note 8)

LOCA 5.1 a 104 48 hr 3.5 a 10 6.3 a 106 5

sone 6 Gamma 4

Toru s Neutron 0.1 3.5 a 10 seta (4) 1.5 a 106 at 3.5 a 108 48 hr 1.

The unit of dose rate is made/tr. The units 5.

Gamma, beta, and neutron radiation shall all of dose is rade.

he considered for equipment quellfication inside the drywell.

2.

Normal integrated dose is calculated for a peraod of 40 years.

6.

Gamma 1.9 a 107 (airborne) 1.9 a 105 (plateout at I hr) 3.

DOE dose rate is the dose rate lamediately following the Det except otherwise specified.

meta 3.1 s 108 (airborne) 1.1 a 107 (plateout at I hr) 4.

The beta dose is not significant compared to others during normal operation.

7.

Gamma 2.6 a 107 (airborne) 3.4 x 106 (plateout)

Beta 9.5 a 108 (airborne)

K9/19 6.7 a 108 (plateout) 8.

20' away from 20" recire. line

10855-D7.5 Rov. 2 0

9 Table 5-Environmental Conditions Inside Primary Containment for the Design Basis Event Drywell Temperature Pressure Humidity Time (Deerees F)

(psia)

(%)

0 - 20 sec 340 0 - 62 100 20 sec - 5 min 340 62 100 5 min - 3 hr 340 40 100 3 - 6 hr 320 40 100 6 - 24 hr 250 25 100 1 - 4 days 200 25 100 4 - 180 days 200 10 100 4

'1 40 OO67H i

10855-07.5, Rev. 2 Table 6 Normal and Maximum Plant Environmental Conditions 4

INDEX Reactor Building Pages 42 - 50 Turbine Building Pages 51 - 58 Auxiliary Building Pages 59 - 79 Intake Structures Page 80 Notes Page 81 41 OO87H

i 1

4 1

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I 5

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ML20125B183

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